西安脉冲堆大破口失水事故分析

分析了西安脉冲堆大破口失水事故的特点,建立了适用的数学模型,编制了计算程序。结果表明：在大破口失水事故下,部分燃料芯体最高温度将超过设计限值,但不会发生燃料元件熔毁事故。     

西安脉冲反应堆超设计基准事故动态特性分析

为填补以往西安脉冲反应堆(脉冲堆)超设计基准事故研究的不足,利用RELAP5/SCDAP/MOD3.4程序对脉冲堆系统进行了建模计算,给出了脉冲堆在断电ATWS事故和大破口失水ATWS事故下的瞬态响应特性。计算结果表明:发生断电ATWS事故后,在无人为干涉情况下,反应堆部分燃料可能熔毁;发生大破口失水ATWS事故后,破口位置和尺寸对事故后果的严重程度有重要影响,破口位置在堆池底部时,燃料最高温度低于1 800℃,而破口位置高于堆芯下栅板时,将导致燃料元件熔毁。根据脉冲堆在超设计基准事故下的动态响应,针对两种事故工况分别提出了相应的缓解措施。     

新型转换堆(ATR)下降段大破口失水事故的实验研究

为了分析新型转换堆(ATR)的大破口失水事故,在日本动力堆和核燃料开发集团大洗工程中心的 ATR 安全分析实验回路上,进行了三次不同破口直径的 ATR 下降段大破口失水事故实验。通过实验,对影响燃料元件安全的主要参数进行了测量和分析讨论。     

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。     

船用核动力装置小破口失水事故放射性后果分析

建立了小破口失水事故下热工水力分析与放射性源项计算耦合模型,利用研发的反应堆源项放射性计算软件(Nuclear source radioactive compute,NSRC),分别就不同破口尺寸的堆舱放射性泄漏进行了分析和研究,进一步研究了小破口失水事故,冷端安注和热端安注对堆舱放射性影响。结果表明:破口尺寸大小、安全注射位置及破口隔离时间直接影响堆舱放射性泄漏大小。本工作的分析结果为小型船用堆在小破口设计基准事故下,放射性污染后果分析及事故处置提供了依据。     

脉冲堆失水事故计算程序的开发和应用

由于西安脉冲堆的特点，致使国际上通用的瞬时堆芯裸露模型不能使用。中国核动力研究设计院建立了反映西安脉冲堆失水事故机理和过程的真实真芯裸露模型，开发了相应的计算机程序，用于分析和评价西安脉冲堆的安全特性。分析结果表明，真实堆芯裸露模型具有广泛的实用性，可用于计算全部侧面破口和底部破口的失水事故。在破口直径相同的条件下，西安脉冲堆侧面破口失水事故后果比底部破口失水事故严重。在目前的设计条件下，即使发生失水事故，西安脉冲堆也能满足安全准则的要求。     

船用堆小破口失水事故处置的影响因素分析

针对船用堆小破口失水事故处置复杂的特点,利用运行安全分析平台对事故进行了仿真研究,探讨了补水系统、危急冷却系统、二回路设备等对事故处置过程和后果的影响,为运行人员的处理和操作提供了参考,有助于失水事故应急处置规程的制定。     

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。     

小型堆严重事故下安全壳内氢气行为分析

采用MELCOR程序,对小型堆破口叠加全部电源丧失的典型严重事故进行计算,并对安全壳内发生氢气燃烧、爆炸的可能性进行分析。结果表明:主管道直径3.72%的破口叠加全部电源丧失后,堆芯裸露,出现熔堆事故;同时锆水反应产生的大量氢气进入安全壳,使安全壳内氢气含量上升,在安全壳局部空间、屏蔽水箱内出现氢气燃烧。但由于小型堆安全壳净容积较小,水蒸气含量较高,氧气含量较少,不会导致氢气爆炸。     

