主蒸汽管道破裂叠加蒸汽发生器传热管破裂事件树分析

用事件树分析方法对压水堆核电厂主蒸汽管道破裂诱发的蒸汽发生器传热管断裂进行了事故序列分析 ,找出了引起堆芯裸露的支配性事故序列。结果表明 ,由主蒸汽管道破裂诱发的蒸汽发生器传热管断裂导致堆芯裸露的频率为1 04×10-6/堆·年     

AP1000核电厂SGTR事故工况下CMT水位分析

本文使用LOFTTR2AP-1.6程序分析了AP1000核电厂在蒸汽发生器传热管破裂（SGTR）事故工况下堆芯补水箱（CMT）的水位变化情况.分析结果表明,即使在极端的情况下,SGTR工况也不会导致CMT的水位下降到触发自动卸压系统（ADS）动作的整定值,不会导致更为严重的瞬态,符合压水堆用户要求文件（URD）的规定.     

基于全范围模拟机对压水堆MSLB叠加SGTR事故分析

主蒸汽管道断裂事故叠加蒸汽发生器传热管破裂事故属于核电厂超设计基准事故。为研究国内M310系列机组对该种事故的处理能力,采用了以宁德核电厂1号机为原型的全范围模拟机对此次事故进程进行模拟,选择了放射性释放较为恶劣的蒸汽管道破口(MSLB)叠加100根蒸汽发生器传热管破裂(SGTR)事故,并应用了最新的SOP规程中的操纵员动作以缓解事故后果,分析了事故发生后一回路压力、蒸汽发生器压力、堆芯出口温度以及一次侧至二次侧破口流量的变化。分析结果表明了在核电厂自动动作和操纵员有效及时干预下,在一定情况下可以避免进入严重事故中,最终可以处于安全可控状态。     

SGTR事故SG满溢分析扩展研究

采用热工水力系统程序进行核电厂蒸汽发生器传热管破裂（SGTR）事故蒸汽发生器（SG）满溢分析，验证在该事故下SG不会发生满溢；对SGTR事故进行扩展研究，考虑多种传热管破裂情况，包括单根传热管双端断裂、多根传热管双端断裂和传热管破口，并将3种情况的分析结果进行比较，给出SGTR事故最极限的工况。研究结果表明，单根传热管双端断裂工况下，SG不会发生满溢，且与其他2种工况相比满溢裕量最小，在所有分析工况中最极限。      

EPR缓解SGTR事故的设计特点

蒸汽发生器传热管破裂(SGTR)事故可能造成安全壳旁通,是一个特殊而重要的设计基准事故。本文归纳了EPR缓解SGTR事故的主要设计特点:(1)中压安注(MHSI)泵关闭扬程低于主蒸汽安全阀(MSSV)开启整定值,避免MSSV开启;(2)大气旁排系统(VDA)通过降低其整定值自动启动部分冷却,使一回路快速冷却、降压;(3)蒸汽发生器排污系统(APG)增加转移管线,有助于以排污和蒸汽排放组合方式最终冷却、降压。这些EPR设计特点可供CPR1000核电厂系统设计改进参考。     

蒸汽发生器传热管的降质及对其完整性的评估

蒸汽发生器传热管是反应堆冷却剂压力边界的主要组成部分，这就意味着必须保持传热管的完整性。然而，运行经验表明，蒸汽发生器传热管会出现各种降质。这些降质可能会导致管子的泄漏或破裂，使反应堆冷却剂丧失，并提供了直接通向二回路和释放到环境中去的途径。本文将介绍几种已知的传热管降质，传热管完整性性能准则．并对蒸汽发生器传热管完整性进行评估。     

蒸汽发生器传热管断裂事件树分析

对压水核堆电厂蒸汽发生器传热管断裂（ＳＧＴＲ）事故进行了概率安全分析,给出了功率运行状态下一根或两根ＳＧＴＲ事故导致堆芯裸露的频率为１．２６×１０＾－６／堆·年,并找出了支配性序列及其主要贡献。文章指出了模拟培训中对ＳＧＴＲ事故下正确干预训练的重要性。     

传热管泄漏对蒸汽发生器动态特性的影响

以大亚湾核电站蒸汽发生器为研究对象,建立了基于漂移流理论的蒸汽发生器一维动态数学模型及传热管泄漏模型,并进行了蒸汽发生器不同工况下的稳态仿真。在验证所建立漂移流模型和传热管泄漏模型的基础上,研究了不同工况下传热管泄漏位置及泄漏流量对蒸汽发生器关键参数的影响。研究结果表明,所建立的漂移流模型和传热管泄漏模型能准确反映不同泄漏情况下蒸汽发生器质量含汽率及蒸汽压力等关键参数的变化规律,泄漏发生在热端沸腾段入口处时各参数变化最显著,泄漏量为冷却剂流量的5%时出口质量含汽率由0.261降到0.163。基于漂移流理论传热管泄漏对蒸汽发生器动态特性影响的成功预测,为蒸汽发生器传热管泄漏事故的监测与防范措施的制定提供一定参考。     

