田湾核电站主数据管理创新

针对主数据管理面临的持续发展问题,通过对田湾核电站主数据全生命周期管理、主数据移交管理、主数据结构关系以及主数据治理流程的设计与描述,阐述了田湾核电站实现主数据持续发展与改进的理念与措施,首次完整地将主数据管理与核电站实体建设相结合,并将相关管理理念融合在一起,明确主数据管理的演进路线与发展方向。     

嵌入式数据库SQLite在核仪器数据管理中的应用

针对核仪器中采用文件方式管理二进制谱线数据块时操作繁杂、格式不一、升级困难等问题,本项目通过嵌入体积小、速度快、稳定可靠的SQLite数据库,采用BLOB类型存储二进制谱线数据块,利用事务传输批量数据,提高了谱线数据管理性能与安全性。根据谱线数据管理要求,设计实用的谱线数据管理系统,验证SQLite数据库在核仪器数据管理中高效、便捷、可靠等优势。     

核燃料产品全生命周期数据管理系统的实现

上海核工程研究设计院在西门子Teamcenter(TC)系统上开发出核燃料产品全生命周期数据管理系统(SNPLM)。SNPLM涵盖了核燃料产品的材料研发、结构设计、试验验证、制造、辐照历史和后处理共6个数据范畴,采用客户端和服务器构架(CS),使数据库层、应用层和用户层3层隔离,提高了数据存储效率,保障了数据的安全。SNPLM具有加强的数据模型、检索系统、需求跟踪系统和设计分析接口系统,为核燃料产品的研发提供了平台。     

小堆发展的法规标准问题分析及建议

当前小堆从设计研发阶段正逐步推向市场应用。小堆自身特点和设计理念与传统大型核动力堆不尽一致,这对我国当前的法规标准的优化和改进提出了挑战。文章介绍了法规标准在小堆方面的安全要求,分析了当前小堆发展的法规标准存在的共性问题,并提出了我国法规标准支持小堆发展方面的优化和改进建议,可供我国核电行业法规标准后续建设进行参考。     

海洋工况下海洋小堆热工水力特性分析

海洋的开发,对核能有很大的需求空间.相较于陆基堆,海洋小堆在堆形的结构设计、设备及系统布置等,有着不同的特征.为适应船舶舱室与海洋环境,本文参照某型模块化小堆的主要设备结构和系统布置,得到一种改进后的海洋一体化小堆设计方案和设计参数.使用添加了海洋工况计算模块的RELAP5软件,对其中一个设计方案,引入海洋工况;通过和...     

美俄小型水冷堆发展现状及趋势

<正>2023年11月,俄罗斯国家原子能集团宣布已批准RITM-200N小型水冷堆的技术设计,首座陆基小堆电厂部署达到重要里程碑节点。国际原子能机构数据显示,全球已有超80种小堆设计,其中美俄型号数量位居世界前列。在各类小堆技术路线和设计中,小型水冷堆成熟度高、研发应用进展快,成为美俄近期推动研发和出口的重点。     

LabVIEW在西安脉冲堆辐射剂量监测与数据管理中的应用

西安脉冲堆辐射剂量监测及数据管理系统用单片机完成信号的采集处理及数据传送,用RS-485总线构成分布式的测试网络,用功能强大的LabVIEW软件开发了上位机监测及数据管理系统软件,实现了脉冲堆各个区域辐射剂量实时监测及剂量数据库管理功能.系统的试运行表明,系统运行稳定可靠,达到了设计指标,能够用于实际应用.     

核电厂企业内容管理信息系统建设

数据已逐渐成为企业可持续发展、提高决策水平和创新能力的战略性资源;企业数据量快速膨胀,并呈现出来源多样化和类型复杂化的特点,传统关系数据库管理系统（RDBMS）处理的结构化数据仅占数据信息总量的15%,而全球85%的信息是非结构化的。如何管理这些非结构化信息,成为传统结构化数据管理的一大难题,而企业内容管理系统恰好是解决这一问题的有效方式。文章通过分析国内部分核电企业内容管理系统现状,提出了企业内容管理系统选型方面的建议,从实际应用出发探讨了核电厂企业内容管理系统的建设。     