二回路耗汽运行对船用堆全部电源丧失事故进程缓解分析

针对船用核动力装置的特点,建立了船用堆一、二回路及控制系统的RELAP5模型,用RELAP5/MOD3.2程序对典型船用堆经济巡航工况下发生全部电源丧失事故进行模拟,分析了4种耗汽工况对事故进程的缓解效果。分析表明：事故后合理的耗汽运行方案能明显延缓事故进程,延缓时间为小时级别;耗汽量越小,二回路设备运行时间越长,二回路热阱持续时间也越长,一回路事故进程越缓慢。但太小的耗汽量会引起事故过程中蒸汽发生器（SG）二次侧水位过高甚至满水,进而威胁二回路设备的正常运行。同时,二回路设备数量众多,不同设备的最低运行汽量也不尽相同,应选择最有用和最低耗汽量的设备耗汽运行。本研究能对实际船用堆事故下的应急处理提供参考。     

安注方式对设计基准失水事故下元件包壳破损份额影响的分析

以某船用压水堆为研究对象,采用RELAP5/MOD32程序,分析了发生在主管道冷端的极限中破口失水事故中,采取冷端、热端安注方式时不同的事故过程。引入临界管概念,确定了包壳破损临界功率因子。对全堆进行精细功率重构,确定每根燃料元件功率因子,最终确定不同安注方式下的元件包壳破损份额,并指出：对破口出现在主管道冷段的设计基准事故,热端安注能减轻事故后果,减少破损份额。     

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.     

Thermohydraulic and safety analysis on China advanced research reactor under station blackout accident

A thermohydraulic and safety analysis code-TSACC has been developed using Fortran90 language to evaluate the transient thermohydraulic behavior of the China advanced research reactor (CARR) under station blackout accident (SBA). For the development of TSACC, a series of corresponding mathematical and physical models were applied. Point reactor neutron kinetics model was adopted for solving the reactor power. All possible flow and heat transfer conditions under station blackout accident were considered and the optional correlations were supplied. The usual finite difference method was abandoned and the integral technique was adopted to evaluate the temperature field of the plate type fuel elements. A new simple and convenient equation was proposed for the resolution of the transient behaviors of the main pump instead of the complicated four-quadrant model. Gear method and Adams method were adopted alternately for a better solution to the stiff differential equations describing the dynamic behavior of the CARR. The computational result of TSACC showed the adequacy of the safety margin of CARR under SBA. For the purpose of Verification and Validation (V&V), the simulated results of TSACC were compared with those of RELAP5/MOD3 and a good agreement was obtained. The adoption of modular programming techniques enables TASCC to be applied to other reactors by easily modifying the corresponding function modules.     

池式钠冷快堆事故余热排出系统一回路仿真研究

池式钠冷快堆事故余热排出系统采用了非能动工作原理,依靠液态钠及空气的自然对流排出堆芯余热。为研究事故工况下余热排出系统一回路的换热能力,基于FORTRAN语言,建立堆芯单通道及盒间流模型,采用全隐二阶迎风差分格式及改进的欧拉法离散求解,对事故余热排出系统一回路系统进行数值模拟,并对全厂断电事故进行仿真计算验证。结果表明:该程序能较好地反映事故余热排出系统瞬态变化过程,并可达到超实时仿真。     

秦山第二核电厂乏燃料水池失去冷却后最短响应时间分析

全厂断电事故工况下,反应堆乏燃料水池冷却和处理系统存在较大的停运风险。为避免反应堆乏燃料水池失去冷却事故工况的进一步恶化,使用ORIGEN-S程序计算了不同状态下从乏燃料水池失去冷却到乏燃料组件裸露的最短时间。结果表明,在最恶劣工况下,乏燃料组件裸露的最短时间为79.2h,该结果也被用于制定秦山第二核电厂的应急响应行动计划。     

Assessment of fuel-rod meltdown in a severe accident at Bushehr nuclear power plant(BNPP)

After the Fukushima disaster, interest in the evaluation of severe accidents in nuclear power plants and off-site consequences has significantly increased. Because experimental studies are difficult to conduct, computational methods play a substantial role in accident analysis. In this study, a severe accident in the Bushehr pressurized water reactor power plant caused by a station blackout with a total loss of alternating current power supply has been evaluated. This analysis presents the in-core damage of fuel rods and the release of fission products as well as the thermal hydraulic response of the station components during the loss of active emergency cooling systems. In this manner, a perfect model of the Bushehr nuclear power plant using the MELCOR code is prepared. The accident progression is simulated, and the thermal responses of the fuels and hydraulic components are presented. It is shown that, without operator intervention, steam generators will become dry in approximately 3000 s, and the heat sink of the reactor will be lost. The simulation results show that at approximately 8600 s, the upper parts of the core start melting. This model calculates the shortest available time for accident prevention and proves that the time available is sufficient for operator manual action to prevent a nuclear disaster.     

