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为开展关于核热推进反应堆堆芯的稳态热工水力计算,基于现有针对压水堆的系统分析程序,添加了氢气的物性模型及流动换热和摩擦阻力关系式,并采用公开文献中的数据进行验证。结果表明采用上述模型计算得到的结果与参考值符合较好,二次开发的程序适用于氢气的流动换热计算。针对一种折流式核热推进反应堆堆芯,使用该系统程序建模并计算,得到了堆芯的流量、焓升等分布情况。研究结果表明,对于折流式核热推进反应堆,内外堆芯燃料元件之间的导热会增强堆芯释热不均,对堆芯的稳态热工水力特性有较大影响,堆芯物理方案的设计应结合热工水力方面的计算。本研究可为核热推进系统内氢气流动换热计算提供借鉴。 相似文献
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为开展关于核热推进反应堆堆芯的稳态热工水力计算,基于现有针对压水堆的系统分析程序,添加了氢气的物性模型及流动换热和摩擦阻力关系式,并采用公开文献中的数据进行验证。结果表明采用上述模型计算得到的结果与参考值符合较好,二次开发的程序适用于氢气的流动换热计算。针对一种折流式核热推进反应堆堆芯,使用该系统程序建模并计算,得到了堆芯的流量、焓升等分布情况。研究结果表明,对于折流式核热推进反应堆,内外堆芯燃料元件之间的导热会增强堆芯释热不均,对堆芯的稳态热工水力特性有较大影响,堆芯物理方案的设计应结合热工水力方面的计算。本研究可为核热推进系统内氢气流动换热计算提供借鉴。 相似文献
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基于RELAP5 MOD3.2的钠冷快堆热工水力系统分析程序开发及验证 总被引:1,自引:1,他引:0
对大型核反应堆热工水力分析程序RELAP5 MOD3.2进行了改造,使之适用于钠冷快堆系统安全分析。在不影响原程序功能的基础上添加了气液两相钠物性和液态金属对流换热模型,并改造了相应的初始化模块和计算模块。改造后的程序可正确模拟钠的流体力学特性和热物性,搭建钠冷快堆热工水力流体网络进行分析计算。对EBR-Ⅱ试验堆基准题进行了稳态模拟和失流事故分析,其中稳态计算主要参数与实验值相对偏差小于1%,瞬态计算相对偏差小于10%,各参数变化趋势与实验值相符良好,初步验证了改造程序的可靠性。 相似文献
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《核动力工程》2017,(5):34-39
为研究热管冷却双模式空间堆(HP-BSNR)堆芯稳态热工水力安全特性,基于改进后的双模式反应堆初步概念设计方案建立了其堆芯热工水力模型,包括推进模式和电源模式下的燃料元件单通道模型、换热模型、压降计算模型以及热管模型等,开发了堆芯稳态热工水力分析程序STHA_HPBSNR。采用文献的实验数据以及程序ELM的计算结果与程序STHA_HPBSNR的氢气物性计算模块和热力学参数计算模块进行对比,初步验证了程序STHA_HPBSNR用于双模式空间堆系统热力学稳态计算分析的可靠性。此外分析了不同换热关系式和摩擦阻力关系式对通道壁面温度的影响,为后续将STHA_HPBSNR程序应用于双模式空间堆堆芯瞬态安全分析奠定了基础。 相似文献
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将堆芯子通道热工水力分析程序COBRAⅢC/MIT-2的水物性、临界热流关系式、泡核沸腾起始点判断公式等加以修正或扩充,使之能用于低温低压下研究堆或实验堆的分析。利用改进的COBRAⅢC/MIT-2,对日本板状元件高通量研究堆JRR-3M在不同基准流速下以及不同流道阻塞率下的热工水力特性进行了分析计算,所得结果与日本原子能研究院开发的热工水力分析软件COOLOD的相应预测结果符合良好。 相似文献
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Masa-Aki Ochiai 《Nuclear Engineering and Design》1981,66(2):223-232
A computer code WTRLGD has been developed to describe the transient internal pressure of a waterlogged fuel rod during power burst and also to predict the possibility of the rod failure in the mode of cladding rupture. The code predicts transient thermal behavior of the fuel rod on the basis of an assumption of axisymmetry, and thermal-hydraulic transients of the internal water on the basis of a homogeneous volume-junction model modified so as to involve the cladding deformation. Calculated transients of the rod pressure are in fairly good agreement with those measured in the NSRR experiments, simulating the fuel rod behavior under an RIA condition. The comparison between calculation and experiment verifies that the code is an effective tool for the prediction of the failure of a waterlogged fuel rod. 相似文献
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The RALIZA-2 computer program was designed for thermal-hydraulic analysis of flow channel and fuel element of PWR/BWR at steady-state and transient conditions. A nonhomogeneous, nonequilibrium model of a two-phase flow and a two-dimensional heat conduction model of fuel pin are used in the program. A fully implicit integration scheme for both models is used. The steady-state constitutive correlations set is used. The void fraction, pre- and post-DNB heat transfer mechanism are compared with data. Also a loss of flow experiment was calculated and compared with nuclear heated rod bundle experimental data for typical PWRs. A very good agreement was obtained. 相似文献
