未来核电机组安全壳消氢系统的工艺方案研究

本文立足于新版HAF法规的要求，参考大田湾核电站、秦山三期和岭澳二期的安全壳消氢系统设计，对用于未来核电机组安全壳内可燃气体控制的安全壳消氢系统的设计功能、消氢设备选择配置及系统运行方案进行初步研究，重点对非能动氢复合器（PAR）和点火器这两种消氢设备的特点进行比较，并针对未来核电机组提出PAR＋点火器的系统初步工艺方案。     

一个换料周期前后非能动消氢复合器催化片消氢效率对比试验

从设备功能、设备组成、现场布置、在役试验方法和周期等方面介绍秦山第二核电厂3、4号机组使用的非能动消氢复合器。对经过1个换料周期使用前后的4组催化片进行消氢效率对比试验,对比分析调试与生产过程中容易发生的锈蚀、油污等问题对催化片消氢效率的影响。试验结果表明:经过1个换料周期后,由于受油污、灰尘等因素的影响,催化片的消氢成功时间与初始结果相比略有延长;少量锈蚀和油污对催化片的消氢效率未产生显著影响,但锈蚀容易造成催化片穿孔、破损,油污板片试验过程中产生的烟气容易对消氢过程产生干扰。     

大型干式安全壳消氢系统的初步设计

以岭澳核电站为分析对象,利用MELCOR和TONUS(CEA)程序进行分析计算,给出了初步的消氢系统设计方案,对不同核电站的消氢系统设计方案进行了对比和讨论.结果表明:安全壳内安装33个FR750型或者17个左右的FR1500型氢气复合器可以满足氢气控制要求.     

核电厂非能动氢气复合器消氢特性试验研究

根据非能动氢气复合器(PAR)的工作状态特点和启动阈值、停止阈值、消氢能力、点火阈值等关键特性参数的要求,设计建立能够模拟安全壳内事故环境条件、在非能动条件下开展PAR关键特性参数验证试验的试验装置,制定相应的试验方法,开展启动阈值试验、启动时间试验、消氢能力试验和点火阈值试验等,获得PAR的关键特性参数。试验结果表明:PAR关键特性在不同的试验参数条件下测试结果也不同;在制定PAR消氢特性参数要求时需要限定试验方法和试验参数条件,以便获得统一的、定量的PAR的消氢特性参数。     

大亚湾和岭澳核电站安全壳氢气风险及消氢措施有效性分析

使用MAAP程序计算大亚湾和岭澳核电站严重事故条件下安全壳内的相关质能释放和氢气源项;利用TONUS程序建立安全壳集总参数模型,计算分析氢气在安全壳内的分布情况;结合非能动氢复合器消氢性能、现场条件和氢气分布情况,提出氢复合器布置方案;借助TONUS和GASFLOW程序,分别使用集总参数法和CFD法,验证消氢方案的有效性。验证结果表明,安全壳内氢气浓度满足相关法规要求。     

非能动氢复合器性能验证试验方法研究

消氢启停阈值和消氢速率是非能动氢复合器的关键性能参数。本文设计了一种直观方便的非能动氢复合器性能验证试验方法:将非能动氢复合器放于密闭容器中，并通入氢气，只要氢复合器启动消氢反应且整条消氢过程曲线在给定值直线A以下，则验证了启动阈值不大于给定值A；只要消氢过程曲线最终的水平段在给定值直线B以下，则验证了停止阈值不大于给定值B；只要氢复合器达到稳定消氢状态，通入容器的氢气质量流量即为消氢速率。本文设计并搭建了试验装置，采用非能动氢复合器样机PARQX-15进行消氢性能验证试验，成功验证了消氢启动阈值<2%（体积浓度，下同），停止阈值<0.5%，消氢速率大于536 g/h，证明了试验方法的实用性和有效性。      

