主泵参数变化对压力容器液位测量影响分析

为了量化分析CPR1000核电厂主泵特性及相关参数（如电网频率、空泡份额等）变化对堆芯冷却监测系统（CCMS）压力容器液位(L VSL)测量引入的误差,评价该误差对事故处理进程的影响,基于CCMS L VSL测量原理,推导出主泵各参数变化对L VSL测量引入误差的计算公式,并进行量化计算。计算结果表明,除主泵本身的性能降级会导致L VSL较大的低估误差外,其余参数变化对L VSL测量引入的误差可忽略。结合状态导向法事故运行程序（SOP）,分析了主泵本身性能降级导致的低估误差对操纵员关键安全操作的影响。结果表明,该误差可能干扰SOP中主泵的相关操作,但不会阻碍SOP事故处理中关键安全操作的执行。     

硼浓度和不凝气体对压力容器液位测量影响分析

对硼浓度和不凝气体给堆芯冷却监测系统CCMS(Core Cooling and Monitoring System)压力容器液位L VSL测量引入的误差进行了量化计算。计算结果表明,2 000 ppm的硼浓度对L VSL测量引入的误差可以忽略,对40 000 ppm的硼浓度对L VSL测量引入的误差可达13%。对压力容器内充满不凝气体的极端情形,对L VSL测量引入的误差不超过2%。这些误差不会阻碍事故处理安全重要操作的执行。     

水平分层两相流动现象的试验研究和CFD模拟

1引言 在不同类型的小破口失水事故（SBLOCA）情况下,压水堆（PWR）主管道中可能出现分层两相流动区域。UPTF—TRAM试验计划对其进行了详细研究（Liebcrt等,1997）,实施该计划的目的在于模拟事故状态及其处置。失水事故（LOCA）造成的后果包括反应堆压力容器（RPV）的水位下降和蒸汽开始倒向流入蒸汽发生器,因此在热管段中将发生两相混合流动。     

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。      

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。     

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。     

堆芯流量变化对压力容器水位测量影响分析

CPR1000核电厂在每次换料大修期间需执行CCMS(Core Cooling and Monitoring System)校验试验,以获得计算压力容器水位L_(VSL)所需的堆芯动态压头损失系数,完成该试验耗时较长。论文依据调试和换料大修期间一回路冷却剂流量的变化情况评估堆芯动态压头损失系数的变化,并定量评价对L_(VSL)测量的影响。分析结果表明,在回路水力特性未发生明显变化的情形下,对L_(VSL)测量引入的误差很小。建议在L_(VSL)测量不确定度评定时引入堆芯动态压头损失变化的影响,在换料大修时校验流量变化对堆芯动态压头损失的影响是否在允许范围之内,可简化CCMS校验试验,提升机组的经济性。     

用SCDAP／RELAP5评价高压序列下RCS卸压策略

反应堆冷却剂系统（RCS）不同的降压过程是严重事故管理策略的一部分，本文用SCDAP／RELAP5程序对优化动力反应堆（OPR1000）RCS进行了评价。通过给排水操作使二回路降压，从而间接使RCS卸压，评价了无安全注入（SI）情形的小破口失水事故（LOCA）。此外，通过直接降低安全卸压系统（SDS）压力，选择评价了全部给水丧失事故（LOFW）过程。结果表明：二回路给排水操作可以使RCS降压，但不能使其充分卸压。因此，对于1．35in无SI的LOCA破口事故，有必要应用SDS系统使得RCS降压更大，合理的RCS降压时间和能力，可使压力容器有7．5～10．7h的延迟失效时间；而对于LOFW事故。开通两个SDS阀门可使RCS充分卸压，合理的RCS降压时间和能力会使压力容器的失效时间延迟大约5h，只打开一个SDS阀门不能使RCS充分卸压。     

基于神经网络方法的LOCA事故诊断

冷却剂丧失事故(Loss of Coolant Accident,LOCA)是核电厂安全分析中的一类典型事故,不同的破口位置和破口尺寸将直接影响到事故的处置和后果。为判断LOCA事故的破口位置和尺寸,可以借助于神经网络的模式识别功能。针对CPR1000核电系统,利用CATHARE软件建模并仿真不同破口位置和尺寸的LOCA事故,提取事故发生时的6类热工水力参数对BP(Back Propagation)神经网络、Elman神经网络、RBF(Radial Basis Function)神经网络和支持向量机进行训练,再将训练后的神经网络用于破口位置和尺寸的诊断。结果表明,在4种神经网络中,参数优化后的支持向量机对破口位置和尺寸的诊断准确率较高且诊断稳定性较好。在LOCA事故发生时,可以利用支持向量机获取破口的详细信息,辅助操纵员高效地处理事故。     