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The generalized simple, transient, integral energy balances based on the average properties for the fuel and cladding have been used in our new multichannel thermal-hydraulic model for calculating the transient behavior of coolant in the rod bundle. This model was developed to provide a simple useful tool for analyzing the flow and thermal transients in a rod bundle with reasonable accuracy, and to understand the fundamental characteristics of flow in the rod bundle under both normal and abnormal condition of reactor-core operation. 相似文献
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《Annals of Nuclear Energy》1999,26(4):301-326
This paper examines the applicability of a mathematical dynamic model developed here for the simulation of the thermal-hydraulic transient analysis for light water reactors (LWRs). The thermal-hydraulic dynamic modeling of the fuel pin and adjacent coolant channel in LWRs is based on the moving boundary concept. The fuel pin model (FUELPIN) with moving boundaries is developed to accommodate the core thermal-hydraulic model, with detailed thermal conduction in fuel elements. Some results from transient calculations are examined for the first application of the thermal-hydraulic model and the fuel pin model with moving boundaries in a boiling water reactor (BWR). An accurate minimum departure from nucleate boiling ratio (MDNBR) and its axial MDNBR boundary versus time within the fuel channel are predicted during transients. Transient analysis using a known thermal-hydraulic code, COBRA and FUELPIN linked with a PWR systems analysis code show that the thermal margin gains more by a transient MDNBR approach than the traditional quasi-steady methodology for a pressurized water reactor (PWR). The studies of the overall nuclear reactor system show that moving boundary formulation provides an efficient and suitable tool for thermal transient analysis of LWRs. 相似文献
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事故条件及海洋条件下反应堆处于非稳态工况,堆芯燃料组件内热工水力行为具有瞬变及多因素耦合特性,对反应堆的安全提出更高挑战,因此有必要对燃料组件内瞬态特性进行研究。本文通过测量棒状燃料组件内压降和流量之间延迟时间开展棒束通道脉动流条件下相位差研究,对比了相位差在不同振幅、不同流动状态下的变化特性,并分析了定位格架对脉动流相位差的作用特点。另外,基于粒子图像测速(PIV)技术开展了脉动流条件下棒束通道内流场分布特性研究,对比了相同流量条件下稳态工况与瞬态工况下流场分布差异,分析了主流具备不同加速度时棒束通道内流场分布特征。实验结果表明:定位格架可减小脉动流下棒束通道内相位差;棒束通道内流场演化滞后于主流量变化。实验结果有助于揭示燃料组件在非稳态条件下瞬态特性,并为燃料组件的设计和优化奠定基础。 相似文献
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Amjad Nawaz Hidekazu Yoshikawa Anwar Hussain 《Journal of Nuclear Science and Technology》2016,53(11):1794-1808
During reactor operation, many complex changes occur in fuel rod which affects its thermal, mechanical and material properties. These changes also affect the reactor response to the transient and accident situations. Realistic simulation of fuel rod behavior under transients such as reactivity-initiated accident (RIA) is of great significance. In this study, thermal hydraulic analysis code THEATRe (Thermal Hydraulic Engineering Analysis Tool in Real-time) has been modified by addition of fuel rod behavior models for dynamic simulation of nuclear reactor. Transient changes in gas-gap parameters were taken into account by modeling the gas-gap behavior. Thermo-mechanical behavior of fuel rod is modeled to take into account the thermal, elastic and plastic deformation. To simulate RIA, point reactor kinetics model is also incorporated in the THEATRe code. To demonstrate the transient fuel rod behavior, AP1000 reactor is modeled and three hypothetical RIA cases are simulated. The RIA is considered at three different reactor power levels, i.e. 100, 50 