铅基研究实验堆无保护瞬态安全特性分析

发布时间：2018-02-01 04:00

本文关键词： 铅基反应堆无保护瞬态固有安全特性蒸汽迁移　出处：《中国科学技术大学》2017年博士论文　论文类型：学位论文

【摘要】：革新型核能系统是未来核能之路。铅基反应堆是最具潜力堆型之一。不少国家、机构均制定了庞大的铅基堆发展规划,中国在最近几年来开始加大对铅基反应堆的投入,作为国内铅基反应堆的主要推动者,中科院核能安全技术研究所FDS团队目前已完成中国铅基研究实验堆详细方案设计。反应堆无保护瞬态分析是评估其固有安全特性的有效手段,也是检验、指导其设计的重要途径。鉴于此,本文采用NTC-2D对上述铅基研究实验堆开展了系统、深入的无保护瞬态安全特性分析。本文分析工具是NTC-2D程序,该程序是由FDS团队自主开发的中子-热工耦合安全分析程序,已开展大量验证工作。在此基础上,本文开展了基于"PDS-XADS安全分析例题"的程序对比验证,以验证其整体性功能;同时开展了基于"JAEA水注入铅铋实验"的实验对比验证,以验证其"铅铋与水相互作用"模型。对比发现,NTC-2D与成熟商业软件及实验结果均吻合较好。首先,根据国内外无保护瞬态安全分析的经验,同时考虑该铅基研究实验堆自身设计特点,最终确定了本文无保护瞬态安全分析的具体内容。即:无保护超功率瞬态、无保护失流瞬态、无保护失热阱瞬态、无保护失流合并失热阱瞬态、无保护主换热器破口瞬态。基于NTC-2D对以上五种工况进行瞬态模拟,分析各种工况下的具体瞬态响应过程及机理,最终完成对其无保护瞬态安全特性的分析,具体分析结果如下:前四种瞬态工况,该反应堆展现出良好的固有安全特性,主要体现在以下三个方面:(1)该反应堆具有良好的负反馈特性:超功率工况下它阻止了功率的持续上升,在失流、失热阱及失流合并失热阱工况下,它自动降低反应堆功率,甚至停堆以保障反应堆安全;(2)主回路热容量很大,热惰性强,大大减缓了失热阱,失流合并失热阱工况下堆芯温度增长速率,大大缓解了事故后果;(3)主回路拥有良好的自然循环特性:反应堆在失流工况下依靠自然循环自动达到稳定状态,在失流合并失热阱工况下,反应堆仅靠自然循环进行堆芯冷却,大大缓解了事故后果。主换热器破口模拟揭示了 "铅铋与水相互作用"的相关现象与机理,重点揭示了水蒸汽在主容器内的迁移规律:(1)下封头断裂工况下,部分水蒸汽向下迁移,发现水蒸汽进入堆芯;(2)上封头断裂工况下,几乎所有水蒸汽向上迁移,未发现水蒸汽进入堆芯。本文研究意义:(1)可以为该铅基研究实验堆安全评价工作(初步安全分析报告撰写等)提供一定借鉴意义;(2)可以为该铅基研究实验堆设计优化提供初步依据,进而实现设计的"built in"而不是"added on";(3)可以为未来商业铅基反应堆发展提供一定参考价值。
[Abstract]:The innovative nuclear energy system is the road of nuclear energy in the future. Lead based reactor is one of the most potential reactor types. Many countries and institutions have formulated a huge lead based reactor development plan. In recent years, China has begun to increase its investment in lead-based reactors as the main promoter of domestic lead-based reactors. At present, the FDS team of the Institute of Nuclear Safety Technology of the Chinese Academy of Sciences has completed the detailed scheme design of the lead based experimental reactor in China. Unprotected transient analysis of the reactor is an effective means to evaluate its inherent safety characteristics and is also a test. In view of this, NTC-2D is used to carry out the system of the lead-based experimental reactor mentioned above. The NTC-2D program, which is developed by the FDS team, is a neutron thermal coupled safety analysis program. A lot of verification work has been carried out. On this basis, the program comparison and verification based on "PDS-XADS Security Analysis example" is carried out to verify its integrity function. At the same time, based on the "JAEA water injection lead bismuth experiment" experimental verification, to verify its "lead-bismuth and water interaction" model. NTC-2D is in good agreement with commercial software and experimental results. Firstly, based on the experience of unprotected transient safety analysis at home and abroad, the design characteristics of the lead based research reactor are considered. Finally, the specific content of the unprotected transient safety analysis is determined, that is, the unprotected superpower transient, the unprotected out-of-current transient, the unprotected heat-sink transient, the unprotected de-current combined with the heat-trap transient. Based on the transient simulation of the above five working conditions, the transient response process and mechanism of the unprotected main heat exchanger are analyzed based on NTC-2D. The results are as follows: the first four transient conditions, the reactor shows good inherent safety characteristics. The reactor has good negative feedback characteristics in the following three aspects: it prevents the power from rising continuously under the condition of overpower, under the condition of loss of current, heat trap and combined loss of current and heat trap. It automatically reduces the power of the reactor and even stops the reactor to ensure the safety of the reactor; 2) the thermal capacity of the main loop is very large and the thermal inertia is strong, which greatly slows down the loss of heat trap, and the temperature growth rate of the core under the condition of the lost current combined with the lost heat trap greatly alleviates the consequence of the accident; (3) the main loop has good natural circulation characteristics: the reactor can be stabilized automatically by natural circulation under the condition of loss of flow, and the core is cooled only by natural circulation under the condition of combined loss of flow and heat trap. The accident consequence was greatly alleviated. The main heat exchanger break simulation revealed the related phenomena and mechanism of the "lead bismuth and water interaction". The migration rule of water vapor in the main vessel is revealed emphatically. Under the condition of head fracture, part of the water vapor migrates downward, and the steam enters the core of the reactor. (2) almost all the water vapor migrates upward under the condition of the fracture of the head of the seal. Water vapor has not been found to enter the core of the reactor. The significance of this paper is: 1) it can be used for reference for the safety evaluation of the lead based experimental reactor (preliminary safety analysis report, etc.); 2) it can provide a preliminary basis for the design optimization of the lead based research reactor, and then realize the design of "built in" instead of "added on". It can provide some reference value for the future development of commercial lead-based reactors.
【学位授予单位】：中国科学技术大学
【学位级别】：博士
【学位授予年份】：2017
【分类号】：TL411

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