研究堆稳压器隔离及卸压方案仿真研究

建立了某气体稳压型研究堆在失水事故(LOCA)下局部破口及整体系统的数值仿真模型。针对主管道破口进行数值分析,研究系统流量、压力和破口流量的关系,获得破口的特性参数。通过在系统仿真模型中耦合破口特性参数,对隔离及卸压2种事故下防止稳压器上部气体进入反应堆的应对方案进行了研究。结果表明,破口截面的流量主要取决于系统的压力,采用卸压方案要优于隔离方案。     

Analysis of LSC phenomena of ATLAS cold leg SBLOCA tests using MARS-KS code

Loop seal clearing (LSC) is an important phenomenon for the safety of a pressurized water reactor (PWR) during a small-break loss-of-coolant accident (SBLOCA). The investigation on an LSC phenomenon of 4″, 6″, and 8.5″ break cold leg SBLOCAs simulated by Advanced thermal–hydraulic Test Loop for Accident Simulation (ATLAS) is performed using a Multi-dimensional Analysis of Reactor Safety-KINS Standard (MARS-KS) code. The LSC triggers earlier for larger break sizes during tests and calculations. LSCs occur during the simultaneous sudden decrease of steam condensation rate and the sudden increase of the break volumetric flow rate while the core volumetric flow rate increases slowly in calculation. The five phases of an SBLOCA transient are blowdown, pressure plateau, LSC, boil-off, and core-recovery phase, which can be identified by observing the volumetric flow rate and the time-dependent pressure variation. Loop seal refilling (LSR) occurs due to the trivial steam flow rate to the crossover leg inlet in calculation. The sensitivity analysis shows that the combination of countercurrent flow limitation (CCFL) model option for hot leg and steam generator (SG) inlet (Kutateladze, c = 1.36, m = 1), crossover legs (Kutateladze, c = 1, m = 1), and SG U-tubes (Wallis, c = 1, m = 1) provide good prediction of the LSC phenomenon and thermal-hydraulics behaviors in an SBLOCA transient by MARS-KS code calculation.     

Investigation on two-phase critical flow for loss-of-coolant accident of pressurized water reactor

The previous investigations were mainly conducted under the condition of low pressure,however,the steam-water specific volume and the interphase evaporation rate in high pressure are much different from those in low pressure,Therefore,the new experimental and theoretical investigation are performed in Xi‘an Jiaotong University.The investigation results could be directly applied to the analysis of loss-of -coolant accident for pressurized water reacor.The system transition characteristics of cold leg and hot leg break loss-of -coolant tests are described for convective circulation test loop.Two types of loss-of-coolant accident are identified for :hot leg” break,while three types for “cold leg”break and the effect parameters on the break geometries.Tests indicate that the mass flow rate with convergent-divergent nozzle reaches the maximum value among the different break sections at the same inlet fluid condition because the fluid separation does not occur.A wall surface cavity nucleation model is developed for prediction of the critical mass flow rate with water flowing in convergentdivergent nozzles.     

Accident progression within the secondary containment boundary of large LMFBRs

This study deals with the sodium spillage phenomenon as it relates to accident energetics and containment integrity. Sodium spillage has been identified as an important issue for large LMFBRs because of the large inventory of sodium present and the potential for energetic accidents. Energetic core-disruptive events leading to slug impact could open leak paths in the reactor cover and vent sodium into the secondary containment. Sodium fires in the containment building could lead to pressurization and thermal stressing of the surrounding structure and jeopardize containment integrity. The potential consequences of such a scenario have prompted the development of analytical tools to quantify the spillage process.One of the primary concerns in assessing the integrity of secondary containment is the amount and velocity of sodium which may be ejected from the primary vessel. A parametric study has been performed, the purpose of which was to study the sensitivity of sodium spillage to accident energetics. Treatment of the spillage process was accomplished with the ICECO code employing a quasi-Eulerian method. A 1000 MWe reactor, with prescribed leak paths, is modelled and analyzed during the slug impact phase. Leak paths are assumed to exist as annular penetrations in the reactor cover and as a gap at the vessel-head junction. The behavior of sodium spillage is investigated under conditions of different accident energetics, various opening sizes, and multiple leak paths, with both stationary and moving reactor covers. The relative influence of short and long term spillage is also addressed.During the transient period immediately following slug impact it was found that spillage from annular penetrations in the reactor cover is only weakly sensitive to changes in slug velocity. The same conclusion applies to spillage from a fixed gap at the vessel-head junction. Significant sensitivity of spillage to accident energetics was seen only in cases of spillage from the vessel-head junction when the reactor cover was movable. The influence of slug impact on the motion of the reactor cover leads to the conclusion that sodium spillage is most sensitive to accident energetics inasmuch as it affects the size of the leak path.     

