乏燃料贮存格架热工水力分析

通过计算流体力学的方法对新型国产乏燃料贮存格架进行热工水力分析,评估新型CPR乏燃料贮存格架在乏燃料池中的局部热工性能,计算在最大水力阻力下,包含放热量最大的乏燃料组件的格架贮存单元的局部最高温度。同时,经过理论计算分析了乏燃料池失去冷却水的极端工况下,乏燃料池的沸腾时间和贮存格架裸露时间。数值计算应用CFX流体分析软件,基于多孔介质模型完成计算分析。分析结果表明乏燃料池局部最高温度低于当地压力下水的饱和温度,满足格架的应用要求;在功率运行工况下失去冷却水,乏燃料水池沸腾时间足以用于采取有效措施应对极端工况。     

某1000 MW压水堆核电厂乏燃料水池扩容热工冷却能力验证

针对国内某1000MW压水堆核电厂乏燃料水池扩容项目,使用计算流体力学(CFD)和理论分析方法,验证了扩容后的乏燃料水池热工冷却能力。在乏燃料水池至少存在一列反应堆水池和乏燃料水池冷却和处理系统(PTR)运行的冷却工况下,乏燃料水池平均水温均满足相应的验收准则,局部最高水温和燃料包壳最高温度均低于当地水的饱和温度。在2列PTR系统均失效的失去冷却工况下,计算出了乏燃料水池平均水温加热到沸腾温度的时间和燃料格架裸露的时间,为运行干预提供了指导。     

用RETRAN程序进行乏燃料元件贮存水池的热工水力安全分析

开发了对核电厂乏燃料贮存水池进行热工水力分析的ＲＥＴＲＡＮ模型,按照最大热功率工况,即在乏燃料贮存水池中装满乏燃料组件（其中包括换料期间刚卸出的全堆芯燃料组件）的条件下用ＲＥＴＲＡＮ模型来评估乏燃料贮存水池冷却系统的冷却能力,并进行了几个假想方案的瞬态计算和校对计算。利用ＲＥＴＲＡＮ模型来评估乏燃料贮存水池稳态和瞬态的热工水力安全分析既方便,又精确,还可用于申请许可证的计算和估算水池的温度分布。     

乏燃料水池冷却的工况分类和安全评价原则研究

确保乏燃料水池内燃料组件的充分冷却是核动力厂设计中要考虑的一个重要问题，新的核安全导则针对确定论分析及燃料装卸和贮存系统设计，要求考虑与乏燃料水池相关的核动力厂状态，包括正常运行、预计运行事件、设计基准事故和设计扩展工况。本文根据我国核安全导则和美国国家标准对于乏燃料水池相关假设始发事件的要求，并参考国内工程实践，给出了针对乏燃料水池冷却应考虑的工况分类。此外，通过调研分析以及根据单一故障准则应用的范畴和例外条款，对乏燃料水池冷却相关工况的温度限值准则和单一故障假设提出建议和指导。     

CPR1000核电厂乏燃料水池重力补水的有效性分析

CPR1000核电厂维修冷停堆工况下如发生全厂失电,汽动辅助给水泵等不依赖动力电源的设备不再可用,将增加了堆芯损坏的风险。设计可利用乏燃料水池和一回路之间的水位差可实现向堆芯的重力补水。本文对影响乏燃料水池重力补水效率的现象进行了分析,并进行建模计算。结果表明,在维修冷停堆工况下,在安全壳未打开条件下,安全壳内压力的升高是降低重力补水效率的主要因素;在最不利的工况下,从乏燃料水池通过重力向一回路补水确保至少在2.7小时内堆芯不会裸露。     

模块式小型堆乏燃料水池冷却系统设计

在福岛核电站事故后,乏燃料贮存安全的重要性得到了广泛重视,业界根据福岛核电站事故的教训,加强了相关研究。多用途模块式小型堆示范工程吸收了福岛核电站事故的经验反馈,在保证乏燃料贮存安全性的同时,兼顾提高模块式小型堆的经济性,在其乏燃料水池冷却系统设计时结合了其他堆型乏燃料水池系统的设计优点。本文从系统调研入手,通过归纳总结三代核电机组乏燃料水池冷却系统的配置特点,研究模块化小型堆的乏燃料水池冷却系统设计方案,并通过使用Flowmaster软件模拟各个工况下乏燃料水池冷却系统的流体特性,对现有的布置条件和设备选型进行校核计算,并基于计算得到的流体参数确定各工况下限流孔板的特征参数和主要工作泵的工况参数等,为设备的设计和采购提供了依据。     

基于SCALE的压水堆乏燃料贮存水池临界安全分析

基于国际先进的核设计与安全分析计算程序SCALE,针对我国自主研发的先进压水堆乏燃料贮存水池,建立恰当的计算模型,并选取合理的保守假设,计算乏燃料水池正常贮存及事故工况下的反应性,评估计算模型的临界安全,分析该程序对我国先进反应堆乏池计算的适用性。计算结果表明该先进压水堆乏燃料贮存水池正常贮存工况及事故工况的有效增值因子均小于0.95,处于次临界状态。该设计模型可确保燃料堆内贮存区域临界状态安全可控。SCALE计算程序适用于我国自主研发的先进压水堆乏燃料水池临界安全计算。     

