压水堆的刹车失灵预期瞬变事故研究

作为“未解决的安全问题”之一,对压水堆的刹车失灵预期瞬变事故(ATWS)已作了多年研究.作者所进行的最新分析考虑到合并小破口失水事故时压水堆的固有稳定性及其长期表现,并研究了各种不同假定的灵敏度;结果是积极的.     

先进压水堆大破口失水事故耦合特性研究

先进压水堆采用非能动安全壳冷却系统作为事故后安全壳排热手段,事故后以钢安全壳为换热面将释放到安全壳的能量传递到环境中。失水事故后非能动安全壳冷却系统带热能力的好坏关系到整个反应堆的安全,事故进程中反应堆冷却剂系统的非能动特性与安全壳的非能动特性相互耦合,需要将非能动安全壳冷却系统和反应堆冷却剂系统进行耦合分析,了解事故后反应堆冷却剂系统与安全壳的耦合特性。本文通过开展大破口失水事故下反应堆冷却剂系统和安全壳的耦合分析,了解各非能动系统在大破口失水事故工况下的耦合特性。分析结果显示:大破口失水事故下,耦合分析中非能动余热排出系统、非能动堆芯冷却系统、自动卸压系统和非能动安全壳冷却系统的特性尤其是非能动余热排除系统排热功率、内置换料水箱注入时机和流量、自动卸压阀流量、安全壳压力温度等均与单独计算有较大差异,大破口失水事故下耦合分析得到的事故前期安全壳压力、温度峰值小于单独计算,事故后期安全壳压力在地坑水蒸发的作用下会逐步高于单独计算结果。     

压水堆冷管段 2% 小破口失水事故实验研究

在高压综合实验装置（ＨＰＩＴＦ）上进行了压水堆冷管段２％小破口失水事故实验（ＮＳＢ－６），破口方向为冷管段底部，破口面积为２％。实验再现了核电厂发生小破口失水事故时的热工水力学现象，实验结果与ＲＥＬＡＰ５／ＭＯＤ２系统分析程序的计算结果作了比较，验证了该程序对小破口失水事故的分析能力。     

小破口失水事故非能动系统瞬态特性研究

为了解先进压水堆小破口失水事故下非能动安全壳冷却系统、非能动堆芯冷却系统、非能动余热排出系统的瞬态响应特性,需开展小破口失水事故下反应堆冷却剂系统和安全壳的耦合响应特性研究。分析结果表明,小破口失水事故下,耦合分析中非能动余热排出系统、非能动堆芯冷却系统、自动卸压系统和非能动安全壳冷却系统的特性与独立计算有较大差异,小破口失水事故下耦合分析得到的安全壳压力峰值小于独立计算。     

秦山核电厂二期压水堆TRACE程序模型的建立

为进一步深化核安全审评工作,国家核安全局从美国核管制委员会(NRC)引进了大型热工水力最佳估算程序TRACE程序。本文利用TRACE程序以及辅助建模工具SNAP程序,建立了秦山核电厂二期两环路压水堆热工水力模型,进行了大破口失水事故模拟计算,得出了合理的计算结果。     

压水堆核电站大破口失水事故分析

压水堆核电站安全分析报告是核安全监管部门对其进行安全审查的重要文件,大破口失水事故是核电站运行的设计基准事故,是安全分析报告中的重要内容。本文使用RELAP5/MOD3.2进行压水堆冷管段大破口失水事故的计算,对比发现一回路冷管段发生双端断裂大破口时燃料元件包壳温度峰值（PCT）最高,且长时间维持在较高温度,此条件下反应堆最危险。计算结果表明,事故发生后,一回路压力迅速下降,堆芯冷却剂的流动性变差,导致堆芯裸露,燃料包壳温度又重新回升。通过安注系统和辅助给水系统等一系列动作,能保证燃料元件包壳温度不超过1204 ℃的限值。     

SBLOCA叠加高压安注失效事故全范围事故分析

目前,我国现有"二代加"压水堆核电厂的事故分析通常只分析至可控状态,但福岛核电厂事故提示,事故分析仅分析至可控阶段不足以揭示事故后果的全面影响,必须进行全范围事故分析。小破口失水事故(SBLOCA)在核电厂堆芯熔化频率(CDF)中的贡献是概率最大的事故序列之一。本文采用CATHARE程序详细分析了"二代加"压水堆核电厂发生SBLOCA叠加高压安注(HHSI)失效状况下的全范围事故,根据该分析结果初步识别出了现有核电厂安全设计的薄弱环节,对此,本文提出了初步改进建议措施。     

压水堆大破口失水事故高压安注的缓解能力研究

以西屋公司典型的三环路压水堆为参考对象,采用基于RELAP/SCDAPSIM程序开发的压水堆严重事故分析平台,对没有缓解措施的热段25 cm大破口失水事故进行了计算分析,详细研究了堆芯表面峰值温度分别达到1 100K、1 300 K和1 500 K时进行高压安全注射对大破口失水事故的缓解情况.结果显示,高压安全注射的时机对大破口失水事故的进程有着重要的影响,较早阶段的注水能够有效阻止堆芯熔化,较晚阶段的注水会恶化事故进程,加速堆芯熔化.     