AP1000非能动安全壳冷却水WGOTHIC分析

本文应用WGOTHIC程序对AP1000核岛整体分工况建模,系统分析了多种情况下冷却水装量对安全性的影响。结果表明：非能动安全壳冷却系统失效1 000 s后,安全壳超压;冷却水冷却72 h后得不到冷却水的补充,0.9 d后安全壳超压;冷却水冷却19.6 d后,安全壳虽超压,但小于安全壳屈服极限压力;冷却水冷却30 d后,空气冷却已足够带走堆芯衰变热,而不需人为干预。结果为应急计划制定和设计改进提供了依据。     

冷凝水收集装置对非能动安全壳热量导出系统影响试验研究

非能动安全壳热量导出系统（PCS）作为三代核电厂重要的安全系统,用于事故后安全壳的非能动冷却。利用大型安全壳综合试验装置,可开展安全壳内复杂的热工水力现象与安全系统之间耦合行为的研究。本文利用大型安全壳综合试验装置开展了PCS换热器冷凝水收集装置对PCS排热影响及收集率试验。结果表明,在工况范围内,换热器下方安装冷凝水收集装置对PCS的换热能力没有明显的不利影响,且其收集率较高。     

氢气催化复合器对核电厂严重事故的缓解效果

在严重事故条件下,安全壳内的氢气燃烧或爆炸威胁安全壳完整性,必须采取措施减小或消除安全壳的氢气风险。针对600MWe级核电厂的大型干式安全壳,以小破口失水诱发的严重事故序列为基准事故,计算分析了氢气催化复合器（PAR）消除安全壳内氢气的效果,及复合效应对安全壳压力温度的影响。研究表明：氢气催化复合器能够持续稳定地消除安全壳内氢气,但对于极其快速的氢气释放,它的消氢能力受到一定限制。     

安全壳模型装置内氢气分布特性及影响因素分析

采用计算流体力学方法,首先利用THAI HM-2实验对CFX分析模型的适用性进行验证,通过与实验数据的比对,表明计算结果与实验数据基本吻合,从而验证选用的模型适合对安全壳模拟装置氢气分布特性的分析。之后,建立待研究中等规模安全壳模型实验装置的三维几何模型和网格模型,采用基准工况+单因素对比的方式,分别模拟湍流浮力射流中心喷射和近壁面喷射工况以及考虑蒸汽壁面冷凝情况下安全壳模型内的氦气(氢气替代工质)流动扩散分布,讨论喷射位置因素、壁面蒸汽凝结效应对氦气分布的影响。分析结果表明,喷射位置对氦气分布的影响主要体现在壁面引流现象上,即氦气流更倾向于沿着安全壳壁面进行流动和扩散;而与安全壳壁面的换热和蒸汽的冷凝会进一步促进大空间自然对流的建立,从而较为显著地提高氦气在安全壳内的扩散和混合效果。     

AC600非能动安全壳冷却系统长期效应分析

利用自主开发的用于先进压水堆AC600非能动安全壳冷却系统的专用三维热工水力分析程序PCCSAC－3D，对AC600安全壳在大破口失水事故情况下进行了长期效应分析，该程序把钢安全壳内部的工质分为水蒸汽，不可凝干空气，连续相水和非连续相水，对气相引入k-ε湍流计算模型并考虑由于气体浓度差引起的扩散效应。PCCSAC－3D程序充分考虑了各种空间非均匀的物理因素的影响，能够较精细描述在发生核电厂设计基准情况下出现与安全壳非能动冷却系统有关的各种物理现象，本文对安全壳进行长期效应的分析结果表明，AC600非能动安全壳冷却系统能够保证安全壳的完整性。     

