Status of the construction of the THTR-300 MW and the design of a 500 MW follow-on plant

The further completion of the THTR-300 MW_c prototype nuclear power plant will be performed by a consortium of the companies BBC/HRB/Nukem within the planned time schedule: The start of the nuclear commissioning began at the end of August 1983 by loading the first spherical fuel elements into the core. First critically was reached in September 1983. Handover of the plant is scheduled for October 1985 after a 10 months period of test power operation.For the HTR-500 MW_c, which is under discussion as the THTR-follow-on plant, a conceptual design analysis was performed by BBC/HRB on the basis of a private business contract placed by the Arbeitsgemeinschaft Hochtemperaturreaktor (AHR) uniting 16 power industry companies. The HTR-500 is the consequent continuation of the THTR concept. The use of HTR-specific safety characteristics as well as a further optimization of component structures and circuits result in almost the same electricity generation costs with substantially lower absolute capital costs as compared to large pressurized power reactors.     

Fuel element design for the 300 MW(e) gas-cooled fast breeder reactor

The fuel element design for a 300 MW(e) gas cooled fast breeder reactor (GCFR) is presented. The design is the result of a program sponsored by Kernforschungsanlage, Julich (KFA) to develop and fabricate a full size fuel element model under extension of an agreement between General Atomic (GA), Kraftwerk Union (KWU), and KFA to exchange information from GCFR irradiation experiments. The resulting fuel element model design was achieved by joint participation between GA and KWU and relies on the experience and knowledge of the two companies. The model, which will be manufactured by KWU using prototypical materials and specifications, except for dummy fuel pellets, will establish manufacturing feasibility and identify areas for future cost reduction improvements. The evolved designs, particularly the fuel rods, are very similar to those employed in the liquid metal fast breeder reactor (LMFBR) programs. These similarities enable the GCFR to use the vast amount of data being generated for the LMFBR programs, with only an incremental development plan needed to verify certain unique features inherent to the use of helium as the primary coolant.     

Two passive safety systems in a 200 MW demonstration nuclear heating plant

Two new passive safety systems for a demonstration nuclear heating plant, the residual heat removal system (RHRS) and the boron injection system (BIS), are introduced in this paper. Their common characteristic is that they have no driving equipment, therefore fluid circulation depends only on gravity (in BIS) or natural circulation (in RHRS). The inherent safety and realizability of both systems are illustrated in this paper.     

Modbus通信协议在300 MW核电站计算机系统中的设计与应用

采用Modbus协议实现核电站计算机系统与第三方计算机系统之间的数据通讯.详细介绍了Modbus接口协议的设计与实现.通过在协议中定义开关量质量位、浮点型模拟量、从站冗余地址等方式增强了系统数据集成的安全性与可靠性.     

Summary of NET plant probabilistic safety approach and results by means of ENEA fusion plant safety assessment (EFPSA)

This paper summarizes the probabilistic safety assessment for the main accident scenarios associated with failures originating in the In-Vessel Plant Area of the Next European Torus (NET). The assessment refers to the Basic Performance Phase of operation under normal running and conditioning. For the corresponding accident sequences, the values of the annual expected frequency and the seriousness of consequences expressed as early dose to the Most Exposed Individual (MEI) of the public are listed.     

Irradiation testing of design models for the GCFR fuel pressure equalization (vent) system

The major consideration in the design of the pressure equalization system for the gas-cooled fast breeder reactor is the release and venting of gaseous and volatile fission products. Single vented rods have been irradiated in the thermal flux of the Oak Ridge Research Reactor (ORR) at GCFR operating conditions of 12–15 kW/ft and 565–685°C cladding outside temperature to determine the fission product release and to verify the design concept. Results obtained to date from measurements of fission gas release and transport have been compared with predictions based on design assumptions to verify analytical models and have established a degree of conservatism of design assumptions.The release of radioactive gases from the fuel matrix was measured directly at 12 kW/ft in an operating fuel rod and was found to be significantly less than the design assumption of 100% instantaneous release and less than predictions using the diffusion model with Findlay's coefficients. Although solid state diffusion was found to be the dominant process delaying the venting of fission gases in the experimental irradiation, fission gas interdiffusion in helium will be the dominant venting transport process for the reactor design. Delay of fission gases by adsorption on charcoal was verified at trap operating temperatures for burn-ups up to 54 000 MWd/t. Volatile fission products (cesium and iodine) did not migrate beyond the fuel-blanket interface. The feasibility of the vented-fuel-rod design concept has been established.     

