共查询到20条相似文献,搜索用时 140 毫秒
1.
2.
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
3.
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
4.
中国科学院上海应用物理研究所设计的10 MW固态燃料钍基熔盐实验堆(Thorium-based Molten Salt experimental Reactor with Solid Fuel,TMSR-SF)是一种球床型氟盐冷却高温堆,其三维球床随机堆积结构对于堆芯反应性变化和换料分区方案以及热工水力分析中的流动传热都具有重要的影响。为了研究TMSR-SF堆芯球床随机堆积规律,基于PFC3D(Particle Flow Code 3 Dimensional)程序,分别计算并分析了填充颗粒球大小、颗粒间摩擦系数以及填充颗粒球数对球床稳态堆积结构的影响。结果表明:填充直径小的颗粒球,可使球床孔隙率振荡减弱,有利于堆芯功率密度分布的展平;通过减小颗粒间摩擦系数能较好地模拟振动压实效果,压实过程中颗粒间摩擦系数与稳态堆积球床孔隙率和配位数满足负指数关系。当填充颗粒球数超过约8 000个后,球床平均孔隙率和配位数变化趋于恒定,分别稳定在0.43和5.6左右;此时颗粒球数量对球床稳态堆积结构的影响降至最低,可被用来代替满装堆时的球床用于TMSR-SF自然循环的热工水力数值分析,达到节省计算资源的目的。本研究对理解TMSR-SF堆芯球床结构特性与堆积规律具有参考价值。 相似文献
5.
《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
6.
7.
更准确地模拟球床式高温气冷堆堆芯温度分布,是反应堆安全分析尤其是超高温运行研究中的关键问题之一。由于堆芯球流运动具有不确定性,石墨块和碳砖等结构材料采用散体布置,堆内冷却剂流道复杂,对热工水力准确模拟造成困难,可进一步优化。本文结合HTR 10的结构特点和流道特征,简要分析了堆芯传热过程,说明了在热工模拟中准确划分结构和流道对获取更精确的堆芯温度分布的重要意义。详细梳理了冷却剂流动路径,改进了在THERMIX程序下建立的HTR 10原有热工分析模型,更合理地模拟了堆芯冷却剂漏流行为,使得模型对堆芯冷却剂流动和传热过程的描述更准确。与试验数据对比,改进后的模型对堆芯外围系统的温度分布模拟准确性显著提升。计算结果表明,反应堆在额定设计工况下满功率稳态运行时,燃料和反射层最高温度均未超过材料的耐热限值。 相似文献
8.
核热泉(NHS)堆是一种新型熔盐球床概念设计堆,其冷却剂径向流过堆芯,具有满功率自然循环特性。基于多孔介质局部非热平衡模型,利用计算流体力学(CFD)通用软件Fluent计算核热泉堆径向流堆芯的热工水力特性,并比较了不同的内、外孔板开孔率的影响。结果表明,内孔板开孔率对冷却剂流量分布影响较大;燃料中心温度具有相当的安全裕量,冷却剂横向流过堆芯的阻力远低于浮升力,能够实现全回路的自然循环。 相似文献
9.
由于环型球床高温气冷堆特殊的堆芯结构,使其在失冷失压事故下堆内最高温度能够明显低于模块式球床高温气冷堆在相同事故下堆内最高温度。当堆芯热功率有较大幅度提高时,环型堆芯仍然能够凭借自身传热机能将衰变热量及时排出,满足失冷失压事故下燃料最高温度限制。这不仅增大了反应堆的安全性能,同时也能够有效地增加反应堆单堆功率,使环型球床高温气冷堆在经济上更具竞争力。本文研究环型球床高温气冷堆在提高功率水平时,反应堆在失冷失压事故下堆内的热工特性,并综合分析了几个重要的结构尺寸热工参数对失冷失压事故下燃料最高温度的影响。 相似文献
10.
11.
12.
Jae Jun Jeong Seung Wook Lee Jin Young Cho Bub Dong Chung Gyu-Cheon Lee 《Annals of Nuclear Energy》2010
A coupled system thermal-hydraulics (T-H) and three-dimensional reactor kinetics code, MARS/MASTER, was developed to attain more accurate predictions for nuclear system transients that involve strong interactions between neutronic and T-H phenomena. In this paper, a 12-finger control element assembly (CEA) drop event in a two-loop pressurized water reactor (PWR) plant under a full power condition was analyzed, where the 12-finger CEA that is nearest to the hot leg of Loop 2 is assumed to incidentally drop. This instantaneously results in an asymmetric radial power distribution and, in turn, asymmetric loop behavior, which may lead to a reactor trip due to a low departure from nucleate boiling (DNB) ratio at the intact side of the core or an excessive difference between the cold leg coolant temperatures. This event clearly requires a coupled calculation of system T-H and three-dimensional reactor kinetics to realistically investigate the thermal-hydraulic behavior of the reactor core. A simple theoretical modeling is also devised to evaluate the cold leg temperature difference under a quasi-steady state. 相似文献
13.
14.
15.