环形燃料元件小堆的概念设计

环形燃料元件小堆是世界上目前比较先进的堆型。研究设计了一个环形燃料元件小堆,开发出适于环形燃料堆计算的软件和方法。采用整组件束棒计算堆芯少群参数的方法大大提高了计算精度。计算了堆芯的有效增殖系数、所有控制毒物的单个价值以及总价值、堆芯从室温到工作温度的温度效应等堆芯参数。结果表明:设计的环形燃料元件堆具有良好的稳定性和安全性,可以作为一代新堆。     

BESⅢ实验离线数据管理系统的设计和实现

北京正负电子对撞机Ⅱ(BEPCⅡ)上的北京谱仪Ⅲ(BESⅢ)是用τ-粲物理研究的一台大型通用磁谱仪,它将产生PB量级的实验数据.主要研究BESⅢ实验数据在离线数据处理和物理分析过程中的管理.首先分析了BESⅢ离线数据管理系统的需求,然后详细介绍基于J2EE和Web技术的数据管理系统的整体软件架构和各个系统模块的功能实现.最后,给出了BESⅢ数据管理系统的硬件构成、软件部署以及测试结果.     

Materials and design concepts of gas-cooled reactor systems

The family of gas-cooled reactors being developed in the United States by Gulf General Atomic consists of the steam-raising and direct cycle versions of the high temperature gas-cooled reactor (HTGR) for electric power generation, the hydrogen-producing HTGR for chemical process applications, and the gas-cooled fast reactor (GCFR), a high gain breeder. The aim of this paper is to describe the underlying design concepts that are common to all of these reactors and relate these design concepts to the choice of both structural and fuel materials for the wide variety of environmental conditions encountered throughout the world. Interwoven with this discussion are typical examples of the interaction of design activities and materials selection required to give a reactor system of maximum safety and reliability, favourable environmental features, and minimum cost.     

Flexible conversion ratio fast reactors: Overview

Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. The performance achievable by the unity conversion ratio cores of these reactors was compared to an existing supercritical carbon dioxide-cooled (S-CO₂) fast reactor design and an uprated version of an existing sodium-cooled fast reactor. All concepts have cores rated at 2400 MWt. The cores of the liquid-cooled reactors are placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchangers (IHXs) coupling a primary coolant to a compact and efficient supercritical CO₂ Brayton cycle power conversion system. The S-CO₂ reactor is directly coupled to the S-CO₂ Brayton cycle power conversion system. Decay heat is removed passively using an enhanced reactor vessel auxiliary cooling system (RVACS) and a passive secondary auxiliary cooling system (PSACS). The selection of the water-cooled versus air-cooled heat sink for the PSACS as well as the analysis of the probability that the PSACS may fail to complete its mission was performed using risk-informed methodology. In addition to these features, all reactors were designed to be self-controllable. Further, the liquid-cooled reactors utilized common passive decay heat removal systems whereas the S-CO₂ uses reliable battery powered blowers for post-LOCA decay heat removal to provide flow in well defined regimes and to accommodate inadvertent bypass flows. The multiple design limits and challenges which constrained the execution of the four fast reactor concepts are elaborated. These include principally neutronics and materials challenges. The neutronic challenges are the large positive coolant reactivity feedback, small fuel temperature coefficient, small effective delayed neutron fraction, large reactivity swing and the transition between different conversion ratio cores. The burnup, temperature and fluence constraints on fuels, cladding and vessel materials are elaborated for three categories of material - materials currently available, available on a relatively short time scale and available only with significant development effort. The selected fuels are the metallic U-TRU-Zr (10% Zr) for unity conversion ratio and TRU-Zr (75% Zr) for zero conversion ratio. The principal selected cladding and vessel materials are HT-9 and A533 or A508, respectively, for current availability, T-91 and 9Cr-1Mo steel for relatively short-term availability and oxide dispersion strengthened ferritic steel (ODS) available only with significant development.     

