压水堆核电站一回路硼浓度监测

简要介绍了目前我国在役压水堆核电站用硼浓度监测仪的工作原理．分析了该仪器使用过程中的标定方法及提高硼浓度测量可靠性的技术措施。     

压水堆核电站发展趋势及其对策

文中简要介绍了国外压水堆核电站的发展趋势。结合我国秦山二期工程的具体情况,探讨了我们应该采取的对策。     

压水堆核电站不锈钢主管道铸造

用电弧炉和ＡＯＤ双联冶炼核电站主管道Ｚ３ＣＮ２０－０９Ｍ,并根据Ｓｈａｅｆｆｌｅｒ图计算结果调整Ｚ３ＣＮ２０－０９Ｍ的铁素体含量。在离心铸管工艺中,用加大型筒壁厚,减小挡枝内孔直径,选大的重力加速度ｇ值,增加内孔加工余量等措施铸造出主管道样件,测试结果表明,主管道样件各项性能指标均满足ＲＣＧ－Ｍ的要求。     

压水堆核电站反应堆压力容器锻件国产化探讨

高质量要求及大型化是反应堆压力容器锻件的两个关链问题。本文首先简介了反应堆压力容器对材料的特殊性能要求和锻件大型化的情况,然后分析了大型锻件的生产工艺路线及目前我国存在 的主要差距,最后介绍了提高大型锻件质量的措施。     

压水堆核电站安全壳研究

本文介绍了事故状态下安全壳压力和温度瞬态变化计算程序,重点讨论了CONTEMPT—LT/028程序,并用该程序分析了秦山核电站安全壳在在失水事故状态下的各种响应,研究了一些影响因素。     

用于压水堆负荷跟踪运行的硼浓度模糊控制系统

对压水堆负荷跟踪运行进行了研究，提示将一种三维模糊控制系统应用于硼浓度自动调节的设计方案，在设计中，将核电站全范围模拟机的数学模型移植到微机上作为控制对象，对硼浓度模糊控制系统进行仿真实验的结果证实，该模糊控制方法不仅可行，而且效果良好。     

压水堆核电站氚的辐射风险分析

通过物理化学的方法分析了水中氚的浓度、水温及空气中氚的平衡浓度的关系,并得到了三者之间的理论关系式。空气中氚的平衡浓度随水温和水中氚的浓度的升高而升高。分析表明,当水温为３０℃时,只有当水中的氚的浓度高达２８ＧＢｑ／ｍ＾３时,达到平衡后空气中氚的浓度才接近１ＤＡＣ（导出空气浓度）的水平。而由于厂房通风系统的运行,进风中湿度的存在,空气中氚的实际浓度要远低于其平衡浓度。加上压水堆核电站开放性系统水中     

Core cooling in pressurized-water reactor during water injection

In this paper,the reactor core cooling and its melt progression terminating is evaluated,and the initiation criterion for reactor cavity flooding during water injection is determined.The core cooling in pressurized-water reactor of severe accident is simulated with the thermal hydraulic and severe accident code of SCDAP/RELAP5.The results show that the core melt progression is terminated by water injection,before the core debris has formed at bottom of core,and the initiation of reactor cavity flooding is indicated by the core exit temperature.     

Single-loop nuclear power plant with a pressurized-water reactor

Trust Tsentroénergomontazh. Translated from Atomnaya Énergiya, Vol. 71, No. 5, pp. 458–460, November, 1991.     

Fatigue growth and fracture risk from postulated cracks in pressurized-water reactor piping welds

The necessity of taking quick corrective action when a crack indication is discovered in a nuclear piping weld, has led Framatome to evaluate beforehand the potential risk of such a situation by investigating postulated cracks. Considering pessimistic loading conditions to act on a postulated crack of a given shape and orientation enables the determination of the critical size of such a defect. Introduction of fatigue crack growth then yields the maximum crack size that can be tolerated, given the remaining lifetime of the unit. Additionally, detailed analysis of the scenario that leads to these results contributes to the understanding of the potential risk and helps in alleviating it. In this paper, a review of the basic principles and the application to the case of a branch connection weld are presented.     

Fluid mixing and flow distribution in the reactor circuit, measurement data base

Experimental investigations and computational fluid dynamics (CFD) calculations on coolant mixing in pressurised water reactors (PWR) have been performed within the EC project FLOMIX-R. The project aims at describing the mixing phenomena relevant for both safety analysis, particularly in steam line break and boron dilution scenarios, and mixing phenomena of interest for economical operation and the structural integrity. Measurement data from a set of mixing experiments have been gained by using advanced measurement techniques with enhanced resolution in time and space. Slug mixing tests simulating the start-up of the first main circulation pump are performed with two 1:5 scaled facilities: the Rossendorf Coolant Mixing model ROCOM and the Vattenfall test facility. Additional data on slug mixing in a VVER-1000 type reactor have been gained at a 1:5 scaled metal mock-up at EDO Gidropress. Experimental results on buoyancy driven mixing of fluids with density differences have been obtained at ROCOM and the Fortum PTS test facility.Concerning mixing phenomena of interest for operational issues and thermal fatigue, flow distribution data available from commissioning tests at PWRs and VVER are used together with the data from the ROCOM facility as a basis for the flow distribution studies.In the paper, the experiments performed are described, results of the mixing experiments are shown and discussed. Efforts on computational fluid dynamics codes validation on selected mixing tests applying Best Practice Guidelines in code validation will be reported about in a separate paper.     

Release of tritium from boron carbide irradiated with reactor neutrons

The release of tritium from irradiated boron carbide in a pure Ar atmosphere was investigated between 500 and 900°C. The sintered B₄C samples with densities between 75 and 95% of the theoretical density were irradiated with reactor neutrons with total neutron doses up to 5 × 10²⁰/cm². Effective diffusion coefficients, D_eff, were derived from the release data using the model “diffusion out of a sphere”. D_eff decreases by about 3 orders of magnitude with increasing total neutron dose, levels off at about 10¹⁸n/cm² and increases at very high doses ( > 10²⁰ n/cm²). The decrease in the tritium mobility is attributed to the radiation defects formed in the B₄C. The activation energy of 210 ± 30 kJ/mol for the tritium diffusion in the irradiated B₄C is much higher than the value found for unirradiated material. D_eff depends also very strongly on the density of the sintered material.