CAP1000核电厂全功率范围SGTR事故研究

对CAP1000非能动核电厂在部分功率、零功率和热备用条件下发生的蒸汽发生器传热管破裂（SGTR）事故进行蒸汽发生器满溢评价。对典型的部分功率、零功率和热备用运行条件下的SGTR事故分别进行横向敏感性分析,选取每个运行条件下的保守工况,结合满功率事故工况进行纵向功率谱对比,根据其瞬态特性,分析事故进程,评价极限运行工况和关键参数。结果表明：CAP1000核电厂在全功率范围内发生SGTR事故均不会导致蒸汽发生器满溢,且最严重的工况发生在满功率条件下。     

基于RELAP5程序的AP1000典型事故瞬态特性研究

基于最佳估算程序RELAP5/MOD3.3,对AP1000系统进行了详细的建模分析,选取冷却剂泵卡轴事故、蒸汽发生器(SG)传热管破裂事故和直接注射管线双端断裂事故作为典型事故,获得了典型事故工况下关键参数的瞬态特性和非能动系统响应特性。结果表明:对于冷却剂泵卡轴事故,一回路最大压力为16.82 MPa,燃料包壳表面温度最大值为1 299K,满足验收准则的要求;对于SG传热管破裂事故,破损SG的水体积为231.54m3,小于AP1000蒸汽发生器255.563m3的总容积;对于直接注射管线双端断裂事故,AP1000的非能动堆芯冷却系统能对一回路进行冷却和降压,并防止堆芯裸露和燃料包壳过热。     

Effects of tube rupture modeling and the parameters on the analysis of multiple steam generator tube rupture event progression in APR1400

A multiple steam generator tube rupture (MSGTR) event in APR1400 has been investigated using the best estimate thermal hydraulic system code, MARS1.4. The effects of the parameters such as the number of ruptured tubes, rupture location, affected steam generator on the analysis of the MSGTR event in APR1400 are examined. In particular, tube rupture modeling methods, single tube modeling (STM) and double tube modeling (DTM), are compared. The APR1400 is found to have the capability of allowing more than 30 min to operators for the MSGTR event of five tubes. The effects of rupture location on the MSSV lift time is not significant in the case of STM, but the MSSV lift time for tube-top rupture is found to be 25.3% larger than that for rupture at the hot-leg side tube sheet in the case of DTM. The MSSV lift time for the cases that both steam generators are affected (4C5x, 4C23x) are found to be larger than that of the single steam generator cases (4A5x, 4B5x) due to a bifurcation of the primary leak flow. The discharge coefficient of Cd is found to affect the MSSV lift time only for a smaller value of 0.5. It is found that the most dominant parameter governing the MSSV lift time is the leak flow rate. Whether any modeling method is used, it gives the similar MSSV lift time if the leak flow rate is close, except in the case where both steam generators are affected. Therefore, the system performance and the MSSV lift time of the APR1400 are strongly dependent on the break flow model used in the best estimate system code.     

Analysis of severe core damage accident progression for the heavy water reactor

In this study,the severe accident progression analysis of generic Canadian deuterium uranium reactor 6 was preliminarily provided using an integrated severe accident analysis code.The selected accident sequences were multiple steam generator tube rupture and large break loss-of-coolant accidents because these led to severe core damage with an assumed unavailability for several critical safety systems.The progressions of severe accident included a set of failed safety systems normally operated at full power,and initiative events led to primary heat transport system inventory blow-down or boil off.The core heat-up and melting,steam generator response,fuel channel and calandria vessel failure were analyzed.The results showed that the progression of a severe core damage accident induced by steam generator tube rupture or large break loss-of-coolant accidents in a CANDU reactor was slow due to heat sinks in the calandria vessel and vault.     

The effect of tube rupture location on the consequences of multiple steam generator tube rupture event

A multiple steam generator tube rupture (MSGTR) event has never occurred in the commercial operation of nuclear reactors while single steam generator tube rupture (SGTR) events are reported to occur every 2 years. As there has been no occurrence of a MSGTR event, the understanding of transients and consequences of this event is very limited. In this study, a postulated MSGTR event in an advanced power reactor 1400 (APR1400) is analyzed using the thermal-hydraulic system code, MARS1.4. The APR1400 is a two-loop, 3893 MWt, PWR proposed to be built in 2010. The present study aims to understand the effects of rupture location in heat transfer tubes following a MSGTR event. The effects of five tube rupture locations are compared with each other. The comparison shows that the response of APR1400 allows the shortest time for operator action following a tube rupture in the vicinity of the hot-leg side tube sheet and allows the longest time following a tube rupture at the tube top. The MSSV lift time for rupture at the tube-top is evaluated as 24.5% larger than that for rupture at the hot-leg side tube sheet.     

