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。     

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。     

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。      

4×4燃料组件考验装置小破口失水事故分析

采用ＲＥＬＡＰ５／ＭＯＤ３热工水力瞬态分析程序，对４×４燃料组件考验装置（以下简称考验装置）小破口失水事故进行了分析计算，预计小破口失水事故下堆芯的热工水力行为（选取当量直径为φ４ｍｍ小破口）。分析结果表明：在发生当量直径为φ４ｍｍ的小破口失水事故下，考验装置专设安注系统能确保考验堆芯安全，且不会危及高通量反应堆。     

AC600全压堆芯补水箱补水实验研究

全压堆芯冰箱（ＣＭＴ）是ＡＣ６００压水堆非能动高压安注系统的主要设备。全压堆芯补水箱补水实验主要研究中,小破口失水事故时ＣＭＴ的重力排放特性,为验证安全分析计算机程序试验数据,中国核动力研究院建造了ＣＭＴ补水实验装置,并在该装置上模拟反应堆主管道中,小破口失水事故动态工况,完成了ＣＭＴ补水实验,本文给出了小破口失水事故工况堆芯水箱补水试验结果与分析。,     

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。     

压水堆SBLOCA事故环路水封清除现象机理与模型影响研究

环路水封清除（LSC）是压水堆冷管段小破口失水事故（SBLOCA）的典型事故特征之一。为确定LSC现象的物理模型影响,探究准确复现LSC现象的物理模型设置,从LSC现象物理机理的角度,对影响LSC的主要物理模型进行梳理和分析。结合LOBI台架SBLCOA系列实验,对LSC现象物理机理、物理模型影响进行模拟和验证。结果表明,在对影响LSC现象的物理模型进行合理设置后,RELAP5程序模型能较好地复现LOBI台架实验工况中的LSC现象,验证了LSC现象物理模型影响及模型设置的合理性。      

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。     

低温堆上空腔失水事故模拟实验研究

叙述了位于低温堆上空腔位置的中小尺寸管道破裂引起的小破口失水事故研究。在核供热堆热工水力学实验系统ＨＲＴＬ－５上，对停堆后堆内有剩余功率的上空腔小破口失水事故进行了模拟实验，分析了小破口失水事故发生后，系统运行重要参数的变化，给出了上空腔小破口失水事故对低温安全性的影响。     

基于PSA的压水堆LBLOCA不确定性分析

为了结合确定论与概率论分析开展更加真实的核反应堆事故工况安全分析,提出了一种结合概率安全分析（PSA）和最佳估算加不确定性（BEPU）分析的方法,并以典型三环路压水堆冷管段双端断裂大破口失水事故（LBLOCA）的极限事故为对象,首先基于PSA开展了应急堆芯冷却系统的事故失效分析,而后结合BEPU分析评估了事件树中各事故序列的包壳峰值温度（PCT）分布及条件堆芯损坏概率（CCDP）,最终确定了压水堆在该事故工况中的堆芯损坏频率（CDF）。分析结果表明,压水堆在冷管段双端断裂工况中应急堆芯冷却系统能够保证反应堆的安全,且一列低压安注系统足以排出堆芯余热及保证反应堆安全。      

RELAP5/MOD3.3程序对非能动核电厂小破口失水事故的适用性研究

AP1000核电厂采用非能动堆芯冷却系统来缓解小破口失水事故(SBLOCA),缓解事故的理念是流动冷却。RELAP5/MOD3.3程序适用于传统核电厂SBLOCA研究,对于非能动电厂SBLOCA研究的适用性需重新研究与评估。本工作基于非能动电厂小破口失水事故的分析,结合RELAP5/MOD3.3的结构与模型,对其进行评估和改进。为验证改进后的RELAP5/MOD3.3的适用性,以AP1000小破口失水事故的验证试验台架APEX-1000为模拟对象,分析模拟DBA-02、NRC-05事故工况。分析结果表明,改进后的RELAP5/MOD3.3的计算结果与试验数据符合较好。     

A blind simulation of RD-14M small-break LOCA experiments using CATHENA code

The results of two Small Break Loss of Coolant Accident (SBLOCA) experiments in RD-14M test facility and their predictions by CATHENA code, which is used to analyze postulated events in CANDU reactors, are compared in this paper. Two specific SBLOCA experiments selected for the CATHENA code predictions are B9006 and B9802. Test B9006 is a 7-mm inlet header break experiment with pressurized accumulator emergency coolant injection and represents most complete SBLOCA test conducted in RD-14M. Test B9802 is a 3-mm inlet header break experiment with full channel power to study boiling in channels and condensation in steam generators in a slowly depressurizing loop rather than a blow-down.These blind simulations demonstrate that CATHENA code is capable of adequately predicting the primary pressure depressurization, channel flow rate, channel voiding for tests B9006 and B9802, and the high pressure core injection flow by CATHENA accumulator model and switching time from high pressure to low pressure injections for test B9006. The CATHENA predictions of fuel sheath temperatures for test B9006 are in much better agreement with the test measurements than those for test B9802, because CATHENA code could not capture the oscillatory behavior of channel flow and consequently sheath temperature when local fuel sheath surface was being intermittently dried and rewet under near stagnant flow and full power conditions of test B9802. Therefore, it is concluded that further works are required to appropriately predict the sheath temperature spike and its fluctuation during the transient of test B9802 where the critical heat flux and post-dryout heat transfer are important governing phenomena.     