and 1% of nominal power. The investigated parameters are fuel temperature, cladding stress and strain, fuel and cladding thermal conductivity and heat transfer coefficient in gas-gap. Modified code calculates the fuel rod temperatures according to updated fuel, clad and gas-gap parameters at the onset of steady-state operation and during the transient. The modified code provides lower steady-state fuel temperature as compared to the original code. Stress and strain analyses indicate that the hoop and radial strain is higher at high power locations of the fuel rod; therefore, gap closure process will initially occur in the central portion of the fuel rod and it should be given more emphasis in the safety analysis of the fuel rod and nuclear reactor during accidents and transients. 相似文献
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Robert T. Anderson 《Nuclear Engineering and Design》1982,67(3):397-405
Shipment of spent nuclear fuel from operating reactors is an important link in resolving the fuel storage and nuclear waste problems. Certain thermal problems must be considered. The nuclear spent fuel, even after a period of pool storage, has sufficient decay heat to necessitate special handling when being shipped to an off-site location. This paper presents the results of development related to the thermal interaction between dry spent fuel casks and nuclear fuel under operating situations. The tests were performed at the Barnwell Nuclear Fuel Plant (BNFP) using full-sized truck and rail casks and electrically heated dummy fuel assemblies. The safe and practical operation of the equipment developed has been shown. 相似文献
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Robert T. Anderson 《Nuclear Engineering and Design》1982,67(3):407-414
Shipment of spent nuclear fuel from operating reactors is an important link in resolving the fuel storage and nuclear waste problems. Certain thermal problems must be considered. The nuclear spent fuel, even after a period of pool storage, has sufficient decay heat to necessitate special handling when being shipped to an off-site location. This paper presents the results of development related to the thermal interaction between dry spent fuel casks and nuclear fuel under operating situations. The tests were performed at the Barnwell Nuclear Fuel Plant (BNFP) using full-sized truck and rail casks and electrically heated dummy fuel assemblies. The safe and practical operation of the equipment developed has been shown. 相似文献
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FARST, a computer code for the evaluation of fuel rod thermal and mechanical behavior under steady-state/transient conditions has been developed. The code characteristics are summarized as follows:
- 1. (i) FARST evaluates the fuel rod behavior under the transient conditions. The code analyzes thermal and mechanical phenomena within a fuel rod, taking into account the temperature change in coolant surrounding the fuel rod.
- 2. (ii) Permanent strains such as plastic, creep and swelling strains as well as thermoelastic deformations can be analyzed by using the strain increment method.
- 3. (iii) Axial force and contact pressure which act on the fuel stack and cladding are analyzed based on the stick/slip conditions.
- 4. (iv) FARST used a pellet swelling model which depends on the contact pressure between pellet and cladding, and an empirical pellet relocation model, designated as “jump relocation model”.