DVI管小破口失水事故实验研究

在模块化小型反应堆非能动安全系统综合模拟实验装置上进行了压力容器直接注入（DVI）管小破口失水事故实验,研究了DVI管小破口失水事故过程中的热工水力现象和非能动安全系统运行特性。研究结果表明：模块化小型反应堆DVI管小破口失水事故中,非能动安全系统可对堆芯进行注水,有效导出堆芯衰变热量,保护堆芯安全。     

自然循环反应堆NHR-200Ⅱ失水事故初步研究

破口事故是压水堆最为关注的一类重要事故,其失水量与事故后果严重程度密切相关。NHR-200Ⅱ是由清华大学核能与新能源技术研究院经过多年研究和不断改进,设计的一种全功率自然循环低温供热反应堆,其设计中采用了多种先进的非能动和固有安全设计。本研究针对NHR-200Ⅱ反应堆,选取后果最为严重的控制棒引水管断裂且无法隔离事故,利用系统热工瞬态分析程序对事故过程进行了模拟和分析。结果表明,即使在最严重的破口失水事故下,NHR-200Ⅱ主回路中剩余的冷却剂始终能覆盖反应堆堆芯,并有效通过非能动余热载出系统带走堆芯热量,从而保证反应堆堆芯不会因裸露造成烧毁,这表明NHR-200Ⅱ具有很好的安全特性。     

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。     

多用途一体化轻水堆初步设计方案和安全分析

为满足未来区域性核能供电、核供热、大规模制氢、海水淡化等需求,迫切需要一种结构简单、固有安全性高、经济性高的多用途反应堆.基于此,一种多用途的一体化轻水堆设计概念被提出,包括不同设备的初步设计方案和参数;根据其特点,利用最佳估算程序RELAP5对其中一个设计方案进行了稳压器汽腔破口事故和主泵断电引起的丧失流量事故的确定论安全分析.结果表明,在保守假设条件下,其固有特性和安全系统仍能保证堆芯始终处于被淹没状态,非能动余热排出系统可有效导出停堆后的长期衰变热,从而为进一步研究一体化轻水堆的设计和运行安全特性打下了基础.     

典型压水堆小破口LOCA参数重要度排序表开发

现象识别排序表（PIRT）是反应堆热工水力分析的重要依据,传统PIRT的建立依赖于专家经验,因此缺乏专家经验时难以开展参数的识别工作。本文开展在缺乏专家经验时确定各输入参数重要度排序的研究,选定的工况为典型三回路压水堆（PWR）小破口失水事故（SBLOCA）。参考已有的SBLOCA PIRT,并基于基准计算结果,筛选和补充了可能对目标输出（FOM）具有影响的54个不确定性输入参数。使用一种优化矩独立全局敏感性分析方法计算得到了各输入参数对FOM的敏感性度量和重要度排序。将参数的重要度排序转换为Savage分数,按照Savage分数定性地将所有输入参数进行重要度分组,从而得到了SBLOCA的参数重要度排序表,为压水堆SBLOCA工况的参数排序提供了参考。     

先进堆非能动余热排出系统应对全厂断电事故的能力分析

采用RELAP5/MOD程序对先进堆全厂断电事故进行分析计算,论证非能动余热排出系统对事故的缓解能力.分析表明,先进堆在发生全厂断电事故后,完全能够依靠非能动余热排出系统导出堆芯余热,保证反应堆的安全;先进堆非能动余热排出系统的设计总体上是成功的.     

PRHRS隔离阀前后破口事故对非能动堆芯冷却系统的影响分析

对于AP型核电站小破口失水事故(SBLOCA)试验进程,国内外有较为一致的认识,但对于相同尺寸破口在不同破口位置对试验进程、非能动堆芯冷却系统的影响仍需进一步研究。本文利用大型非能动堆芯冷却整体试验台架ACME开展了非能动余热排出系统（PRHRS）隔离阀前后破口事故试验工况研究,并以堆芯补水箱（CMT）侧冷管底部破口事故工况作为对比工况。试验结果表明：ACME开展的PRHRS隔离阀前后破口事故模拟工况事故进程符合典型SBLOCA进程,堆芯始终处在良好的冷却状态,非能动堆芯冷却系统的安全性得到有效验证;相同破口尺寸工况下,不同破口位置对事故进程有一定的影响,其中破口位置对CMT液位、安注流量的影响较为关键。对比工况中,PRHRS设备换热量也有较大不同,冷管破口和隔离阀后破口工况较隔离阀前破口工况换热量更大,但PRHRS换热管内部流动换热机理需进一步研究。     

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.     