600MW压水堆安注箱设计研究

本文用美国核管会热工水力程序TRACE和图形化建模软件SNAP,建立了600 MW两环路压水堆一回路和二回路热工水力系统分析模型,并对安注箱的各设计方案进行大破口失水事故(LBLOCA)模拟计算,通过对比各设计方案在LBLOCA事故下计算出的峰值包壳温度,研究安注箱在大破口失水事故工况下的安注性能,最后给出了优化的设计方案,并提出了可行的设计改进建议。研究结果表明,上腔室和下降段同时注入的方式较冷段注入和下降段注入更有效,且恰当地选取初始安注箱压力,可有效降低峰值包壳温度,提高LOCA裕量。     

小破口失水事故研究综述

对小破口失水事故（ＳＢＬＯＣＡ）及其研究状况进行了综述。描述了典型的压水堆和沸水堆小破口失水事故过程和破口位置、破口尺寸及反应堆冷却泵对失水过程的影响，对现有文献按实验和数值模拟两大类进行了归纳，给出了目前世界上用于小破口失水事故研究的主要设备，对小破口失水事故的研究进行了总结。     

Effect of Loop Seal on Reflood Phenomena in PWR

The blockage of the primary coolant loop with water filled at the crossover legs, denoted “loop seal”, is expected to give a significant influence on core cooling during the reflood phase of a PWR-LOCA. However, the effect of the loop seal has been little investigated. Therefore, it was studied experimentally by using Cylindrical Core Test Facility (CCTF).

The loop seal was cleared in a short time (40–60 s) after reflood initiation by pushing-off of stagnating water in the crossover legs due to steam accumulation and resultant pressure increase in the upper plenum.

Although the core cooling was degraded during the loop seal period, it recovered after the loop seal clearing. The degradation of core cooling during the loop seal period is considered to be caused by the low core-inlet water flow rate and resultantly by the small volumetric fraction of water in the core.

The quantitative estimation about the loop seal effect on the clad temperature was made and it was indicated that the maximum clad temperature would not exceed the allowable upper limit (1,473K) specified in the licensing about the reactor safety even with the loop seal at the beginning of the reflood phase.     

压水堆LOCA放射性源项计算模型及应用研究

根据压水堆冷却剂丧失事故（LOCA）后核素从堆芯迁移、释放至安全壳及环境过程中的产生和消减机理,建立了完整的LOCA放射性源项计算模型,并对模型计算结果进行对比分析,最终将模型应用于第3代压水堆LOCA源项计算分析中。结果表明：本文模型与TACTⅢ程序计算结果的相对偏差在±0.05%以内,与TITAN5程序的碘计算结果的相对偏差在±0.5%以内,本文模型计算准确。对于压水堆各种核电机型,安全壳内核素的去除机制及去除速率不同,导致释放到环境中的I和Cs核素活度变化曲线也不同,¹³¹I、¹³⁴Cs、¹³⁶Cs、¹³⁷Cs在事故后30 d内释放到环境中的累积活度逐渐增大。建立的模型基于完整的核素衰变链,考虑了母核衰变对子核源项的贡献及喷淋或自然去除等作用对元素碘的有效去除过程,通用性强。     

Research of Radioactive Source Term Calculation Model and Application in LOCA for PWR

According to the mechanism of the generation and reduction of the nuclide in the process of migration and release from the core to the containment and the environment after the pressurized water reactor (PWR) loss of coolant accident (LOCA), the calculation model of radioactive source term for LOCA was established. The comparative analysis of model calculation results was carried out. Finally, the model was applied to the source term analysis for the third generation PWR LOCA. The results show that the relative deviation between the calculation results of the model and TACTⅢ code is within ±0.05%, and the relative deviation between the iodine calculation results of the model and TITAN5 code is within ±0.5%, so the model calculation is accurate. For various nuclear motor types of PWRs, the removal mechanism and removal rate of nuclide in the containment are different, resulting in different I and Cs radioactivity release curves. The cumulative radioactivity of ¹³¹I, ¹³⁴Cs, ¹³⁶Cs and ¹³⁷Cs released into the environment within 30 d gradually increases. The established model is highly versatile, which is based on the complete nuclide decay chain, considering the contribution of the precursor nuclides decay to the daughter nuclides, and the effective removal process of elemental iodine by spraying or natural removal.     