The effects of break location on PWR small break LOCA: Experimental study at the ROSA-IV LSTF

This paper discusses the effect of break location on the break flow rate and break flow quality transitions during a small-break loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). Results from five experiments conducted at the ROSA-IV Large Scale Test Facility (LSTF) are compared for this purpose. These experiments simulated a 2-inch break at the lower plenum, upper head, pressurizer top, cold leg, and hot leg, respectively. The controlling phenomena for the break flow quality transitions in cold-leg and hot-leg break experiments are described.     

基于ECC安注热混合试验的LOCUST 1.2分析与验证

在失水事故（LOCA）工况下安注系统投入使用时,蒸汽与安注冷却剂会发生流体热力学混合,热混合过程中冷腿段的冷却是直接影响堆芯再淹没与否的重要因素。中国广核集团有限公司自主研发了一款两相流热工水力系统分析软件LOCUST,可用于压水堆核电厂事故工况的分析计算。基于西安交通大学堆芯应急冷却系统（ECCS-XJTU）试验台架进行的堆芯应急冷却（ECC）安注热混合试验,本文使用LOCUST软件对ECC热混合试验进行了几何建模及计算分析。ECC热混合试验工况主要为不同流量下主管纯蒸汽与安注管过冷水的混合,蒸汽流量为25~125 kg/h,过冷水流量为100~500 kg/h。模拟计算结果和试验结果的对比分析表明：试验段出口质量流量计算值的最大相对误差在13.8%以内,混合后温度计算值的最大相对误差在8%以内,LOCUST在计算高温蒸汽和过冷水混合时的计算结果相对保守,总体上验证了LOCUST在LOCA下两相热混合安注计算的可靠性和准确性。     

新风量对主控室可居留性的影响分析及优化

主控室是对核电厂正常运行和事故状态实施控制的重要场所,应当采取适当措施和提供足够的信息保护室内的工作人员。就核事故而言,目前的可居留性评价通常考虑相对固定的新风量,没有考虑非过滤渗入途径与新风量的相互影响和制约。本文首先对主控室内人员受照剂量的计算方法进行了讨论,分别分析了事故源项以惰性气体为主、以气溶胶和碘为主以及两者并存时人员受照剂量随新风量的变化。在此基础上结合典型的主控室设计参数和LOCA事故源项,对主控室可居留性系统的新风量进行了敏感性分析,尝试确定最优新风量。此外分析了非过滤渗入与新风量相互制约、非过滤渗入相对固定等多种情形下对主控室人员受照剂量的影响,并初步讨论了动态调整循环回风过滤对降低事故后主控室工作人员剂量的可行性。通过本研究,可以为不同的核电厂主控室可居留系统设计方案的改进和优化提供参考。     

百万千瓦级压水堆严重事故后再注水的有效性评价

根据现有的设计资料,使用一体化严重事故分析程序MELCOR1.8.6建立了核电厂一、二回路系统,非能动堆芯冷却系统和安全壳系统的模型,并模拟冷段2英寸(5.08cm)小破口叠加重力注入失效的严重事故发生后,将冷却剂注入堆芯的情形,分析其对严重事故进程的缓解能力。本文选取3个严重事故的不同阶段,将冷却剂分别以小流量(10kg/s)、中流量(50kg/s)和大流量(200kg/s)的速率注入堆芯,通过比较氢气产生量、堆芯放射性产生量及堆芯温度等数据来评估在严重事故不同阶段再注水的可行性。结果表明:在堆芯损伤初期,可认为10kg/s以上的流量足以冷却百万千瓦级事故安全。而当严重事故发展到堆芯开始坍塌阶段,200kg/s的注水流量可认为是基本可行的,而小于此流量的注水应慎重考虑。