Influence of dynamic sodium environment on the creep-fatigue behaviour of Modified 9Cr-1Mo ferritic-martensitic steel

The use of liquid sodium as a heat transfer medium for sodium-cooled fast reactors (SFRs) necessitates a clear understanding of the effects of dynamic sodium on low cycle fatigue (LCF), creep and creep-fatigue interaction (CFI) behaviour of reactor structural materials. Mod. 9Cr-1Mo ferritic steel is the material of current interest for the steam generator components of sodium cooled fast reactors. The steam generator has a design life of 30-40 years. The effects of dynamic sodium on the LCF and CFI behaviour of Mod. 9Cr-1Mo steel have been investigated at 823 and 873 K. The CFI life of the steel showed marginal increase under flowing sodium environment when compared to air environment. Hence, the design rules for creep-fatigue interaction based on air tests can be safely applied for components operating in sodium environment. This paper attempts to explain the observed LCF and CFI results based on the detailed metallography and fractography conducted on the failed samples.     

Perspective channel-type reactor with enhanced safety

Following the search for new design solutions to develop within the framework of channel trends the reactor with enhanced safety the Research and Development Institute of Power Engineering has developed the design of the multiloop boiling water reactor (MKER). The MKER enhanced safety is attained when involving the inherent safety features, passive safety systems as well as the accident consequences confinement devices. The design realizes several advantages which are typical of the channel-type reactors, namely: the design desintegration simplifying the manufacture, control, equipment delivery and decreasing, versus the pressure vessel reactors, the accident effect if it proceeds in an explosive manner; small operating reactivity margin and fuel burnup increased due to continuous refuelling; fuel cycle flexibility allowing comparatively easily to adopt the reactor to the conjuncture of the country fuel balance; multiloop circuit of the main coolant which reduces the degree and effect of the accidents connected with the equipment and pipings rupture; monitoring of the channels and fuel assemblies leak-tightness.     

Fracture mechanics issues for irradiated blanket structures

This paper reviews some of the factors that will affect fracture behavior of fusion reactor structures and summarizes some component life predictions based on linear elastic fracture mechanics analysis. The review includes discussion of the environments to which the components will be subjected, the response of materials to these environments, the time dependent nature of the structural response, and the fracture related failure mechanisms.Radiation environments and complex loading conditions in a fusion reactor cause a variety of material phenomena. These phenomena include irradiation swelling and creep, strength changes due to matrix hardening, helium embrittlement, and surface effects such as sputtering and blistering.The interaction of thermal creep, irradiation creep, and swelling results in complex time, temperature, and neutron fluence dependent stress histories in first wall and blanket structures. These effects reduce compressive thermal stresses during the burn portion of a reactor operating cycle and result in residual tensile stress during the non-burn portion of the cycle. The cyclic nature of these stresses, particularly in a tokamak reactor, and the presence of undetected flaws provide a basis for the application of fracture mechanics. Linear elastic fracture mechanics analysis techniques have been applied to predict component life for several conceptual tokamak fusion reactor designs. These analyses show that the structural life may be limited by growth of initial flaws to a coolant leakage. Results indicate that for neutron wall loadings below 2 to 3 Mw/m², life is likely to be controlled by stresses during the burn period and, at higher wall loadings, by residual stresses during the non-burn period.Fracture toughness properties tend to be reduced by irradiation. Therefore, brittle fracture will be a potentially critical failure mode. Fatigue crack growth and fracture characteristics of the design will affect the operating mode of a reactor and influence the performance of different types of reactors. Tests are currently planned to develop material crack growth and fracture toughness data [1] for candidate alloys because these properties have been shown to be important.     

系统工程方法论在核反应堆数字实验中的应用研究

核反应堆工程实验系统的复杂性一直是制约核反应堆工程实验技术攻关和创新的重要因素之一。为提升应对核反应堆工程实验系统复杂性的能力和手段,引入数字实验概念,目的是建立适用于核反应堆工程实验全生命周期的统一高效的业务执行环境。本文基于系统工程方法论详细阐述了数字实验平台的顶层架构,包括由V模型和业务场景图构成的业务流程架构,由数据模型化知识化逻辑图构成的实验基础架构,以及由业务层的业务管理系统、应用层的实验设计仿真环境系统、知识层的实验知识系统和资源层的基础功能系统构成的平台功能分层架构,并以“华龙一号”（ACP1000）二次侧非能动余热排出系统（PRS）实验系统为对象进行了应用验证。验证结果表明:上述的架构具有较强的可行性,可作为数字实验平台开发的整体逻辑框架。      

