共查询到19条相似文献,搜索用时 171 毫秒
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本文利用笔者导出的计算公式,对脉冲堆中子源元件进行了计算校核.最终得出脉冲堆中子源元件可以随堆运行的结论,并进行了必要的讨论。 相似文献
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中国实验快堆的安全特性 总被引:8,自引:0,他引:8
钠冷快堆因钠具有好的热物理特性而具有固有安全性,同时也因钠是活泼的碱金属,也难免会有钠的泄漏、钠火和钠水反应等工业事故.本文介绍了中国实验快堆利用钠冷快堆的固有安全性,装设了单靠自然循环和自然对流的事故余热导出系统等多项非能动安全系统及完善的能动安全系统,其安全性达到了第Ⅳ代先进核能系统的安全要求.对于大型快堆,因其保... 相似文献
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为研究海洋条件对海上浮动堆全厂断电事故后的事故进程及非能动安全系统运行特性的影响,通过建立海洋条件加速度场模型,基于RELAP5程序开发获得了适用于海上浮动堆的系统分析程序,并对程序进行了实验验证。利用所开发的程序通过建立双环路海上浮动堆及二次侧非能动余热排出系统的计算模型,开展了不同摇摆运动参数下海上浮动堆全厂断电事故的计算分析。计算结果表明,船体的横摇运动可加快全厂断电事故后浮动堆系统压力和温度的下降速度,堆芯余热能够被二次侧非能动余热排出系统有效导出;但横摇运动会造成事故后堆芯自然循环流量的显著降低,引起一回路系统和非能动余热排出系统中自然循环流量的大幅度振荡及周期性倒流。本文计算结果可为海上浮动堆非能动安全系统的设计提供参考。 相似文献
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反应堆停堆后的余热导出是反应堆的重要安全功能之一,停堆初期余热由裂变功率和衰变热构成,停堆后期余热主要取决于衰变热。本文开发了应用于钠冷快堆系统分析程序FR-Sdaso的衰变热计算模型,该模型可考虑裂变功率和功率历史的影响。通过与ANSI/ANS-5.1-2005标准和SAS4A/SASYS-1程序对比进行了模型验证。FR-Sdaso程序的计算结果与ANSI/ANS-5.1-2005标准的最大相对偏差约为0.1%,与SAS4A/SASYS-1的最大相对偏差在10-8量级,初步证明了所开发模型的正确性。最后,基于中国实验快堆的设计数据,分析了紧急停堆过程中裂变功率对衰变热的影响,结果表明,忽略裂变功率的影响导致衰变热的最大相对偏差约-7%,出现在停堆初期。因此,计算停堆初期衰变热时应考虑裂变功率的影响。 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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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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Meng Lu 《Journal of Nuclear Science and Technology》2019,56(8):731-743
Small break loss of coolant accident (SBLOCA) is one of the most important severe accidents in nuclear heating reactor. Nuclear heating reactor designed by Tsinghua University, whose primary loop is integrated layout and designed without main pump. The initial water volume in the reactor vessel is important to determine whether the reactor will be cooled or not as no safety injection system is designed for coolant makeup during the whole scenario. This paper simulates SBLOCA in nuclear heating reactor based on RELAP5. Transient behavior of relevant thermal parameters is specifically analyzed. Moreover, investigation also has been made on SBLOCA scenario based on different residual heat removal correlations and found the long-term residual heat removal capacity is decisive in determining the loss of coolant. The mathematical form of residual heat removal correlation is specifically deducted and can be widely applied to different situations. The envelope line that differentiates the region whether the core is safe or not under different maximum PRHRS capacity is also given. 相似文献
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SMART is an integral type reactor of 330 MW, which enhances its safety by adopting inherent safety design features. Thermal hydraulic characteristics of transients in heat removal by a secondary system for the SMART have been carried out by means of the TASS/SMR and MATRA codes. The primary, secondary, and passive residual heat removal systems RHRS of the SMART were modeled properly. Then, a set of transients for the whole system was investigated. The results of the analyses using the conservative initial and boundary conditions showed that the safety features of the SMART design carried out their functions well and there was a strong moderator temperature coefficient due to the soluble boron free reactor affected by the transient behavior. The natural circulation was well established in the primary and passive residual heat removal systems during the transients and was enough to ensure a stable plant shutdown condition after a reactor trip. 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