Simulation of the small modular reactor severe accident scenario response to SBO using MELCOR code

Advanced small modular reactors (SMRs) use different design in the systems, structures, components from large reactors for achieving a high level of safety and reliability. In present work, the SMRs severe accident caused by the station blackout (SBO) was modeled and analyzed using MELCOR code, and the simulation of the accident scenario response to SBO was conducted. Based on the steady state calculation, which agrees well with designed values, we introduced the SBO accident for transient calculation. First, the case of the SBO accident without the passive core cooling system (PXS) was calculated. The progression and scenario in the reactor pressure vessel (RPV) and the containment were simulated and analyzed, including the transient response, cooling capacity and thermal-hydraulic characteristics and so on. The station black-out transient in the SMR can be simulated accurately, and the main failure model in the accident process can be concluded. Then three other cases of the SBO accident with different passive safety systems (core makeup tank (CMT), accumulator (ACC), passive residual heat removal system heat exchanger (PRHR HX), automatic depressurization system (ADS)) of the PXS were calculated respectively, and the results for different passive safety systems were compared. The passive core cooling system can not only provide water to the primary coolant system, but also take away the reactor decay residual heat. So in a station black-out transient, we can get more time for restoring AC power, and effectively prevent the accidents such as Fukushima.     

非能动堆芯冷却系统LOCA下冷却能力分析

本文基于机理性分析程序建立了包括反应堆一回路冷却剂系统、专设安全设施及相关二次侧管道系统的先进压水堆分析模型,对典型的小破口失水事故和大破口失水事故开展了全面分析。针对不同破口尺寸、破口位置的失水事故,分析了非能动堆芯冷却系统（PXS）中非能动余热排出系统（PRHRS）、堆芯补水箱（CMT）、安注箱（ACC）、自动卸压系统（ADS）和安全壳内置换料水箱（IRWST）等关键系统的堆芯注水能力和冷却效果。研究表明,虽然破口尺寸、破口位置会影响事故进程发展,但所有事故过程中燃料包壳表面峰值温度不超过1 477 K,且反应堆堆芯处于有效淹没状态。PXS能有效排出堆芯衰变热,将反应堆引导到安全停堆状态,防止事故向严重事故发展。     

一体化小型氟盐冷却高温堆瞬态特性分析

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。     

一体化多用途的非能动小型压水反应堆发生假定事故的初步分析

采用RELAP5/MOD3.2系统程序建立一体化小型反应堆的事故分析模型,包括反应堆冷却剂系统(RCS)、简化的二回路系统和专设安全设施。一体化多用途的非能动小型压水反应堆(SIMPLE)热功率为660 MWt(电功率大于200 MWe)。针对SIMPLE的直接安注管线(DVI)双端断裂事故和DVI2英寸(50.8mm)小破口失水事故(SBLOCA)进行分析。计算结果表明:对于直接安注管线双端断裂事故,破口和自动降压系统(ADS)能有效地使反应堆冷却系统降压,安注箱(ACC)和安全壳内置换料水箱(IRWST)能实现堆芯补水,确保堆芯冷却;对于DVI的SBLOCA,非能动专设安全设施能有效对RCS进行冷却和降压,防止堆芯过热。     

ACME台架PRHR管线破口位置敏感性试验研究

为研究先进非能动（AP）型核电厂在非能动系统失效条件下的安全性能,利用我国先进堆芯冷却机理整体试验台架（ACME）开展了非能动余热排出（PRHR）管线破口失水试验研究,分析了主要的试验进程和破口位置对事故过程各阶段关键参数的影响。结果表明,ACME PRHR管线破口试验进程与冷管段小破口失水事故（SBLOCA）进程基本一致,再现了非能动核电厂自然循环阶段、自动卸压系统（ADS）喷放阶段和安全壳内置换料水箱（IRWST）安注阶段的安全特性;在不同破口位置的试验中,非能动堆芯冷却系统（PXS）均可保证堆芯得到补水,堆芯活性区始终处于混合液位以下;破口位置对ACME LOCA事故进程、反应堆冷却剂系统（RCS）初期降压速率、PRHR热交换器（HX）流量、喷放流量、堆芯液位、IRWST安注流量等参数具有显著影响,对堆芯补水箱（CMT）和蓄压安注箱（ACC）安注流量的影响较小。      

