共查询到18条相似文献,搜索用时 790 毫秒
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研究了池式快堆自然循环模拟实验的模拟准则,根据模拟准则和自然循环守衡方程式,对池式实验快堆自然循环模拟实验装置,在各种模拟准则条件下的几何与热工设计参数进行了计算。研究了模型比例,事故冷却器一次侧进出口温差和阻力系数等对相似准则数的影响,并且确定了模拟实验装置的设计参数范围,从理论上解决了池式实验快堆自然循环模拟实验装置的模拟问题。 相似文献
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多孔介质方法在池式快堆系统分析软件SAC-CFR三维钠池计算模型中的应用 总被引:1,自引:1,他引:0
为准确分析池式快堆热钠池内的热工水力学特性,在已开发出的用于池式快堆系统分析的钠池三维计算模型的基础上,应用多孔介质方法建立钠池内中间热交换器、主泵、事故热交换器及屏蔽柱模型,完成了含有多孔介质模型和复杂几何边界的钠池三维计算模型开发。将该模型嵌入池式快堆系统分析软件SAC-CFR后,分析了中国实验快堆在稳态运行和紧急停堆工况下钠池内的流场分布,初步证明了所采用的多孔介质模型的合理性,为下一步非能动余热排出系统模型的开发做准备。 相似文献
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快堆(Fast Reactor)具有燃料利用率高、可嬗变核废料的优势,是目前较为理想的先进堆型之一。快堆广泛采用池式回路布置,因此对池式快堆(Pool-type Fast Reactor)进行安全分析具有重要意义。本文采用集总参数法建立池式快堆的一回路模型,基于MATLAB编写核热耦合程序并对其进行无保护失流事故工况的安全分析,并将计算结果与实验值及其他机构计算结果进行了对比。结果显示:集总参数法的计算结果与实验和其他机构计算值均符合较好,验证了程序的可靠性。使用该程序可对池式钠冷快堆在无保护失流事故中的堆芯行为与固有安全性做出较为准确的预估计算。 相似文献
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针对池式钠冷快堆特点,建立了三维系统分析模型,并结合热分层现象演化机制,提出了准确模拟热分层的关键处理方法,包括能量源项处理、三维动量方程对流项处理及三维空间进口效应处理。在此基础上,采用KALIMER及MONJU热分层实验对所开发的三维系统分析模型进行验证。结果表明模型有效解决了池式钠冷快堆三维热工水力分析的难题,实现了对钠池内温度场瞬态变化及热分层现象演化进程的快速准确模拟,同时也能够确定热分层过程中池式结构表面热应力最大位置,为池式快堆安全设计提供参考。 相似文献
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基于临界/次临界点堆中子动力学模型、燃料棒传热模型、热交换器和多孔介质等辅助热工水力模型,采用显式迭代和动态链接库技术(DLL),利用商用计算流体力学(CFD)程序FLUENT的用户自定义函数(UDF)实现中子动力学、燃料棒热传导等和快堆堆池冷却剂流动换热的耦合计算,开发池式快堆多物理耦合计算程序CFD/PF。采用CFD/PF开展小型自然循环铅铋快堆SNCLFR-10无保护超功率事故(UTOP)模拟,并与国际知名快堆多物理耦合分析程序SIMMR-Ⅲ的计算结果开展Code-to-Code对比分析。研究结果表明:CFD/PF与SIMMER-Ⅲ的分析结果吻合良好,耦合程序的开发取得了初步成功,可用于分析池式快堆堆池内的复杂三维流动和换热现象。 相似文献
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池式快堆系统瞬态分析软件开发 总被引:3,自引:3,他引:0
为实现快堆系统分析软件国产化,在已开发的适用于稳态计算的池式快堆系统分析软件SAC-CFR的基础上,进一步开发了系统各部件的瞬态模型、控制系统和保护系统模型、瞬态工况热工水力学的求解逻辑,完成瞬态计算功能的开发。通过对日本文殊快堆45%功率汽机跳闸工况进行建模分析,验证了SAC-CFR用于系统瞬态分析的有效性,为进一步开发非能动余热排出系统分析模型打下了基础。 相似文献
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事故余热排出系统是池式钠冷快堆最重要的专设安全设施之一,是实现反应堆相关事故工况下余热排出安全功能的主要手段,如全厂断电工况,而独立热交换器是快堆事故余热排出系统的关键设备之一。本文以ANSYS FLUENT为工具,对中国实验快堆现有的独立热交换器和一种改进的新型独立热交换器布置在快堆热池中的情况进行了瞬态数值模拟,并分析比较其结果,证明了改进型独立热交换器在热工水力上的可行性。本文工作对大型快堆的独立热交换器的设计具有一定的借鉴意义。 相似文献
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REN Li-xia 《中国原子能科学研究院年报》2005,(1):6-6
CEFR is a sodium cooled fast reactor of pool-type with 65 MW thermal power and matched with a 25 MW turbine-generator. As a design basis accident of CEFR, a primary pump shaft seizure is a typical one of loss of coolant accidents. The paper gives the analysis of the accident with the liquid metal fast reactor system code DINROS introduced from Russian. The results shows that the consequence of the accident is in accordance to the acceptance criteria. There is no fuel pin failure and no radioactivity release. 相似文献
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WANG Fenglong YANG Yong HUANG Shuming ZHANG Qiang WANG Shixi WU Mingyu XU Zhilong SHAO Jing WAN Haixia 《原子能科学技术》1959,54(10):1849-1857
To deal with the problem that the traditional light water reactor accidental source term calculation method is not suitable for sodium-cooled fast reactor, calculation methods for accidental source term of pool-type sodium-cooled fast reactor, including core damage type, leak type and sodium fire type, were studied and derived on basis of the analysis of release path of potential design basis accidents and beyond design basis accidents. The methods were applied to six typical accidents of the demonstration fast reactor, including the total instantaneous blockage of one fuel assembly, the leakage of cover gas region of reactor main vessel, the damage of primary argon decay tank, the leakage of main vessel, the leakage of sodium purification pipeline without protective sleeve outside the primary circuit, and the leakage of external auxiliary pipeline without protective sleeve outside the primary circuit or the isolation valve tube not be closed. The calculation of accidental source terms and their radiological consequences were carried out. The results show that the radioactive dose consequences of the six accidents are lower than the requirements of GB 6249-2011. The methods proposed can provide reference to the calculations of accidental source term of loop-type sodium-cooled fast reactor, lead-cooled fast reactor and gas-cooled fast reactor. 相似文献