The design of 2300 MW(e) twin high temperature gas-cooled reactors for the philadelphia electric company

The basic design features of a 2300 MW(e) twin high temperature gas-cooled reactor (HTGR) power plant are described. The reactor core consists of vertical columns of hexagonal graphite fuel-moderator elements and graphite reflector blocks which are grouped into a cylindrical array and supported by a graphite core support structure. Reactivity control is accomplished by means of 146 control rods. The distribution of helium coolant flow through the core is controlled by variable orifice valves. Each of the six primary coolant loops is equipped with a helium circulator. The main steam/water section of each steam generator consists of a single helical tube bundle arranged in an annulus around the center duct. A core auxiliary cooling system is provided to furnish an independent means of removing reactor afterheat. The inherent safety characteristics and the design safety features of the large HTGR are discussed. Station arrangement, steam cycle and twin turbine generators, plant performance and control, containment and fuel handling, and environmental controls, are described.     

Plan for first phase of safety demonstration tests of the High Temperature Engineering Test Reactor (HTTR)

Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) will be conducted for the purpose of demonstrating inherent safety features of High Temperature Gas-cooled Reactors (HTGRs) as well as providing the core and plant transient data for validation of HTGR safety analysis codes. The first phase safety demonstration test items include the reactivity insertion test and the coolant flow reduction test. In the reactivity insertion test, which is the control rod withdrawal test, one pair out of 16 pairs of control rods is withdrawn, simulating a reactivity insertion event. The coolant flow reduction test consists of the partial loss of coolant flow test and the gas circulators trip test. In the partial loss of coolant flow test, primary coolant flow rate is slightly reduced by control system. In the gas circulators trip test one and two out of three gas circulators are run down, simulating coolant flow reduction events. The gas circulators trip tests, in which position of control rods are kept unchanged, are simulation tests of anticipated transients without scram (ATWS).     

Implications of the EBR-II inherent safety demonstration test

Extensive thermal-hydraulics testing at EBR-II culminated in the Inherent Safety Demonstration Test on April 3, 1986. This work may well lead to fundamental changes in the approach to the design and licensing of liquid-metal-cooled reactor (LMR) power plants. The EBR-II test program has thus far demonstrated (1) passive removal of decay heat by natural circulation, (2) passive reactor shutdown for a loss of flow without scram, and (3) passive reactor shutdown for a loss of heat sink without scram. Supporting analyses indicate that these characteristics can be incorporated into larger commercial LMRs and be used as the basis for a totally new passive control strategy. Analyses and tests are now in progress to show that LMRs with these characteristics and the passive control strategy are also inherently safe for unprotected overpower accidents.     

Seismic model test of the GCFR core and core support structure

A 15% scale model was constructed to study the dynamic structural behavior of the GCFR (gas cooled fast breeder reactor) core support structure during seismic excitation. The model contains a perforated aluminum plate with a diameter of 20 in. and 265 model core elements constructed from 7/8 in.-diameter aluminum tubes. The proper frequency and mass ratios of the core elements and the perforated plate was ensured by placing steel inserts in the tubes. The natural frequencies, mode shapes and damping factors were individually measured for each of the components and for the complete system. Harmonic and simplified seismic forcing functions were applied to study the dynamic behavior of the core and its support structure. The test results were compared with both analytical and computer code results. Applying thick plate theory, the effective elastic modulus is 27% lower than that given in the ASME code. The resonant frequencies and the mode shapes of the “combined” core and grid plate assembly were also calculated. Applying thick plate theory to the analytical method, the two lowest frequencies were determined and the comparison with the test results shows differences od 3 and 6%.     