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。     

核电站乏燃料水池冷却系统虹吸破坏管性能安全分析

国外核电站的运行经验表明,核电站乏燃料水池冷却（PTR）系统的虹吸破坏管性能存在安全隐患,在某些工况下不能有效阻断虹吸流。本文采用RELAP5软件对国内某典型核电站的虹吸破坏管性能进行安全分析。结果表明,在现有的设计条件下,虹吸破坏管无法及时、有效阻断管道断裂后产生的虹吸流动,乏燃料水池冷却水持续从断裂处泄漏,并导致冷却水管道入口露出水面,从而引起乏燃料水池冷却能力丧失,为核电站安全带来极大风险。进一步分析表明,虹吸流引起的乏燃料水池水位下降幅度受断裂点处距水面的高度差、管道流动阻力和PTR系统的管道结构3个因素的共同影响;管道流动阻力可有效缓解和降低由管道断裂引发的虹吸流动的危害性。     

CPR1000核电机组乏燃料水池后备冷却方式设计研究

针对CPR1000核电机组反应堆水池和乏燃料水池冷却以及处理（PTR）系统在某些情况下存在失去设备冷却水的风险,从冷却水源单一的角度分析机组PTR系统存在的问题,结合PTR系统现有的设备,创新性设计出采用其他冷却水源的备用冷却方式。分析研究表明,该设计方案提高了持续冷却乏燃料水池的可靠性,为PTR系统冷却方式增加了多样性和冗余性。      

压水堆核电厂乏池冷却系统扩容改进研究

在分析国内二代改进型百万千瓦核电机组成熟技术的基础上,通过Flowmaster软件计算及设计优化等手段,针对目前已运行和在建核电站的乏燃料水池冷却系统的冷却能力进行评估,提出改进方案增加电站的乏燃料水池冷却系统的冷却能力,并提出满足第三代核电技术对性能及安全性的要求的乏燃料水池冷却系统设计方案。     

Estimation of the amount of surface contamination of a water cooled nuclear reactor by cooling water analysis

Calculations, based upon on-the-spot measurements, were performed to estimate the contamination of NPP primary circuit and spent fuel storage pool solid surfaces via the composition of the cooling water in connection with a non-nuclear incident in the Paks NPP. Thirty partially burnt-up fuel element bundles were damaged during a cleaning process, an incident which resulted in the presence of fission products in the cooling water of the cleaning tank (CT) situated in a separate pool (P1). Since this medium was in contact for an extended period of time with undamaged fuel elements to be used later and also with other structural materials of the spent fuel storage pool (SP), it was imperative to assess the surface contamination of these latter ones with a particular view to the amount of fission material. In want of direct methods, one was restricted to indirect information which rested mainly on the chemical and radiochemical data of the cooling water. It was found that (i) the most important contaminants were uranium, plutonium, cesium and cerium; (ii) after the isolation of P1 and SP and an extended period of filtering the only important contaminants were uranium and plutonium; (iii) the surface contamination of the primary circuit (PC) was much lower than that of either SP or P1; (iv) some 99% of the contamination was removed from the water by the end of the filtering process.     

Prediction of temperature and water level in a spent fuel pit during loss of all AC power supplies

A prediction method for water temperature in a spent fuel pit of a pressurized water reactor (PWR) has been developed to calculate the increase in water temperature during the shutdown of cooling systems. In this study, the prediction method was extended to calculate the water level in a spent fuel pit during loss of all AC power supplies, and predicted results were compared with measured values of spent fuel pools in the Fukushima Daiichi Nuclear Power Station. The calculations gave reasonable results, but overestimated the decreasing rate of the water level and the water temperature. This indicated that decay heat was overestimated and evaporation heat transfer from the water surface was underestimated. Results of calculations with 80% decay heat and 155% (Unit 4 pool) or 230% (Unit 2 pool) evaporation heat flux were in good agreement with measured values. The data-fitted evaporation heat fluxes agreed rather well with the evaporation heat transfer correlation proposed by Fujii et al.     

VVER机组一回路硫酸根升高原因分析

针对核电厂大修期间一回路硫酸根异常升高的问题,首先对一回路和乏燃料水池可能产生硫酸根的物项成分进行分析,排除了给水、硼酸、氢氧化钾等添加试剂是造成硫酸根的主要来源。在对大修期间一回路硫酸根的变化趋势分析时发现,一回路硫酸根的变化和净化系统有关。通过试验确认乏燃料水池中的硼酸溶液在放射性和富氧条件下生成了氧化物质,当乏燃料水池和硼箱净化系统在净化乏燃料水池时,阳树脂中的磺酸基被氧化脱落进而分解生成硫酸根是导致一回路硫酸根升高的主要原因。根据研究成果通过减少阳床的运行时间有效解决了VVER机组中普遍存在的问题。     