A human–machine interaction design and evaluation method by combination of scenario simulation and knowledge base

A method of designing and evaluating HMI (human–machine interaction) is proposed for the design in supervisory control of fully digitalized I&C (instrumentation and control) and digitalized human–machine interface system, which is a large-scale complex system in the NPPs (nuclear power plants). The proposed method consists of plant accident scenario simulation, knowledge base establishment, and interaction simulation. The plant accident scenario simulation is to analyze the plant behavior and system sequences under the predefined conditions; the knowledge base is modeled based on the simulation results as human and machine roles; and the interaction simulation is to simulate the interactions such as between operator and plant, operator and technical advisor. The proposed method utilizes the object-oriented software named plant DiD (defense-in-depth) risk monitor with the combination of accident simulation by an advanced nuclear safety analysis code such as RELAP5/MOD4. The practical developments for the details are demonstrated using an example practice for the SBLOCA (small break loss of coolant accident) case of passive safety PWR (pressurized water reactor) AP1000.     

PWR冷管段1%小破口失水事故实验研究

在高压综合实验装置（ＨＰＩＴＦ）上进行核电厂反应堆一次系统冷管段小破口失水事故（ＳＢＬＯＣＡ）模拟实验，破口方向为冷管段底部，破口面积为１％（ＮＳＢ－７工况）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分析程序的计算结果上比较，验证了该程序对小破口失水事故的分析能力。     

AP1000主给水管道断裂事故中PRHR系统冷却能力分析

使用机理性分析程序建立包括主冷却剂系统、专设安全设施及相关二回路管道的AP1000核电厂模型,对AP1000核电厂主给水管道断裂事故进程进行计算分析。着重分析了非能动余热排出（PRHR）系统在主给水管道断裂事故工况中的瞬态响应、热工水力行为及其冷却能力,并针对PRHR系统流道阻力特性的不确定性对冷却能力的影响进行分析。分析结果表明,在主给水管道断裂事故中,PRHR系统的热移出功率最终能够与堆芯的衰变功率相匹配,有能力带走衰变热,保证一回路系统最终处于安全停堆状态,不发生堆芯损伤,当PRHR系统阻力系数增加时,PRHR系统的流量和换热功率会降低,对PRHR系统冷却能力造成影响。     

Passive cooldown performance of a 65 MW integral reactor

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.     

福岛条件下AP1000核电厂事故前期研究

假设AP1000核电厂发生类似福岛核事故的初因事件,利用RELAP5/MOD3.3程序对事故早期的一、二回路系统和非能动安全系统进行模拟计算,得到了反应堆冷却剂系统压力、堆芯冷却剂温度、非能动安全系统流量等重要参数的瞬态变化。分析表明：在非能动余热排出系统完好的情况下,反应堆系统能顺利进入热停堆状态;如果非能动余热排出系统1根换热管发生双端断裂,则反应堆系统将会在5 h内发生严重事故。     

二次侧非能动余热排出系统设计及验证分析

提出了一种新型非能动余热排出系统(PRHRS)设计方案,该方案以高位水箱为最终热阱,采用在蒸汽发生器二次侧建立自然循环的方式间接地带走堆芯余热。以大亚湾核电站主冷却剂系统为载体,用RELAP5/MOD3.2程序分析了全厂断电事故下,PRHRS的运行特性。结果表明:事故发生后,余热排出系统内可较快地建立起循环流动,带走蒸汽发生器二次侧热量,在一段时间内保证反应堆安全,证明系统设计合理、有效。并分析了换热器布置高度、系统投入时间及换热面积对余热排出系统运行特性的影响。     

Boiling water reactor fuel heatup during a loss of coolant while refueling

The design of nuclear power plants includes provisions for heat removal from the reactor core in the event there is a loss of reactor coolant while shut down. Boiloff from decay heat can lead to inventory reduction and fuel heatup if no coolant makeup is available. Certain decay heat removal system failures in boiling water reactors can drain the upper vessel and downcomer. This leaves the water inside the core shroud at the same level as the top of the jet pumps. This becomes the starting point from which further inventory reduction is possible through boiloff. This study investigated the core thermal response following such a scenario. A simple model of the core was used for analysis of this sequence. The goal of the analysis was to determine the time at which the water in the core would boil down and fuel heat up to a specified temperature (1256 K). It is this interval during which the operator can take action that will mitigate the transient.