秦山一期30万千瓦机组热功率下降分析

文章介绍了秦山一期30万kW机组第17循环热功率异常下降的现象。通过分析,确定了其原因是主给水流量孔板的测量特性偏差所致。为了确保在本循环剩余的时间内反应堆安全稳定运行,对后续时间的运行功率进行了预测,并制定了相应的管理应对措施。     

百万千瓦级压水堆核电厂停堆PSA研究

对百万千瓦级核电厂的停堆运行事故风险进行内部事件1级概率安全评价(PSA),并根据不同的停堆进程分别建立停堆PSA模型,分析经历LOI-RRA水位对电厂风险水平构成的影响。分析结果表明停堆工况下的电厂风险不可忽视,在冷停堆工况下经历LOI-RRA水位导致堆芯损坏频率明显增加。     

第三代电子动量谱仪的多参数符合测量系统研制

本文介绍了一种新研制的用于第３代电子动量谱仪的多参数符合测量系统。该系统以６个慢符合道的电荷脉冲信号（它含有位置信息）和一个快符合道中的符合时间信号作为７个参数，采用７个数据获到通道并行获取数据，经计算机系统处理，可同时得到（ｅ，２ｅ）反应相关的符合多道时间谱，电离能谱和电子动量谱。     

Materials aspects of BWR plant life extension

Considerable experience with plant equipment performance in nuclear power stations has indicated that the principal factors limiting the life of BWRs and PWRs are materials related. Specifically, for LWRs it is known that these materials issues generally include parameters related to stress corrosion cracking, corrosion fatigue, wear and radiation embrittlement. Not only do these parameters affect and limit the actual useful design life of plant components but also affect the plant's operating availability. In all these cases, the elimination or control of one or more of these critical parameters should improve the plants availability and significantly extend the useful service life.In the present paper, research performed to address the intergranular stress corrosion cracking (IGSCC) area is described. Specific emphasis is placed on Type 304 stainless steel which has suffered IGSCC in piping in the heat-affected-zone (HAZs) adjacent to the welds in the BWR primary system. Research has developed and qualified a number of techniques which address the three necessary conditions for IGSCC in BWRs: (1) sensitized microstructure, i.e., chromium depletion at the grain boundaries during welding; (2) over yield tensile stress; and (3) oxygenated (200 ppb) high temperature (288Another potential life-limiting IGSCC phenomenon for certain components, irradiation assisted stress corrosion cracking (IASCC) of stainless steel exposed to a high neutron flux, is also discussed. Unlike the IGSCC, IASCC results in intergranular cracking of annealed material at low stress. Fortunately, preliminary research has indicated that some of the techniques utilized for IGSCC control in piping as well as new controlled impurity level stainless steel alloys may reduce the future potential IASCC concern to an insignificant level.     

（e,2e）电子动量谱仪的多参数获取和处理系统

本文介绍研制成功的（ｅ，２ｅ）电子动量谱仪的多参数获取和处理系统。这个系统有几个特点：一是５个数据获取通道并行转换数据；二是在微机支持下，每个通道相当于一个多道；三是根据需要可很方便地分开组合数据获取的通道数，即可作为单通道微机多道分析器使用，又可作为一维或二维位置灵敏探测器的数据获取和处理系统使用；最后一点是系统实时获取时，可在线处理数据并在图形显示上分时显示符合时间谱及电离能谱。     

300MW压水堆核电厂堆芯反应性控制组件的设计和研究

总结了我国３００ＭＷ压水堆核电厂堆芯反应性控制组件设计的基本经验。针对控制棒的主要失效模式，讨论了关键的技术问题，对于首次使用的的硼硅酸盐玻璃可燃毒物，着重研究了抗强辐照性能，以于次级和初级中子源棒，分别阐述了重要的内压问题和有关的安全性能要求.?