FHR在全厂断电事故下对RVACS散热能力的要求

氟盐冷却高温堆(Fluoride salt-cooled High-temperature Reactor,FHR)是一种采用包覆颗粒燃料、高温熔融氟盐冷却剂的先进反应堆。部分FHR概念采用了反应堆容器辅助冷却系统(Reactor Vessel Auxiliary Cooling System,RVACS)导出事故下的堆芯余热。RVACS通过导热、对流换热、辐射换热等非能动过程,在事故发生时将堆芯余热排出至大气中。本文采用中国科学院上海应用物理研究所设计的10 MW FHR作为基准,利用RELAP5-MS程序,对其在全厂断电事故下的瞬态过程进行了模拟,验证了RVACS的余热导出能力。本文进一步研究了高反应堆功率情况下的全厂断电事故的瞬态过程,探讨了不同反应堆功率的FHR对RVACS散热能力的要求。     

Overview of high-temperature fuel behaviour with relevance to CANDU fuel

This paper provides an overview of high-temperature phenomena in nuclear fuel elements and bundles, with particular relevance to the CANDU fuel design. The paper describes heat generation, fuel thermal response, and thermophysical properties of the fuel and sheath that can affect the thermal and mechanical response of the fuel element. Sources of chemical heat that can arise during accident conditions in the fuel element are also detailed. Specific phenomena associated with fuel restructuring, fuel sheath deformation, fuel-to-sheath heat transfer, fuel sheath failure criteria, oxidation, hydriding and embrittlement of the Zircaloy sheath, gap transport processes in failed elements, fuel/sheath interaction and fuel dissolution by molten cladding are detailed as important phenomena that can impact reactor safety analysis. Fuel behaviour during a power pulse and fuel bundle behaviour that occurs during a severe reactor accident are further considered. The review also points out areas of further research that are needed for a more complete understanding.     

压力管式超临界水堆极限事故下“无堆芯熔化”概念评估

在自主研发的事故分析程序SCTRAN的基础上,开发并验证了二维导热模型和辐射换热模型,并将改进后的SCTRAN应用于加拿大压力管式超临界水堆在失水事故（LOCA）叠加丧失紧急堆芯冷却系统（LOECC）事故中的堆芯安全评估,并对燃料棒到慢化剂之间的传热效率以及关键的影响因素进行了评估。计算结果表明,在LOCA叠加LOECC工况下,燃料棒到燃料通道的辐射换热和燃料棒到蒸汽的自然对流换热能够有效导出反应堆的衰变余热,最高功率的燃料组件内、外圈燃料棒的最高包壳温度分别为1278℃和1192℃,均低于不锈钢包壳的熔化温度,因此整个事故过程中不会发生堆芯熔化。      

Simulation of the chernobyl accident

An analysis of the April 26, 1986 accident at the Chernobyl-4 nuclear power plant in the Soviet Union is presented. The peak calculated core power during the accident was 550 000 MW_t. The analysis provides insights that further understanding of the plant behavior during the accident. The plant was modeled with the RELAP5/MOD2 computer code using information available in the open literature. RELAP5/MOD2 is an advanced computer code designed for best-estimate thermal-hydraulic analysis of transients in light water reactors. The Chernobyl-4 model included the reactor kinetics effects of fuel temperature, graphite temperature, core average void fraction, and automatic regulator control rod position. Preliminary calculations indicated the effects of recirculation pump coast down during performance of a test at the plant were not sufficient to initiate a reactor kinetics-driven power excursion. Another mechanism, or “trigger” is required. The accident simulation assumed the trigger was recirculation pump performance degradation caused by the onset of pump cavitation. Fuel disintegration caused by the power excursion probably led to rupture of pressure tubes. To further characterize the response of the Chernobyl-4 plant during severe accidents, simulations of an extended station blackout sequence with failure of all feedwater are also presented. For those simulations, RELAP5/MOD2 and SCDAP/MOD1 (an advanced best-estimate computer code for the prediction of reactor core behavior during a severe accident) were used. The simulations indicated that fuel rod melting was delayed significantly because the graphite acted as a heat sink.