A numerical investigation of SBLOCA scenario in nuclear heating reactor

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.     

压水堆大破口失水事故引发的严重事故研究

采用机理性严重事故最佳估算程序SCDAP/RELAP5/MOD3.2，以美国西屋公司Surry核电站为参考对象，建立了1个典型的3环路压水堆核电站的严重事故分析模型，分别对主回路冷段和热段发生25cm大破口失水事故（LBLOCA）导致的堆芯熔化事故进行研究分析。结果表明，压水堆发生大破口失水事故时，堆芯熔化进程较快，大量堆芯材料熔化并坍塌至下腔室，反应堆压力容器下封头失效较早，且主回路冷段破口比热段破口更为严重。     

大破口失水事故工况下碳化硅惰性氧化模型研究

与传统Zr包壳相比,SiC复合包壳具有更好的辐照稳定性、高温机械性能和抗氧化能力,可有效缓解事故进程,增加事故应对时间。在大破口失水事故工况下,SiC复合包壳会与低压高温水蒸气发生惰性氧化反应而持续损耗。SiC材料的惰性氧化反应分为两个过程：SiC抛物线型氧化过程和SiC表面氧化产生的SiO₂的线性挥发过程。本文应用修正的Deal-Grove模型和传热/传质类比法研究SiC的抛物线型氧化速率和SiO₂的线性挥发速率,并基于纯水蒸气环境下SiC氧化实验数据和SiO₂线性挥发实验数据,获得了SiC抛物线型氧化速率常数模型和SiO₂线性挥发速率常数模型。理论模型分析结果显示,在大破口失水事故后低压高温纯水蒸气氧化条件下,SiC材料的氧化速率常数较Zr合金低约2~3个数量级,导致SiC材料的损耗速率远低于传统Zr包壳的损耗速率。     

小型压水堆完全丧失电源引发的严重事故研究

以压水堆严重事故最佳估算程序RELAP/SCDAPSIM/MOD3.4为核心软件,以假想的小型压水堆为研究对象,建立了1个径向3通道、轴向10节块的核反应堆严重事故计算模型,研究了完全丧失电源初因事件引发的严重事故过程,并对事故停堆后蒸汽发生器给水持续300s的缓解措施进行了分析。计算结果表明：蒸汽发生器辅助给水对于延迟事故进程,缓解事故后果具有重要作用。     

Upper Plenum Dump during Reflood in PWR Loss-of-Coolant Accident

Upper plenum dump during reflood in a large break loss-of-coolant accident of PWR is studied with the emergency core coolant injection into the upper plenum in addition to the cold leg. Transient experiments were carried out by injecting water into the upper plenum and the simple analysis based on a one-dimensional model was done using the drift flux model in order to investigate the conditions under which water dump through the core occurs during reflood.

The most significant result is an upper plenum dump occurs when the pressure (hydrostatic head) in the upper plenum is greater than that in the lower plenum. Under those circumstances the flow regime isco-current down flow in which the upper plenum is rapidly emptied. On the other hand, when the upper plenum pressure (hydrostatic head) is less than the lower plenum pressure (hydrostatic head), the co-current down flow is not realized but a counter-current flow occurs. With subcooled water injection to the upper plenum, co-current down flow is realized even when the upper plenum hydrostatic head is less than the lower plenum hydrostatic head. The importance of this effect varies according to the magnitude of water subcooling.     

Analysis of Effect of SBLOCA before and behind Isolation Valve of PRHRS on Passive Reactor Core Cooling System

For the test process of small break loss of coolant accident (SBLOCA) of AP type nuclear power plant, there is a more consistent understanding at home and abroad. However, the influence of the same size of the break on the test process and passive core cooling system in different locations still needs further study. In this paper, a large passive core cooling integrated test facility ACME was used to study the break accident test conditions of passive residual heat removal system (PRHRS) before and behind the isolation valve, and the bottom break test of the cold pipe of core makeup tank (CMT) was used as the contrast condition. The test results show that the accident process of PRHRS before and behind the isolation valve is in accordance with the process of SBLOCA, the core is always in a good cooling statement and the safety of passive core cooling system is effectively verified. There is a certain impact on the accident process for the same break size and different break locations, and the location of the break has a key impact on the CMT level and safety injection flow. In contrast, the heat transfer of PRHRS equipment is also quite different. The heat transfer of cold pipe break and break behind the isolation valve is greater than break before the isolation valve, however, the flow and heat transfer mechanism of PRHRS heat exchange tube needs further study.