在反应堆物理设计和分析时，经常要进行微扰计算，以快速分析堆芯中子截面扰动下反应性的变化量。
Perturbation calculation will be required to rapidly determine the changes of reactivity caused by core neutron cross sections perturbation in reactor physics design and analysis.
该方法已在大亚湾核电站18个月换料项目的提高堆芯功率因子的分析论证中应用。
The DRM has been used for justification of increasing core power factor in Daya Bay NPS18-month fuel cycle project.
在反应堆1：10模型上进行了堆芯吊篮结构的空气和静水中模态分析试验，获得动态特性。
After having completed the model analysis of the pressurized water reactor (PWR) core barrel in anl: 10 model, the dynamic characteristics are obtained.
在反应堆1：10模型上进行了堆芯吊篮结构的空气和静水中模态分析试验，获得动态特性。
After having completed the model analysis of the pressurized water reactor (PWR) core barrel in an 1:10 model, the dynamic characteristics are obtained.
介绍了压水堆核电厂换料堆芯功率能力验证分析的原理和方法。
The Principle and methodology of pressurized water reactor nuclear power plant core power capability verification for reload are introduced.
分析结果表明，NHR-200含钆可燃毒物棒能很好地满足堆芯设计的要求，并且有较大的安全裕度。
It is demonstrated that NHR 200 gadolinia fuel rods can successfully fulfill the requirement of design criteria and possess high margin in safety.
它包括脉冲堆自然循环分析程序（MC-FLOW）、堆芯热工水力分析程序（MC-THAS）和脉冲堆瞬态分析程序（MC-TRAN）。
They are the natural circulation flow analysis code (MC-FLOW), the core thermal hydraulic analysis code (MC-THAS) and the pulsed reactor transient analysis code (MC-TRAN).
通过理论分析和运行结果比较了高通量工程试验堆（HFETR）80盒、60盒元件堆芯性能。
The core performance of 80 and 60 fuel assemblies of high flux engineering test reactor (HFETR) have been compared by theoretical analysis and operating results.
本报告介绍了堆芯换料设计需提交给核电站的设计文件和所用的计算机软件，并对启动物理试验实测值与设计预计值进行了比较分析。
This report represents the design files needed to submit to the plant and computer software used for reload design, and gives the comparison and analyses between the measured values from t…
堆芯出口热气混合实验用于测量和分析该混流结构的混合性能及其阻力特性。
A hot gas mixing experiment of outlet of HTR-PM reactor core was proposed to measure and analyze the actual mixing performance and resistance property of this mixing structure.
初步分析了以正三角形栅格紧密排列组成的棒形燃料组件和堆芯的优越性。简要介绍了稠密栅在中小型核动力装置中的应用。
This paper preliminarily analyzed the advantage of a reactor core with rod type element in tight triangle lattice, and briefly introduced its application in small and medium nuclear power plants.
通过与实验数据进行了基准测试，以验证和改进的反应堆芯抗震分析中使用的代码。
A benchmark test with experimental data was conducted to verify and improve the codes used for the seismic analysis of reactor cores.
分析结果表明，在失流事故初期阶段，堆芯热通道燃料中心最高温度和MDNBR不超出规定限值，满足安全准则要求。
The results show that at the early stage of the loss of flow accident, the highest fuel central temperature and MDNBR in the hot channel do not exceed specified limits and meet the safety criteria.
本文介绍了CPWR640反应堆的核设计准则、堆芯特性与主要参数，并给出了堆芯核设计的主要计算机程序、计算结果及分析。
The paper presents the nuclear design criterion, reactor core characteristics and main parameters, and shows some computer codes, calculation results and analyses.
大亚湾核电站由年换料改为18个月换料，燃料组件由AF A2g改为afa3g，堆芯中子学参数发生了较大变化。因此，需要对许多事故重新进行分析。
The reload strategy is from 12 months to 18 month of GNPS, the fuel assembly changed from AFA 2g to AFA 3g and core nuclear data been changed, thus many accidents need to be reanalyzed.
大亚湾核电站由年换料改为18个月换料，燃料组件由AF A2g改为afa3g，堆芯中子学参数发生了较大变化。因此，需要对许多事故重新进行分析。
The reload strategy is from 12 months to 18 month of GNPS, the fuel assembly changed from AFA 2g to AFA 3g and core nuclear data been changed, thus many accidents need to be reanalyzed.
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