核事故应急堆芯损伤快速评价系统顶层设计方案研究

通过对国内外堆芯损伤评价方法的详细调研,提出适用于我国目前运行及在建核电厂的堆芯损伤评价方法,即堆芯损伤评价导则(CDAG)和国际原子能机构第955号技术报告(IAEA TECDOC-955)相结合的方法,并给出详细的系统顶层设计方案,为我国核事故应急堆芯损伤快速评价系统顶层设计的最终制定提供有利依据。     

核电厂堆芯损伤评价研究及软件开发

核电厂发生事故后,需要及时准确地判断反应堆堆芯损伤状态,以便为应急决策提供必要的技术支持。基于国际上堆芯损伤评价方法研究现状,重点介绍适用于我国在建和运行压水堆核电厂的堆芯损伤评价方法,并开发堆芯损伤评价软件,从而有效支持核电厂的应急决策,进一步提高核电厂的安全水平。     

压水堆堆芯换料设计优化的研究

讨论压水堆堆芯换料设计优化问题，并研制一套实用的换料优化软件包，可用于低泄漏、外－内和改进的外－内装料方案的优化。在低泄漏方案优化时，利用哈林原理将燃料组件布置与可燃毒物配置的优化问题脱耦成两步优化问题。先用线性规划方法进行无可燃毒物时燃料组件布置的优化，然后再用可变容差法寻找可燃毒物的优化布置。应用该软件对秦山核电厂首次难芯换料方案进行了优化，提出了可供参考的一些优化布置方案。     

压水堆堆芯辐射换热因子研究

严重事故中，反应堆堆芯裸露后，堆芯燃料棒的辐射热被周围热构件的吸收程度对堆芯熔融过程和下封头失效时间的影响很大。辐射换热效果与具体辐射表面形状有关，常用辐射换热因子来衡量。首先对辐射换热因子进行了推导，然后根据计算结果，使用MELCOR程序比较了一种压力水堆LOCA事故下计算值和缺省值对事故进程的影响，差异比较显著。     

基于多模型策略的压水堆堆芯功率控制与仿真

在建立压水堆堆芯非线性模型的基础上求取5个功率水平处的线性化模型作为堆芯局部模型,以局部模型组合替代堆芯非线性模型。利用基于局部模型全维观测器的全状态反馈法,设计带有鲁棒性能的局部模型控制器作为非线性堆芯局部控制器,用于相应功率水平域内的功率控制。仿真结果表明,所设计的堆芯多模型控制系统能很好地控制堆芯功率。     

压水堆堆芯设计特点及其演变

综合论述了压水堆堆芯设计中的化学补偿反应性、标准化无盒大型燃料组件、棒束型控制棒、可燃毒物和采用多区堆芯装料等基本问题。并以上述５大问题为基础,简要叙述了负荷跟踪运行给堆芯设计带来的有关设计问题。此外,简要介绍了当前压水堆堆芯的改进设计及演变过程。     

压力容器水位参数在堆芯损伤评价方法中的应用

事故工况下,堆芯会随着冷却能力的下降而逐步升温,长时间的裸露会导致堆芯损伤,而堆芯出口温度和压力容器水位可直观反映堆芯的冷却能力。以西屋公司堆芯损伤评价导则为基础的堆芯损伤评价方法将堆芯出口温度和安全壳剂量率作为主要参数评价堆芯损伤状态,压力容器水位作为辅助参数之一来验证评价结果的合理性,但一些核电厂堆芯出口热电偶量程并不能满足严重事故条件下的要求,需要其他替代参数。本工作以压水堆核电厂严重事故分析数据为基础,探讨将压力容器水位作为主要参数应用于堆芯损伤评价方法的可行性。     

华龙一号应急工况评价系统初步开发

为了满足华龙一号(HPR1000)事故条件下的应急响应,需要开发一套应急工况评价系统,用于基于征兆的堆芯损伤评价和释放源项估算。本文给出了华龙一号应急工况评价系统(ECAS-HPR1000)的总体设计,包括软件框架、评价模块、平台和接口开发等,该系统采用跨平台的JAVA语言开发,以MySQL数据库作为数据存储,支持Windows和Linux操作系统。该系统包括五个子系统,分别是基础数据采集和管理子系统、堆芯损伤评价子系统、释放源项计算子系统、评价结果展示子系统和用户权限管理子系统。该系统可以基于实时工况数据,评价堆芯损伤状态和程度,并计算出堆芯释放到一回路、安全壳和环境的放射性核素的量,并考虑了华龙一号双层安全壳对计算结果的影响。     

Design features and thermal margin assessment of core protection and monitoring systems of an integral reactor,SMART

The Korea Atomic Energy Research Institute has developed the SMART integral reactor, and SCOPS and SCOMS were also newly developed as advanced real-time core protection and monitoring systems for SMART. SCOPS calculates the minimum DNBR and maximum LPD based on several on-line measured core state parameters, and SCOMS calculates the limiting conditions for operation variables and assists the operator in implementing the technical specification requirements for monitoring. The design features and characteristics of SCOPS and SCOMS were described. The performance of the SCOMS power distribution synthesis method was evaluated and shows negligible power distribution synthesis errors. A technically reliable uncertainty analysis method was developed, and a preliminary uncertainty analysis was evaluated. The overall analysis results are similar or more improved compared to those of cycle 1 for Younggwang units 3&4 of Korea. In particular, uncertainty factors of SCOMS are much improved because of an improvement in the power distribution synthesis and DNBR calculation algorithm. Finally, thermal margins were estimated, and the DNB overpower margin of SCOMS is large enough to accommodate a 40% required overpower margin and 15% top-tier requirement thermal margin.     