基于系统工程方法的HPR1000应急堆芯余热排出系统设计研究

为进一步提高核电厂的经济性与竞争力,基于系统工程方法,针对"华龙一号"核电机组(HPR1000)应急堆芯余热排出系统开展设计研究,综合考虑安全性、经济性以及技术成熟度等要求,以核电厂工程应用和核电厂整体技术指标最优为目标,构建系统评估指标体系,并运用层次分析法(AHP)分析应急堆芯余热排出系统的最优化设计方案。研究表明,取消汽动辅助给水系统,将非能动余热排出系统(PRS)的功能扩展至缓解预计运行事件和设计基准事故可能是HPR1000应急堆芯余热排出系统更为优化的方案。     

CAP1400非能动余热排出系统自然循环试验风险分析与应对措施

大型先进压水堆(CAP1400)非能动余热排出系统(PRHR)自然循环试验是CAP1400首堆试验项目之一,也是调试期间的重大瞬态试验。试验过程中,由于反应堆一回路温度、压力和液位等参数剧烈变化,增大了试验风险,对机组运行控制提出了较高要求。本文在AP1000调试实践的基础上,从降低自然循环试验风险角度分析提出利用功率运行后的真实衰变热执行本试验。同时针对试验过程一回路压力、温度,稳压器(PZR)液位及堆外源量程等参数剧烈变化产生的安全风险分析,并制定相应的应对措施,为后续CAP1400 PRHR自然循环试验安全实施提供有力支撑。     

一体化先进堆全厂断电事故下非能动余热排出系统能力分析

中国核动力研究设计院（NPIC）设计的中国一体化先进堆（CIP）余热排出系统是非能动系统。采用RELAP5／MOD程序分析计算该堆全厂断电事故后堆芯核功率、堆芯平均温度、一回路和二回路压力,以及非能动余热排出系统功率随时间的变化,论证了非能动余热排出系统对事故的缓解能力。分析结果表明,CIP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。     

A PANDA integral test on the effect of light gas on a Passive Containment Cooling System (PCCS)

As part of the Euratom project TEMPEST (Testing and Enhanced Modelling of Passive Evolutionary Systems Technology for Containment Cooling), a series of five tests was performed in the PANDA facility to experimentally investigate the distribution of hydrogen inside the containment and its impact on the performance of the Passive Containment Cooling System (PCCS) designed for the Economic Simplified Boiling Water Reactor (ESBWR). In a postulated severe accident a large amount of hydrogen could be released in the Reactor Pressure Vessel (RPV) as a consequence of the cladding Metal-Water (M-W) reaction and discharged together with steam to the Drywell (DW) compartment. The retention of hydrogen in the DW, instead to be vented in the Wetwell (WW), has a positive effect toward the mitigation of the system pressure build-up. Hydrogen retention in the DW is a consequence of the stratification phenomena driven by the steam-hydrogen density difference. The paper presents the experimental results of the integral Test T1.2 performed in the PANDA facility. Helium was used to simulate hydrogen and the specific PANDA facility configuration included a dead-end volume, allowing for retaining a portion of the released helium in the DW compartment. The results from Test T1.2 showed that the containment end pressure is mainly determined by the redistribution of non-condensable gas inside the containment system and the temporary deterioration of the PCCS performance during the helium release phase plays a minor role.     

非能动余热排出系统安全分析程序SAC-PREARS 的功能与应用

SAC-PREARS 是一个用于分析非能动RHRS稳态和瞬态安全特性的专用程序.通过实验验证的用于AC-600 非能动 RHRS安全分析的MISAP 程序,对SAC-PREARS程序进行了稳态计算验证.并应用SAC-PREARS程序对200 MW 核供热堆非能动RHRS稳态和瞬态热工水力特性进行了分析,得出了具有工程意义的结论.     

