共查询到16条相似文献,搜索用时 187 毫秒
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停堆后冷却问题是中国先进研究堆(CARR)重要的安全问题之一.冷却措施的实施对CARR的安全和建设投资有较重要的影响.CARR采用停堆初期的强迫循环及停堆后期全堆芯自然循环相结合的策略实现正常停堆和事故停堆后的堆芯冷却.停堆冷却的过程具体分为主泵大质量惯性飞轮惰转强迫冷却、应急堆芯冷却系统强迫冷却、自然循环功能部件动作实现全堆芯自然循环3个阶段.3个阶段既相互衔接又相互独立,每个阶段各有特点.停堆冷却策略的实施证明,CARR停堆冷却过程是可靠、有效、合理的,符合先进研究堆的发展趋势. 相似文献
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停堆后冷却问题是中国先进研究堆(CARR)重要的安全问题之一,冷却措施的实施对CARR的安全和建设投资有重要的影响。有关停堆冷却系统应严格遵循核安全法规,确保其可靠性和安全性。CARR采用停堆初期的强迫循环及停堆后期全堆芯自然循环相结合的方式,实现正常停堆和事故停堆后的堆芯冷却。 相似文献
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针对中国先进研究堆(CARR)的具体结构和运行特点,考虑冷却剂所有可能的流动状态以及换热形式,利用FORTRAN程序设计语言开发了CARR瞬态热工水力计算程序TSACC.利用程序对CARR发生全厂断电事故(SBO)时控制棒不能下落,且应急冷却泵不能投入运行这一严重事故工况进行了计算分析.计算结果表明:CARR发生SBO时,在应急冷却系统故障和控制棒不能插入堆芯的严重事故工况下,堆芯功率仍然能够在冷却剂密度反馈、空泡反馈及燃料多普勒反馈等作用下降低至较低的水平,能够保证燃料元件结构的完整性,也说明了CARR具有很高的固有安全性.计算结果同时发现:在自然循环建立过程中,堆芯冷却剂流量出现了短暂的密度波流动不稳定现象. 相似文献
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高温气冷堆紧急停堆后需要快速冷却堆芯,使其达到重新启动条件,制定合理的冷却方案对于减少电厂运行成本和保护设备安全具有重要意义。本文建立了冷却系统的数学模型,对冷却过程中关键设备的传热传质过程进行了动态数值模拟。首先分析了德国高温气冷堆采用的直接冷却方案,结果表明,此方案无法避免对设备形成冷冲击或热冲击,风险性较大。进而提出了适用于我国高温气冷堆的新方案,新方案包括4个步骤:蒸汽发生器排水-卸压-预冷-冷却堆芯。动态分析表明,新方案成功地避免了冷/热冲击,大幅提高了安全性,冷却时间也在可接受范围内。 相似文献
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CARR设置了以下18个流体系统:反应堆冷却剂系统:二次冷却水系统;重水冷却系统;第二停堆系统;氦气系统;真空系统;热水层循环系统;应急堆芯冷却系统;水冷同位素孔道冷却系统;水池充排水系统;反应堆冷却剂净化系统;反应堆池水净化系统;重水净化系统:重水浓缩系统:中放系统;低放系统;去离子水制备系统;压缩空气系统。 施工设计阶段对以下系统作了部分改进。 1)应急堆芯冷却系统 由原方案的2台应急泵应急启动改为2台随堆运行,并与池水冷却系统合并。每台泵的吸入口并联在1个母管上,从700m3的堆水池内吸水,出口并联在1个母管上,与反应堆冷却剂系统的冷段母管相连。应急泵出口旁路管并联在一起,并与池水冷却系统的板式换热器 相似文献
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根据下一代核能系统的发展目标,提出了采用自然循环的一体化小型氟盐冷却高温堆的概念。利用修改后的RELAR5-MS系统分析程序,建立了一体化小型氟盐冷却高温堆模型,并得到其稳态特性参数。在此基础上,对其在满功率运行状态下的反应性引入事故和失热阱事故进行了分析。分析计算表明,在反应性事故工况下,由于自然循环的存在,堆芯冷却剂流量随着堆芯温度发生动态变化,最终达到新的稳态,燃料棒和冷却剂温度均处于安全限值范围内。在失热阱事故下,反应堆负反馈的特性使得堆芯功率逐渐降低并实现自动停堆,即使不考虑余热排出系统的作用,燃料组件和冷却剂温度上升缓慢,在140 h内,燃料棒和冷却剂温度均处于全限值范围内。结果表明,一回路采用自然循环冷却的一体化小型氟盐冷却高温堆具有良好的固有安全性。 相似文献
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为验证和评估棱柱型模块式高温气冷堆设计的固有安全性,需针对代表性事故工况开展计算分析。目前针对棱柱型堆芯的模块式高温气冷堆尚缺少专用的事故分析程序。本研究基于通用CFD程序COMSOL针对堆芯活性区域和压力容器建立三维模型,包括燃料和冷却剂通道、石墨慢化剂、侧反射层以及压力容器;非能动余热排出系统采用对流边界条件简化模拟。采用C++编写点堆模块求解中子动力学,并通过动态链接库(DLL)与COMSOL实现耦合。首先计算了正常运行工况下的稳定状态;然后以该结果作为初始条件,选取3个典型事故瞬态工况开展了数值模拟,包括未失压丧失强迫流动冷却(PLOFC)事故、未失压丧失强迫流动冷却且未能停堆(PLOFC+ATWS)事故以及反应性引入且未能停堆(RIA+ATWS)事故;最后针对压力容器壁与非能动余热排出系统的辐射发射率开展了敏感性分析。计算结果表明:在本文分析的事故条件下,燃料最高温度均低于安全限值(1 620℃)且具有较大的裕量,因此均能保证堆芯燃料结构的完整性。对于PLOFC事故,提高非能动余热排出系统的换热能力能显著缓解事故后果,但对于ATWS类事故影响趋势则正好相反,需进一步开展综合分... 相似文献
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Young-Jong Chung Seong Wook Lee Soo Hyoung Kim Keung Koo Kim 《Nuclear Engineering and Design》2008,238(7):1681-1689
Advanced integral-type pressurized water reactor with a maximum thermal power of 65 MW is under development at the Korea Atomic Energy Research Institute (KAERI). This 65 MW integral reactor incorporates a number of innovative design features. In the case of a transient, the passive residual heat removal system (PRHRS) is designed to cool the reactor coolant system (RCS) from a normal operation condition to a hot shutdown condition by a natural circulation, and the shutdown cooling system (SCS) is designed to cool the primary system from a hot shutdown condition to a refueling condition by a forced circulation. A realistic calculation has been carried out by using the TASS/SMR code and a sensitivity analysis has been performed to evaluate a passive cooldown capability for various system conditions such as natural and forced circulation conditions for the reactor coolant system or the passive residual heat removal system, and number of active PRHRS trains. The reactor coolant system and the passive residual heat removal system adequately remove the core decay heat by a natural circulation and the 65 MW integral reactor can cool the coolant to the SCS entry condition in the primary system for all the possible operational conditions. 相似文献