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针对传统轻水堆事故源项计算方法不适用池式钠冷快堆的问题,分析可能发生的设计基准事故和超设计基准事故的释放路径,研究建立适用于池式钠冷快堆的堆芯损伤类、泄漏类和钠火类事故源项计算方法。结合示范快堆的6种典型事故:1盒燃料组件瞬时全部堵塞事故、反应堆堆本体覆盖气体边界泄漏事故、一次氩气衰变罐破损事故、主容器泄漏事故、一回路外无保护套管的钠净化管道泄漏事故和一回路无保护套管的外辅助管断裂或泄漏合并隔离阀关不住事故,开展事故源项计算及其剂量后果评价。结果表明:6种事故的放射性后果均低于GB 6249-2011的要求。该方法还可为回路式钠冷快堆、铅铋快堆以及气冷快堆事故源项计算提供参考。 相似文献
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铅铋快堆属于第四代反应堆,其一回路采用液态铅铋合金冷却.铅铋快堆一回路充排系统可以调节反应堆主容器内液态金属液位,该系统充满含有放射性物质的液态金属,其可靠性水平对反应堆运行及安全有重要影响.本文以中国科学院核能安全技术研究所·FDS团队自主设计的铅铋快堆一回路充排系统为研究对象,运用故障树分析方法对该系统进行可靠性分... 相似文献
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池式快堆系统分析软件稳态功能开发 总被引:5,自引:5,他引:0
针对目前我国快堆系统分析软件主要采用国外引进方式而导致难以掌握核心物理模型的现状,以中国实验快堆(CEFR)为研究和建模对象,基于中子动力学模型、堆芯及其热钠池模型、中间热交换器模型、一回路和中间回路热量传输系统模型、三回路模型等,自主开发了基于CompaqVisualFortran(CVF)的适用于稳态计算的池式快堆系统分析软件SAC-CFR。通过与中国实验快堆安全分析报告中数据进行对比,验证了所开发模型的精度,为下一步瞬态模型的开发及控制和保护系统的开发做准备。 相似文献
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4S (Super-Safe, Small and Simple) is a small sized sodium-cooled fast reactor being developed for the electricity supply in remote areas, high-temperature steam supply more than 400 °C, seawater desalination, and hydrogen production. The system design of power output of 10 MWe (30 MWt) has been completed. The main feature is that it does not have to be refueled for a long period (i.e. 30 years for 10 MWe version), and enable the reactor closure sealed during plant operation. Furthermore, the small size of the reactor makes the reactor building suitable for below grade installing. These two features can provide resolutions for the issues relevant to safety, security, and safeguard, which become much more serious matter internationally these days.4S is a pool-type reactor which contains the whole primary cooling system in a vessel. For the purpose of reducing the maintenance requirements with the reactor, (1) reflectors to compensate for fuel burn-up instead of control rods, (2) electromagnetic pump (EMP) which has no rotating parts, and (3) residual heat removal system by natural circulation and natural air draft are adopted. Therefore, exchange of the reactor components is not required during plant operation, in addition to no needs for refueling.Toshiba has initiated the U.S. Nuclear Regulatory Commission (NRC) pre-application review of 10 MWe version for the purpose of applying for design approval (DA). A series of public meetings with NRC has been held four times, and five technical reports have been submitted to NRC in preparation for DA application. Topics discussed in these meetings included, plant design, metallic fuel, safety design philosophy, safety analysis, measures against severe accident, phenomena identification and ranking table (PIRT), etc. Some useful comments and questions on the issues regarding the specific feature of 4S as well as sodium-cooled fast reactor were raised by NRC at the public meetings. Among them, those items which are applicable to general sodium-cooled fast reactors, e.g. principal design criteria, guideline for safety analysis, validation and verification for safety analysis code, quality requirements, severe accident, and emergency planning are presented in this paper. 相似文献