共查询到18条相似文献,搜索用时 640 毫秒
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模块式高温气冷堆采用预应力混凝土压力容器的可行性研究 总被引:1,自引:0,他引:1
针对模块式高温气冷堆(MHTR)钢制压力容器存在着制造难度大、运输困难和成本高等缺点.开展了对MHTR采用预应力混凝土压力容器(PCPV)的可行性研究本文简要介绍了PCPV的发展现状与技术特点,分析研究了MHTR采用PCPV作为一回路压力边界部件的技术难点及可行性,给出了.MHTR采用PCPV的初步设计方案。对该方案的分析结果表明.将PCPV应用于MHTR在技术上是可行的.不仅能够解决多腔室PCPV的力学问题以及反应堆余热释放等关键技术问题,而且能使MHTR具有更好的安全性和经济性。 相似文献
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球床模块式高温气冷堆失冷事故特性研究 总被引:2,自引:2,他引:0
利用高温气冷堆专用系统分析软件THERMIX程序,对球床模块式高温气冷堆(HTR-PM)失冷失压和失冷不失压事故的动态特性进行了研究,分析了堆芯功率、燃料最高温度及堆舱水冷壁余热载出功率等关键参数的变化过程,并对影响余热排出功率和燃料最高温度的不确定性进行了评价.研究结果表明,在失冷事故下,堆芯余热可通过热传导、辐射和自然对流等非能动方式传至最终热阱大气,燃料元件和压力容器等重要部件的最高温度均在设计限值内.这为HTR-PM保持模块式高温气冷堆固有安全性不变的同时实现单堆250 MW的功率方案奠定了基础,也为后续高温气冷堆电站示范工程进一步的深入设计研究提供了依据. 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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高温气冷堆的余热排出系统为非能动式系统,是一回路舱室冷却系统的组成部分之一。本文建立了10 MW高温气冷实验堆(HTR-10)余热排出系统在反应堆舱室内结构的三维模型,模拟HTR-10运行过程中余热排出系统的工作状况。在HTR-10上进行余热排出系统试验,获得了HTR-10在最高热功率为3 MW条件下余热排出系统的相关数据。将试验数据与模拟结果进行比对,结果表明:模拟结果与试验数据存在偏差。通过分析,提出从模型设计、工况适应性两方面对模型进行优化。 相似文献
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反应堆在停堆后相当长时间内仍具有较高的剩余发热是核电站的重要特性,也是核电站安全分析的关键。因此,对反应堆余热及其不确定性进行分析,对于合理设计余热排出系统、研究论证燃料元件在事故后的安全特性等均具有重要意义。本工作结合德国针对球床式高温气冷堆制定的余热计算标准,介绍了球床式高温气冷堆剩余发热及其不确定性的计算方法,并结合200 MWe球床模块式高温气冷堆示范工程(HTR-PM)的初步物理设计,对长期运行在满功率平衡堆芯状态下的反应堆停堆后的余热及其不确定性进行了计算分析,为进一步的事故分析提供依据。 相似文献
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一种反应堆非能动余热排出系统的方案设计 总被引:2,自引:0,他引:2
提出了一种反应堆非能动余热排出系统的方案设计。该系统利用 3个回路的自然循环 ,把事故工况下的堆芯余热排出到最终热阱。利用RETRAN0 2程序分析了这种非能动余热排出方案的可行性 ,并结合陆奥堆的参数 ,对该非能动余热排出系统方案在 1 0 0 %额定工况下的余热排出能力进行了数值模拟计算 ,还分析了影响余热排出能力的几个关键因素 相似文献
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Using the grey correlation analysis, it can be concluded that the reactor pressure vessel wall temperature has the strongest effect on the passive residual heat removal system in HTR (High Temperature gas-cooled Reactor), the chimney height takes the second place, and the influence of inlet air temperature of the chimney is the least. This conclusion is the same as that analyzed by the traditional method. According to the grey model theory, the GM(1,1) and GM(1, 3) model are built based on the inlet air temperature of chinmey, pressure vessel temperature and the chimney height. Then the effect of three factors on the heat removal power is studied in this paper. The model plays an import,ant role on data prediction, and is a new method for studying the heat removal power. The method can provide a new theoretical analysis to the passive residual heat removal system of HTR. 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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SAC-PREARS 是一个用于分析非能动RHRS稳态和瞬态安全特性的专用程序.通过实验验证的用于AC-600 非能动 RHRS安全分析的MISAP 程序,对SAC-PREARS程序进行了稳态计算验证.并应用SAC-PREARS程序对200 MW 核供热堆非能动RHRS稳态和瞬态热工水力特性进行了分析,得出了具有工程意义的结论. 相似文献
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The Ignalina nuclear power plant (NPP) is a twin-unit with two RBMK-1500, graphite moderated, boiling water, multichannel reactors. The accident management guidelines for beyond design basis accidents (BDBAs) are in a stage of preparation at Ignalina NPP. The most challenging event from BDBAs is the unavailability of water sources for heat removal from fuel channels (FCs). Due to specific design of RBMK, there are a few possibilities for heat removal from reactor core by non-regular means: depressurisation of reactor cooling system (RCS) (if pressure in cooling circuit is high) and supply of water into cooling system from low pressure water sources, removal of heat from graphite stack by reactor gas circuit, removal of heat from reactor core using cooling circuit of control and protection system channels, etc. The possibility to remove the heat using cooling circuit of control and protection system channels looks very attractive, because the channels with control rods are cooled with water supplied by the system totally independent from the reactor cooling system. The heat from fuel channels, where heat is generated, through graphite columns is transferred in radial direction to cooled channels with control rods. Therefore, the heat removal from RBMK-1500 reactor core using control rods cooling circuit can be used as non-regular mean for reactor cool-down in case of BDBAs with loss of long-term heat removal from the core. 相似文献