基于不同传热管材料的SGTR始发事件频率分析

基于不同材料传热管的运行经验,统计总的管临界年数以区分不同核电厂蒸汽发生器传热管数的影响,并将Jeffreys分布作为先验分布统计分析690TT以及经热处理的传热管(包含600TT和690TT)发生破裂的频率。该方法得到的蒸汽发生器传热管破裂(SGTR)始发事件频率可较合理地体现传热管材料性能的改进对降低该事件导致安全壳旁通失效风险的影响及贡献,与通用数据库中未区分传热管材料对应的频率相比明显降低,且随着690TT传热管运行经验的进一步累积,预期SGTR始发事件频率会进一步降低。     

直流蒸汽发生器传热管破裂事故分析

通过对直流蒸汽发生器传热管破裂（SGTR）事故的分析,可看出RELAP5瞬态分析程序能较好地模拟一体化反应堆在SGTR事故后的事件响应序列及主要热工水力现象,例如环路的不对称效应、主回路的自然循环等。一体化反应堆在发生SGTR事故后,可通过一系列安全与保护系统的动作得到有效缓解,并最终能应用非能动余热排出系统（PRHRS）的自然循环导出堆芯余热,使反应堆处于安全状态。同时,受事故影响蒸汽发生器压力在PRHRS投入运行后会快速升高,最终与一回路压力相平衡,此后,破口处的泄漏也会终止。此外,本文还研究了破口处临界流量及其积分流量结果不确定性的影响因素,其中主要考虑了采用不同的临界流模型和破口建模方式等两个方面。     

Analysis of Overheating Rupture in Heat-Transfer Tubes Causing Corrosive High-Temperature Reaction

In the steam generator of a liquid metal fast breeder reactor, a defect penetrating through heat-transfer tube will cause high-pressure water/steam to spout into the low-pressure sodium filling the space outside the tube, to initiate sodium-water reactions. If the leak exceeds an intermediate level (~2kg/s), the reaction jet may rupture adjoining tubes with overheating in the event of insufficient cooling available inside the tubes. Such phenomenon of overheating tube rupture presents a serious problem to the economy and safety of steam generator. With a view to clarifying the failure behavior of steam generator heat-transfer tubes under such condition a model of the phenomenon is derived through a series of tests on sodium-water reactions making use of a test loop representing the scale model of an actual fast breeder steam generator. Comparison of actual test data with analysis based on the model has yielded the following information: The failure behavior of gas-pressurized tubes fall into two categories: (a) by creep failure—occurring upon increase of cumulative damage with tube wall wastage caused by the reaction jet and (b) by ductile failure accompanied by creep—upon tube heating with the reaction jet to the extent of lowering tube wall strength below the hoop stress exerted by tube pressure. Analysis of the two categories of failure results in estimation of the percentage difference between analyzed and measured times to failure of 35–50% in the case of creep failure and of 20–50% in the case of ductile failure accompanied by creep. In practical application to steam generators in order to provide a safety margin a time factor—i.e., the safety factor indicating multiple of actual time to failure—of 3 is adopted against 1.5–2 indicated from test to be the actually applicable value.     

CPR1000蒸汽发生器二次侧三维稳态热工水力分析

采用三维稳态分析软件GENEPI,对CPR1000蒸汽发生器二次侧管束区进行了热工水力计算,利用多孔介质及局部阻力系数来表征传热管及各几何部件的复杂结构和压降影响,得到了二次侧管束区流场、温度场等的分布情况。计算结果表明:管束区最大干度为0.3;将典型传热管的动能数据提供给流致振动软件进行计算分析,结果显示在本工况下,传热管的流致振动在可接受范围内;对管板附近的流场及温度场进行分析,预测了此模型及工况下的泥渣沉积区域,为排污管的设计提供了输入数据。计算结果验证了CPR1000蒸汽发生器二次侧管束区设计的合理性。     

蒸汽发生器大泄漏钠-水反应引起的二回路压力波传播

研究建立了钠冷快堆蒸汽发生器在单管束发生双端断裂情况下钠－水反应中气泡从球形到柱状的变温绝热生长模型,及采用一维特征线方法建立的压力波在快堆二回路中的传播模型。模型中考虑了爆破膜、管壁弹性变形和气蚀的影响。对在两相汽水混合区发生大泄漏后有、无爆破膜情况下的钠－水反应和二回路压力传播瞬态进行了计算,定性分析了其影响以及爆破膜在钠－水反应中的安全保护作用。