堆芯

HTR-10堆芯的工程化动态模型

Dynamic engineering model of the HTR-10 reactor core

堆芯冷却剂平均流速为3.2m／s，小于临界流速；

The average flowrate of reactor core coolant is 3.2 m / s , lower than the value of critical flowrate ;

紧急冷却系统有可能失灵，核反应堆有可能发生堆芯熔毁。

Emergency cooling systems could fail and a reactor meltdown could occur .

ADS系统由强流质子加速器、散裂靶件和次临界堆芯构成。

ADS is composed by intense-beam proton accelerator , spallation target and subcritical core .

用涡格子CFD方法研究反应堆堆芯元件流致振动的流体力学机理

Investigation of Flow Induced Vibration of Core Rods in Nuclear Reactor Using Vortex Lattice CFD Method

特征统计算法（CSA）在堆芯装料方案优化中的应用

Characteristic Statistic Algorithm ( CSA ) for In - Core Loading Pattern Optimization

CARR堆芯容器用辐照监督装置的设计优化

Design Optimization of Material Irradiation Monitor Device for CARR Core Vessel

PWR堆芯不同状况下安全壳内辐射水平的计算

PWR containment radiation levels calculation relate to core conditions

在堆芯内装载武器级钚，包层内装载Th，选取液态钠为冷却剂。

Weapons grade Pu was loaded into the core with Th in the blanket . Liquid sodium was used as the coolant .

CARR应急堆芯冷却系统停堆冷却措施分析

Analysis of Cooling Measures for ECCS at CARR Shutdown

研究了HTR－10堆芯的动态仿真，为反应堆控制系统的分析和设计服务。

The dynamic simulation of HTR 10 core is studied .

多重网格法及在HTR-10堆芯动态仿真

Application of multigrid method in dynamic simulation of HTR 10 core

加速器驱动次临界系统（ADS）与核能可持续发展三束流驱动ADS次临界反应堆堆芯初步设计

Study on ADS and the sustainable development of nuclear energy Preliminary Design of Sub-critical Reactor Core for Accelerator Driven System with Three Proton Beams

MCNP程序在反应堆堆芯建模中的应用

Application of MCNP in the modeling for reactor core

HTR-10进气事故下堆芯石墨腐蚀分析

Analysis of corrosion of the core graphite in case of air ingress accident

三束流驱动ADS次临界反应堆堆芯初步设计加速器驱动的次临界系统快堆次锕系核素非均匀布置堆芯的中子学研究

Preliminary Design of Sub-critical Reactor Core for Accelerator Driven System with Three Proton Beams Neutronics Study on Minor Actinide Heterogeneous Core of ADS Fast Reactor

作为加速器和次临界堆芯的耦合部件，散裂靶件对系统整体的运行安全起到了至关重要的作用，是ADS系统相关研究最主要的组成部分之一。

As the coupling component between accelerator and subcritical core , the spallation target is of crucial importance to the operation safety of the whole system .

BWR堆芯和区间功率-流量振荡计算

The Calculations of BWR Core-wide and Region Power / Flow Oscillation

文章介绍了ADS反应堆物理研究中的实验装置设计目的、堆芯布置及控制系统，同时给出了ADS反应堆物理研究中的部分内容。

The paper gives the design purpose , reactor core arrangement and control system of ADS experiment facility on reactor physics research . Some experiment work is introduced .

600MW反应堆堆芯入口流量分配实验研究

Test of flow distribution at the core inlet of 600 MW reactor

堆芯容器及堆内构件是中国先进研究堆（CARR）中的关键设备之一。

The core vessel and core structure of China Advanced Research Reactor ( CARR ) are one of the key components .

使用ANSYS软件，在模拟计算过程中，通过提取每一时刻的堆芯组件的间隙值，来改变流体附加阻尼的数值，再进行下一步计算。

Software ANSYS is applied in numerical simulation , by extracting every moment of the gap value in the simulation process to change the fluid added damping value , then calculated the next step .

与实验结果比较，计算得到的临界实验高度误差为0.6%，堆芯有效增值因子Keff误差为0.1%。

The results compared well with experimental results with errors of 0.6 % in the initial criticality height and 0.1 % in the effective multiplication factor , k__eff .

M5锆合金主要应用于核反应堆堆芯的燃料包壳和结构件。

M5 zirconium alloy is mainly applied in the nuclear reactor core as the fuel cladding tubes and structure materials .

基于物理启动、同位素生产以及空间燃耗不同考虑的全堆芯模型，采用少群扩散方法对中国先进研究堆（CARR）堆芯进行计算。

The neutronics analysis of China Advanced Research Reactor is carried out with the diffusion method by using the whole core module in this paper .

将反应堆堆芯元件简化为圆柱排列，用涡格子CFD（计算流体力学）方法对其中基本并列双圆柱元素进行数值研究。

In this paper , the vortex lattice CFD ( Computational Fluid Dynamics ) method is employed to investigate the dynamic response of core fuel and control rods on the fluid flow in nuclear reactors .

本文以核安全法规和导则为前提，以满足系统功能为基础，首先介绍了CARR应急堆芯冷却系统的功能、主要参数和流程。

This paper first introduced the function , main parameters and flow chart of the CARR ECCS , consider - ing the nuclear safety code and guide .

为确保堆芯的充分冷却和防止汽轮机叶片的损坏，将SG水位控制在合理的位置是十分重要的。

Properly controlling the water level of SG is very important in order to secure the sufficient cooling source of nuclear reactor and prevent the damage of turbine blades .

由于堆芯冷却剂密度低，因此SCWR可设计为热堆，也可设计为快堆。

The coolant density of SCWR is low , soit can be designed into a thermal reactor or fast reactor .

NHR-200堆芯旁通区三维流动传热数值分析

Numerical Analysis of Three-dimensional Convective Heat Transfer in the Core Bypass of the Nuclear Heating Reactor NHR-200