聚变发电反应堆双冷锂铅包层热工水力学设计与分析

依据结构设计和中子学计算结果给出了聚变发电反应堆FDS-Ⅱ双冷锂铅(DLL)包层热工水力学设计方案。采用数值计算软件对液态金属增殖区LiPb流场和第一壁热-结构等进行了模拟,并对功率转换系统的热效率进行了计算。通过分析评估,证实该包层热工水力学方案能较好地实现FDS-Ⅱ聚变发电反应堆预期目标。     

ATHLET 程序及其在低温供热堆中的应用

ATHLET 程序是德国核设施安全评审中心GRS开发的一个重要的反应堆热工水力学分析程序.它可分析除堆芯熔化事故之外的其他设计基准事故,可应用于美欧传统和先进的压水堆和沸水堆,原苏联设计的压水堆(VVER)和石墨水冷堆(RBMK),以及加拿大的CANDU堆等反应堆系统的热工分析.介绍了微机版ATHLET程序,并用它对5 MW低温供热堆自然循环稳态运行工况进行了计算,结果和试验数据符合很好.     

Analyses of Power Excursion Event in Chernobyl Accident with RETRAN Code

Nickel ferrite is one of the important corrosion products of PHWR's where Monel and carbon steel are used as the constructional materials in the primary heat transport system. The dissolution of synthetically prepared nickel ferrite was studied in low concentrations (<10 mmol/l) of HEDTA, DTPA, NTA and HIDA. The dependence of the dissolution rate on the ligand concentration was found to be langmuirian in all these cases. The effect of the addition of low concentrations of citric acid, oxalic acid, ascorbic acid and Fe(II)-ligand complex, individually, to each of the above chelating agents was also studied. The effect of pH and temperature on the dissolution rate was determined.     

铅基快堆关键热工水力问题研究综述

铅基快堆是一种极具发展潜力的第4代核能系统,在燃料增殖和嬗变方面具有独特优势,具有良好的非能动安全特性和经济性,且有利于实现小型化,是目前国际核能领域研究的热点。本文总结了国内外主要铅基堆型,指出了小型化是铅基快堆的发展方向,同时也指出了当前铅基快堆发展所面临的主要问题。针对热工水力关键问题的5个方面,即液态铅/铅铋流动换热特性研究、堆芯/组件热工水力分析、铅池内流动换热现象研究、系统热工水力安全分析以及特殊现象的热工水力分析,对国内外研究现状展开了分析,总结了当前研究成果,并分析了研究的发展趋势以及遇到的技术瓶颈。本文可为铅基快堆的设计和热工水力分析提供一定的建议和指导。     

环形燃料热工水力性能分析程序开发及验证

本工作开发了环形燃料子通道分析程序SAAF。采用SAAF计算了西屋公司四环路压水堆所用环形燃料组件的热工水力性能,并与VIPRE-01的计算结果进行比较。结果表明,SAAF与VIPRE-01的计算结果符合较好,SAAF可用于环形燃料热工水力设计分析。     

压水堆重反射层堆芯核热耦合高精度计算分析研究

重反射层的应用可提高反应堆中子经济性,其结构和中子吸收特性均与压水堆常规围板/反射层差异较大,因此对核设计程序的计算分析能力提出了新的要求。为分析重反射层建模方案对堆芯中子学计算结果的影响,使用先进中子学程序SCAP N和确定论堆芯高保真模拟程序NECP X对压水堆重反射层问题进行了高保真模拟,分析了5种反射层建模方案下计算结果的差异,并将高精度计算结果与商用核设计程序系统进行了对比。数值结果表明,重反射层水洞内冷却剂温度变化对计算结果影响较小;相较精确建模方案,重反射层铁水打混建模方案造成的反应性计算偏差在±30 pcm以内、组件相对功率分布计算偏差在±2%以内。     

压水堆环形燃料结构热工水力分析方法研究

以秦山二期压水堆为参考堆型,对压水堆环形燃料结构进行热工水力分析方法研究。应用SAAF程序分析了从11×11到15×15等5种不同排列方式中不同尺寸的环形燃料棒的热工水力性能,综合最小偏离泡核沸腾比、压降和燃料芯块温度等参数确定了环形燃料组件最佳排列方式为13×13。本文研究结果为相关专业分析提供了初始计算模型。     

Survey of experiments and code development for the passive residual heat removal system of PWR in China

Three different kinds of experiments and their typical results are surveyed for the passive residual heat removal system (PRHRS) of PWR performed in Nuclear Power Institute of China (NPIC) recent ten years. The typical results of MISAP. a special code for PWR passive residual heat removal system developed and assessed by NPIC,are also described briefly in this paper.     

倾斜与摇摆条件下一体化反应堆自然循环特性研究

通过建立海洋条件下的附加力模型与控制体空间坐标求解模型,开发了基于RELAP5/MOD3.1程序的海洋条件热工水力分析程序。研究了海洋条件下一体化反应堆IP200的自然循环特性,分析倾斜与摇摆条件对自然循环的影响。计算结果表明:倾斜会降低堆芯流量,导致左右侧环路冷却剂流量不一致,影响直流蒸汽发生器的换热特性;摇摆情形下,环路的附加压降主要由切向力贡献;摇摆轴偏离中心位置以及倾斜和摇摆的叠加运动均会打破环路间的热工水力对称性,增大堆芯流量的波动幅度。     

Thermal-Hydraulics in Uncovered Core of Light Water Reactor in Severe Core Damage Accident, (III)

The SEFDAN is a computer program to analyze the one-dimensional thermal-hydraulics of a partially uncovered core of a light water reactor in a severe degraded-cooling event. In order to verify the code and to obtain better understanding of the severe core damage process, SEFDAN has been applied to analyses of the thermal response of fuel rods in the Power Burst Facility Severe Fuel Damage 1-1 Test.

The calculated results are in good agreement with the experimental results. The analysis indicates that fuel cladding temperature in a portion of the lowest one third of the test bundle would have reached the melting point of the ZrO₂ during a rapid temperature excursion driven by the zirconium-water reaction. The result is consistent with the result of metallographic examination. The crucibilization effect of the ZrO₂ layer played an important role in the reaction. Steam starvation condition would have occurred in contrast to the situation of the Scoping Test of the same test series. Zirconium-water reaction on the inner surface of the fuel cladding was found to have made a strong effect on the fuel rod temperature in the upper part of the test bundle.