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In designing nuclear power plants (NPPs), the evaluation of safety is one of the important issues. As a measure for evaluating safety, this paper proposes a methodology to examine the design process of emergency core cooling systems (ECCSs) in NPPs using Axiomatic Design (AD). This is particularly important for identifying vulnerabilities and creating solutions. Korean Advanced Power Reactor 1400 MWe (APR1400) adopted the ECCS, which was improved to meet the stronger safety regulations than that of the current Optimized Power Reactor 1000 MWe (OPR1000). To improve the performance and safety of the ECCS, the various design strategies such as independency or redundancy were implemented, and their effectiveness was confirmed by calculating core damage frequency. We suggest an alternative viewpoint of evaluating the deployment of design strategies in terms of AD methodology. AD suggests two design principles and the visualization tools for organizing design process. The important benefit of AD is that it is capable of providing suitable priorities for deploying design strategies. The reverse engineering driven by AD has been able to show that the design process of the ECCS of APR1400 was improved in comparison to that of OPR1000 from the viewpoint of the coordination of design strategies. 相似文献
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《Annals of Nuclear Energy》2005,32(3):261-279
The China advanced research reactor (CARR) being built in Beijing, China, is a multipurpose research reactor for a variety of fields. Theoretical calculation of thermal hydraulic characteristics of CARR is presented in this paper. The theoretical analysis consists of initial steady and transient accidental analyses. Point reactor neutron kinetics model with six groups of delayed neutron is adopted for the solution of reactor power. All possible flow and heat transfer conditions are considered and the corresponding optional models are supplied in the theoretical calculations. A new simple and convenient model is proposed for the resolution of the transient behaviors of main pump instead of the complicated four-quadrant model. Gear method and Adams predictor–corrector method are adopted alternately for a better solution to such ill-conditioned differential equations corresponding to detail process. The initial multi-channel analysis shows that the effects of geometrical size on flow distribution play dominant role and the effects of core power distribution may be neglected. The temperature fields of fuel elements under asymmetrical cooling condition are also obtained, which are the bases for further study on transient-induced stress analysis, etc. Accidental analyses show that the activity of emergency cooling system apparently reduces the peak temperatures of fuel and coolant, peak quality and other operation parameters. Thus it effectively ensures the safety in operation of CARR. Because of the adoption of modular programming techniques, this code is expected to be applied to accidental analysis of other types of reactors by easily modifying the corresponding function modules. Also, this code is expected to be validated against experimental data. 相似文献