Scaling methodology for a reduced-height reduced-pressure integral test facility to investigate direct vessel injection line break SBLOCA

A scaling methodology for a small-scale integral test facility was investigated in order to analyze thermal-hydraulic phenomena during a DVI (direct vessel injection) line SBLOCA (small break loss-of-coolant accident) in an APR1400 (advanced power reactor 1400 MWe) pressurized water reactor. The test facility SNUF (Seoul National University Facility) was utilized as a reduced-height and reduced-pressure integral test loop. To determine suitable test conditions for simulating the prototype in the SNUF experiment, the energy scaling methodology was propose to scale the coolant mass inventory and the thermal power for a reduced-pressure condition. The energy scaling methodology was validated with a system code (MARS) analysis for an ideally scaled-down SNUF model and that predicted a reasonable transient of pressure and coolant inventory when compared to the prototype model. For the actually constructed SNUF, the effect of scaling distortions in the test facility's thermal power and the loop geometry was analytically investigated. To overcome the limitation of the thermal power supply in the facility, the convective heat transfer between primary and secondary systems at the steam generator U-tubes was excluded and a modified power curve was applied for simulating the core decay heat. From the code analysis results for the actual SNUF model, the application of the modified power curve did not affect the major events occurring during the transient condition. The results revealed that the scaling distortion in the actual SNUF geometry also did not strongly disturb significant thermal-hydraulic phenomena such as the downcomer seal clearing. Thus, with an adoption of the energy scaling methodology, the thermal-hydraulic phenomena observed in the SNUF experiment can be properly utilized in a safety analysis for a DVI line break SBLOCA in the APR1400.     

燃料棒束换热(RBHT)试验对低压下堆芯漂移流模型的评价

我国目前正在发展基于非能动技术的三代核电,为评价和改进非能动核电厂小破口失水事故在低压下棒束区的漂移流模型,采用燃料棒束换热(RBHT)试验对EPRI[6]、Cunningham-Yeh[4]模型,Bestion[7]漂移流模型进行了计算分析,结果表明燃料棒束换热试验RBHT试验数据工况能涵盖非能动核电厂在低压下的参数,不需要建造针对燃料棒束的试验台架,Cunningham-Yeh[4]和Bestion[7]漂移流模型基本适用,而EPRI[6]在低压区过高预测了空泡份额,不适用非能动核电厂。     

Experiments on slug mixing under natural circulation conditions at the ROCOM test facility using high-resolution measurement techniques and numerical modeling

ROCOM is a four-loop test facility used for the investigation of coolant mixing in the primary circuit of pressurized water reactors. Recently, a new sensor was developed for an improved visualisation and quantification of the coolant mixing in the downcomer. This new sensor array spans a dense measuring grid and covers nearly the whole downcomer. In the presented work, special emphasis was given to the comparison of the data of this sensor with the results of calculations using the Computational Fluid Dynamics (CFD) code ANSYS CFX. A coolant mixing experiment during natural circulation conditions has been conducted. The underlying scenario of this experiment is based on a boron dilution scenario following a SBLOCA event. The corresponding CFD code solution has been obtained using the Best Practice Guidelines. All main effects observed in the measurement are described by the calculation. The detailed comparison reveals that the calculation underestimates the coolant mixing inside the reactor pressure vessel.The measurement data, boundary conditions of the experiment and facility geometry can be made available to other CFD code users for benchmarking.     

SCDAP/Relap5 3.2程序分层夹带模型敏感性分析

在AP1000中,连接堆芯补水箱和冷腿间的压力平衡管线中的气泡份额决定了堆芯补水箱的注入量,其中,气泡源自冷腿中的分层夹带。为研究AP1000核电站中气-液分层夹带现象对堆芯非能动余热排出系统的整体特性的影响,本文以Relap5/Mod3.2作为计算平台,建立了AP1000小破口失水事故模型并进行了数值计算,对比了采用与不采用水平分层夹带模型的计算结果,发现该模型对事故发展有重要的影响。     

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。     

RETRAN's improved small break modeling capability

The RETRAN-02 computer code has been used extensively to analyze both operational transients and small break loss-of-coolant accidents (SBLOCA). While the results of these analyses are generally in good agreement with data, the explicit solution method sometimes requires small time-steps for portions of the small break calculations. The RETRAN-03 code, now under development, is incorporating additional models to extend the capability of RETRAN-02 and improve the code efficiency for SBLOCA's. Included in these models is a stable, implicit solution method which has provided a significant reduction in CPU requirements while maintaining satisfactory analyses results. The new models for RETRAN-03 and results of small break analyses using RETRAN-03 are briefly discussed in this paper.