Functional reliability analysis of Safety Grade Decay Heat Removal System of Indian 500 MWe PFBR

Passive systems are increasingly deployed in nuclear industry with an objective of increasing reliability and safety of operations with reduced cost. Methods for assessing the reliability of thermal-hydraulic passive systems, that is systems with moving working fluid, address the issues in natural buoyancy-driven flow that could result in a failure to meet the design safety limits under accident scenarios. This is referred as design functional reliability. This paper presents the results of functional reliability analysis carried out for the passive Safety Grade Decay Heat Removal System (SGDHRS) of Indian Prototype Fast Breeder Reactor (PFBR). The analysis is carried out based on the overall approach reported in the Reliability Methods for Passive System (RMPS, European Commission) project. Functional failure probability is calculated using Monte-Carlo method and also with method of moments.     

CFD evaluation of hydrogen risk mitigation measures in a VVER-440/213 containment

In the PHARE project “Hydrogen Management for the VVER440/213” (HU2002/000-632-04-01), CFD (Computational Fluid Dynamics) calculations using GASFLOW, FLUENT and CFX were performed for the Paks NPP (Nuclear Power Plant), modelling a defined severe accident scenario which involves the release of hydrogen. The purpose of this work is to demonstrate that CFD codes can be used to model gas movement inside a containment during a severe accident. With growing experience in performing such analyses, the results encourage the use of CFD in assessing the risk of losing containment integrity as a result of hydrogen deflagrations. As an effective mitigation measure in such a situation, the implementation of catalytic recombiners is planned in the Paks NPP. In order to support these plans both unmitigated and recombiner-mitigated simulations were performed. These are described and selected results are compared. The codes CFX and FLUENT needed refinement to their models of wall and bulk steam condensation in order to be able to fully simulate the severe accident under consideration.Several CFD codes were used in parallel to model the same accident scenario in order to reduce uncertainties in the results.Previously it was considered impractical to use CFD codes to simulate a full containment subject to a severe accident extending over many hours. This was because of the expected prohibitive computing times and missing physical capabilities of the codes. This work demonstrates that, because of developments in the capabilities of CFD codes and improvements in computer power, these calculations have now become feasible.     

Sensitivity analysis of the MASLWR helical coil steam generator using TRACE

Accurate simulation of transient system behavior of a nuclear power plant is the goal of systems code calculations, and the evaluation of a code's calculation accuracy is accomplished by assessment and validation against appropriate system data. These system data may be developed either from a running system prototype or from a scaled model test facility, and characterize the thermal hydraulic phenomena during both steady state and transient conditions. The identification and characterization of the relevant thermal hydraulic phenomena, and the assessment and validation of thermal hydraulic systems codes, has been the objective of multiple international research programs. The validation and assessment of the best estimate thermal hydraulic system code TRACE against the Multi-Application Small Light-Water Reactor (MASLWR) Natural Circulation (NC), helical coil Steam Generator (SG), Nuclear Steam Supply System (NSSS) design is a novel effort, and is the topic of the present paper. Specifically, the current work relates to the assessment and validation process of TRACE code against the NC database developed in the OSU-MASLWR test facility. This facility was constructed at Oregon State University under a U.S. Department of Energy grant in order to examine the NC phenomena of importance to the MASLWR reactor design, which includes an integrated helical coil SG. Test series have been conducted at this facility in order to assess the behavior of the MASLWR concept in both normal and transient operation and to assess the passive safety systems under transient conditions. In particular the test OSU-MASLWR-002 investigated the primary system flow rates and secondary side steam superheat, used to control the facility, for a variety of core power levels and Feed Water (FW) flow rates. This paper illustrates a preliminary analysis, performed by TRACE code, aiming at the evaluation of the code capability in predicting NC phenomena and heat exchange from primary to secondary side by helical SG in superheated condition and to evaluate the fidelity of various methods to model the OSU-MASLWR SG in TRACE. The analyses of the calculated data show that the phenomena of interest of the OSU-MASLWR-002 test are predicted by the code and that one of the reasons of the instability of the superheat condition of the fluid at the outlet of the SG is the equivalent SG model used to simulate the different group of helical coils. The SNAP animation model capability is used to show a direct visualization of selected calculated data.