铀浓缩设施核应急实时评价系统开发

本文介绍了我国首次集成开发的铀浓缩设施核应急实时评价系统。系统针对铀浓缩设施可以实时评价核临界事故和UF₆泄漏事故的辐射影响。系统可根据事故γ报警仪剂量率读数估算核临界裂变次数,并自动评价核临界事故后果。针对UF₆烟羽的重气特性,将重气模型与高斯烟羽模型相结合,可更准确地模拟UF₆的扩散过程。系统的开发解决了铀浓缩设施应急准备和响应一直缺乏的技术支持能力。     

Core cooling in pressurized-water reactor during water injection

In this paper,the reactor core cooling and its melt progression terminating is evaluated,and the initiation criterion for reactor cavity flooding during water injection is determined.The core cooling in pressurized-water reactor of severe accident is simulated with the thermal hydraulic and severe accident code of SCDAP/RELAP5.The results show that the core melt progression is terminated by water injection,before the core debris has formed at bottom of core,and the initiation of reactor cavity flooding is indicated by the core exit temperature.     

Identification of initiating events for pressurized water reactor accidents

In order to aid operators in identifying the different initiating events as defined in the Final Safety Analysis Report (FSAR), we develop a novel identification procedure. The procedure is based on the monitoring of three key system parameters in a pressurized water reactor (PWR), i.e., the pressure, the average temperature, and the temperature difference of the hot-leg and cold-leg of the reactor coolant system. By monitoring the system thermal state diagram in a pressure–temperature space, an operator can easily identify what initiating event is taking place while a static point in the diagram starts to move. The event data pool is first established by storing the transient analysis results for events of different types using the optimal estimated RELAP5 model. Since the variation ranges of system key parameters at a specific time represent the specific character for each initiating event, the identification procedure can easily determine which cases in which the event data pool can be fitted to on-line data using only variation range comparison without complex calculations. This identification method is believed to be able to help the plant operator to identify the different events and then execute the Emergency Operating Procedure more effectively.     

中国实验快堆堆芯围桶开孔对正常运行工况影响的分析

堆芯围桶开孔是中国实验快堆(CEFR)事故余热排出系统的重要组成部分之一,是保证该系统形成自然循环排出反应堆事故后剩余发热的关键环节。本文应用通用计算流体力学软件CFX对CEFR堆芯围桶开孔对反应堆正常运行工况的影响进行了模拟,计算了在正常工况运行时,CEFR的反射组件与屏蔽组件热功率对堆芯围桶开孔附近温度场以及流场的影响,给出了堆芯围桶开孔区域的三维温度场、三维流场以及压力分布矢量图。结果表明,目前的设计在满足事故余热排出的要求同时,对反应堆正常运行工况的影响是可以接受的。     

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.     

Development of a short-term emergency assessment system of the marine environmental radioactivity around Japan*

The Japan Atomic Energy Agency (JAEA) has, for many years, been developing a radionuclide dispersion model for the ocean, and has validated the model through application in many sea areas using oceanic flow fields calculated by the oceanic circulation model. The Fukushima Dai-ichi Nuclear Power Station accident caused marine pollution by artificial radioactive materials to the North Pacific, especially to coastal waters northeast of mainland Japan. In order to investigate the migration of radionuclides in the ocean caused by this severe accident, studies using marine dispersion simulations have been carried out by JAEA. Based on these as well as the previous studies, JAEA has developed the Short-Term Emergency Assessment system of Marine Environmental Radioactivity (STEAMER) to immediately predict the radionuclide concentration around Japan in case of a nuclear accident. Coupling the STEAMER with the emergency atmospheric dispersion prediction system, such as Worldwide version of System for Prediction of Environmental Emergency Dose Information version II (WSPEEDI-II), enables comprehensive environmental pollution prediction both in air and the ocean.     

核与辐射应急准备分类及相应的管理要求

根据国际经验给出了基于危害评价的应急准备分类以及各类应急准备需要建立的基本能力,提出了应急管理行动、应急状态、应急响应的启动、缓解行动、应急评价和预测、防护行动和其他响应行动、公众应急通知、应急人员和应急援助人员的保护措施、医学响应行动、应急组织和人员配置、应急信息发布和公众沟通等11个主要要素对应5个应急准备类型的管理要求建议。