Passive cooling of the PBMR spent and used fuel tanks

The pebble bed modular reactor (PBMR) is a new generation high temperature gas-cooled reactor, making use of spherical fuel elements. The spent fuel and partially burnt fuel (called used fuel) is stored in large storage tanks. This paper presents the cooling design of the storage tanks, with special emphasis on its passive cooling ability.For corrosion protection, the tanks are cooled with a closed loop active system, however, passive cooling is seen as the ultimate cooling mode for the storage tanks. If the active cooling fails, the flow automatically bypasses the active system and passive cooling takes over. The active cooling is thus not safety-related; rather its purpose is for investment protection.The storage tank design with its longitudinal internal cooling pipes has a good passive cooling ability. The layout of the tank concrete cubicle ensures that cooling air can flow only in the desired direction. Computational fluid dynamics (CFD) analyses have been done for various heat load scenarios inside the tank. Passive cooling exists for tanks with a low spent fuel fill level with heat load below 25 kW up to a tank containing a full PBMR core (used fuel) with heat load of 640 kW. For all scenarios, the maximum fuel temperature is below 400 °C.A method was developed to calculate the passive cooling characteristics of the tank at a fraction of the time it takes CFD by using the pipe network simulation software Flownex. The method was also used to analyze transient passive cooling events and showed flow phenomenon similar to what CFD analyses have predicted.A small-scale two-dimensional representation of the storage tank and cubicle layout has been built. This experiment demonstrates the passive cooling ability of the tank. It also proved the flow characteristics that were predicted by the CFD and Flownex analyses.It has been shown through diverse techniques that the fuel inside the tanks can be cooled passively. There are still a few aspects which need to be explored in more detail, but overall it can be said that passive cooling of the PBMR spent and used fuel in bulk storage tanks is viable.     

AP1000乏燃料池非能动冷却系统事故后冷却能力分析

AP1000乏燃料池冷却系统采用了先进的非能动设计理念,事故后以池水升温-沸腾的方式带走衰变热,并通过持续的非能动安全补水保证乏燃料安全。对AP1000乏燃料池冷却系统的事故后冷却能力进行分析发现,在核电厂正常换料工况和应急整堆芯卸载工况下,安全水源重力注水能保证事故后72 h内乏燃料安全;在核电厂正常整堆芯换料过程中应等待约56 h,以保证非能动安全壳冷却水箱可为乏燃料池补水,确保堆芯和乏燃料池安全。长期补水可以通过预留的安全接口持续进行。补水手段事故后有效,仅需操纵员有限干预。相对传统乏燃料池冷却系统设计,AP1000能更好地应对冷却丧失的事件。     

Effect of gamma radiolysis on pit initiation of zircaloy-2 in water containing sea salt

In spent fuel pools at the Fukushima Daiichi Nuclear Power Station (1F), seawater was injected for cooling purposes after the tsunami disaster in March 2011. It is well known that the chloride in the seawater has the potential to cause localized corrosion (e.g., pitting corrosion) in metals. In this study, we evaluated the pitting potentials of zircaloy-2, the material used in the fuel cladding tubes in 1F, as a function of chloride concentration. To accomplish this, we used artificial seawater under gamma-ray irradiation and investigated the effect of radiolysis on pit initiation of zircaloy-2 in water containing sea salt. Changes in the composition of water containing sea salt were analyzed as well, both before and after gamma-ray irradiation. The characteristics of the resultant oxide films formed on zircaloy-2 were evaluated by X-ray photoelectron spectroscopy and electrochemical impedance spectroscopy. The experimental results showed that the pitting potential under irradiation was slightly higher than that under conditions in which no radiation was present, and that the pitting potential decreased with increasing chloride concentration in the presence as well as the absence of radiation. Solution analyses for water containing sea salt showed that hydrogen peroxide was generated by irradiation. The oxide film was composed of zirconium oxide and was made thicker during the irradiation. The higher pitting potential could thus be explained by the capacity of hydrogen peroxide to oxidize the surface and enhance oxide film formation. Under gamma-ray irradiation, the zircaloy-2 surface with an oxide film formed by radiolysis products was found to be resistant to pitting in the presence of chloride.     

核电厂乏燃料组件厂内转运关键设备研制

乏燃料组件厂内转运是解决核电厂燃料水池贮存空间不足问题的方法之一。本文分析了乏燃料组件厂内转运的设计准则、安全风险,介绍了用于运输容器内破损组件检测和运输容器内组件冷却用设备的工作原理及其应用情况。应用结果表明:破损检测设备可以快速有效地检测乏燃料运输容器内是否存在破损组件;乏燃料组件冷却设备可以较为安全地冷却装有乏燃料组件的运输容器。      

宁德核电厂PTR系统设计改进方案

指出了M310型反应堆乏燃料水池冷却和处理（PTR）系统运行方面存在的不足,以及宁德核电厂一期工程在该系统上改进的必要性,介绍了其设计改进方案。分析了改进后的系统运行隋况及改进方案的优缺点,提出了进一步改进的建议。