Activation characteristics of candidate structural materials for a near-term Indian fusion reactor and the impact of their impurities on design considerations

Activation analyses play a vital role in nuclear reactor design.Activation analyses,along with nuclear analyses,provide important information for nuclear safety and maintenance strategies.Activation analyses also help in the selection of materials for a nuclear reactor,by providing the radioactivity and dose rate levels after irradiation.This information is important to help define maintenance activity for different parts of the reactor,and to plan decommissioning and radioactive waste disposal strategies.The study of activation analyses of candidate structural materials for near-term fusion reactors or ITER is equally essential,due to the presence of a highenergy neutron environment which makes decisive demands on material selection.This study comprises two parts; in the first part the activation characteristics,in a fusion radiation environment,of several elements which are widely present in structural materials,are studied.It reveals that the presence of a few specific elements in a material can diminish its feasibility for use in the nuclear environment.The second part of the study concentrates on activation analyses of candidate structural materials for near-term fusion reactors and their comparison in fusion radiation conditions.The structural materials selected for this study,i.e.India-specific Reduced Activation Ferritic-Martensitic steel(IN-RAFMS),P91-grade steel,stainless steel 316 LN ITER-grade(SS-316 LN-IG),stainless steel 316 L and stainless steel 304,are candidates for use in ITER either in vessel components or test blanket systems.Tungsten is also included in this study because of its use for ITER plasma-facing components.The study is carried out using the reference parameters of the ITER fusion reactor.The activation characteristics of the materials are assessed considering the irradiation at an ITER equatorial port.The presence of elements like Nb,Mo,Co and Ta in a structural material enhance the activity level as well as the dose level,which has an impact on design considerations.IN-RAFMS was shown to be a more effective low-activation material than SS-316 LN-IG.     

Innovative nuclear technology based on modular multi-purpose lead–bismuth cooled fast reactors

Today's nuclear power is in the state of an intrinsic conflict between economic and safety requirements. This fact makes difficult its sustainable development.

One of the ways of finding the solution to the problem can be the use of modular fast reactors SVBR-75/100 cooled by lead–bismuth coolant that has been mastered in conditions of operating reactors of Russian nuclear submarines.

The inherent self-protection and passive safety properties are peculiar to that reactor due to physical features of small power fast reactors (100 MWe), chemical inertness and high boiling point of lead–bismuth coolant, integral design of the pool type primary circuit equipment.

Due to small power of the reactor, it is possible to fabricate the whole reactor at the factory and to deliver it to the NPP site in practical readiness by using any kind of transport including the railway.

Substantiation of the high level of reactor safety, adaptability of the SVBR-75/100 reactor relative to the fuel type and fuel cycle, issues of non-proliferation of nuclear fissile materials, opportunities of multi-purpose usage of the standard SVBR-75/100 reactors have been viewed in the paper.     

Utilization of Refractory Metals and Alloys in Fusion Reactor Structures

In design of fusion reactors, structural material selection is very crucial to improve reactor’s performance. Different types of materials have been proposed for use in fusion reactor structures. Among these materials, refractory metals and alloys having capability to withstand high temperatures and high neutron wall loads have been considered to get high power density in fusion reactors. However, these materials have insufficient technological database and are very expensive compared to steels. In addition to that, except chromium and some chromium alloys they show no low activation property. This study gives an overview of potential of refractory metals and alloys for possible use in fusion reactors.     

Investigations of BOR-60 Structural Materials and Prospects for Further Work

Since its startup in December in 1969, the BOR-60 reactor has been used effectively for irradiation of structural and fuel materials in a wide range of dose–temperature parameters. Analysis of the actual computational-experimental parameters (irradiation temperature, damage rates) shows that the irradiation conditions are highly reproducible and can be maintained accurately.The investigations made it possible to study phenomena which are important for building reactors using domestic structural materials and to choose the optimal composition and heat treatment of the materials.New directions are indicated for scientific-research work, for improving and increasing the service life of VVÉR type reactors, and for developing new-generation structural materials for fusion reactors being designed.