Abderrahim Hammoud Brahim Meftah Mohammed Azzoune Lila Radji Boumazza Zouhire Mataoui Amina 《Journal of Nuclear Science and Technology》2013,50(9):1154-1160
Nuclear safety analysis remains of crucial importance for both the design and the operation of nuclear reactors. Safety analysis usually entails the simulation of several selected postulated accidents, which can be divided into two main categories, namely reactivity insertion accident (RIA) and loss of flow accident (LOFA). In this paper, thermal-hydraulic simulations of fast LOFA accident were carried out on the new core configuration of the material test research reactor NUR. For this purpose, the nuclear reactor analysis PARET code was used to determine the reactor performance by calculating the reactor power, the reactivity and the temperatures of different components (fuel, clad and coolant) as a function of time. It was observed that during the transient the maximum clad temperature remained well below the critical temperature limit of 110 °C, and the maximum coolant temperature did not exceed the onset of nucleate boiling point of 120 °C. It is concluded that the reactor can be operated at full power level with sufficient safety margins with regard to such kind of transients. 相似文献
16.
破口事故是压水堆最为关注的一类重要事故,其失水量与事故后果严重程度密切相关。NHR-200Ⅱ是由清华大学核能与新能源技术研究院经过多年研究和不断改进,设计的一种全功率自然循环低温供热反应堆,其设计中采用了多种先进的非能动和固有安全设计。本研究针对NHR-200Ⅱ反应堆,选取后果最为严重的控制棒引水管断裂且无法隔离事故,利用系统热工瞬态分析程序对事故过程进行了模拟和分析。结果表明,即使在最严重的破口失水事故下,NHR-200Ⅱ主回路中剩余的冷却剂始终能覆盖反应堆堆芯,并有效通过非能动余热载出系统带走堆芯热量,从而保证反应堆堆芯不会因裸露造成烧毁,这表明NHR-200Ⅱ具有很好的安全特性。 相似文献
17.
大破口失水事故时冷热段同时安注反应堆堆芯会更安全 总被引:1,自引:0,他引:1
大破口失水事故时,安注系统由冷段注入的大量冷却剂从压力壳和吊兰之间的环形通道经破口流入安全壳,只有少量的冷却剂流入堆芯。如果把安注系统同时安装在冷段和热段同时进行安注,从热段注入的冷却剂带走了上腔室和堆芯内的较多热量而降低了上腔室内的压力,使冷段注入的冷却剂较容易流入堆芯。同时,从热段注入的部分冷却剂在上腔室内撞击在导向管上后,沿着导向管流入堆芯,堆芯得到的冷却剂比单一冷段安注时得到的冷却剂要多,堆芯会更安全 相似文献
18.
Yeon-Sik Kim Ki-Yong ChoiHyeon-Sik Park Seok ChoBok-Deug Kim Nam-Hyeon ChoiWon-Pil Baek 《Annals of Nuclear Energy》2008
KAERI recently constructed a new thermal-hydraulic integral test facility for advanced pressurized water reactors (PWRs) – ATLAS. The ATLAS facility has the following characteristics: (a) 1/2-height&length, 1/288-volume, and full pressure simulation of APR1400, (b) maintaining a geometrical similarity with APR1400 including 2(hot legs) × 4(cold legs) reactor coolant loops, direct vessel injection (DVI) of emergency core cooling water, integrated annular downcomer, etc., (c) incorporation of specific design characteristics of OPR1000 such as cold leg injection and low-pressure safety injection pumps, (d) maximum 10% of the scaled nominal core power. The ATLAS will mainly be used to simulate various accident and transient scenarios for evolutionary PWRs, OPR1000 and APR1400: the simulation capability of broad scenarios including the reflood phase of a large-break loss-of-coolant accident (LOCA), small-break LOCA scenarios including DVI line breaks, a steam generator tube rupture, a main steam line break, a feed line break, a mid-loop operation, etc. The ATLAS is now in operation after an extensive series of commissioning tests in 2006. 相似文献
19.
20.
The High Performance Light Water Reactor is a Generation IV light water reactor concept, operated at a supercritical pressure of 25 MPa with a core outlet temperature of 500 °C. A thermal core design for this reactor has been worked out by a consortium of Euratom member states within the 6th European Framework Program. Aiming at peak cladding temperatures of less than 630 °C, including uncertainties and allowances for operation, the coolant is heated up in three steps with intermediate coolant mixing to eliminate hot streaks. Different from conventional reactors, the radial power profile is intended to be non-uniform, with the highest power in the first heat-up step in the core center and the lowest power in the second superheater step to result in the same peak cladding temperatures in each region. The concept has been studied with neutronic, thermal-hydraulic and structural analyses to assess its feasibility. Coupled neutronic/thermal-hydraulic analyses are defining the initial distribution of enrichment, control rod positions and the use of burnable poisons. Sub-channel analyses predict the coolant mixing inside assemblies, and a porous media approach simulates the flow of moderator water between assembly boxes. Finally, structural analyses of the assembly boxes are needed to minimize deformations during operation. Even though the core design cannot yet considered to be final, this state of the art review shall summarize the progress achieved so far and outline the remaining challenges. 相似文献