共查询到18条相似文献,搜索用时 171 毫秒
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RELAP5程序本身具有模拟核电站控制与保护系统的功能,但是,由于该程序采用文本输入方式进行建模,编写复杂,可读性不强,小适合于对大型复杂控制系统进行仿真。而Simulink程序采用图形化建模方式,能够高效、便捷地对核电厂复杂控制与保护系统进行建模。因此,本文将RELAP5程序与Simulink耦合,并利用Simulink扩展RELAP5的控制与保护系统的模拟功能。为了验证两程序耦合方法的准确性,将用Simulink实现的控制与保护系统的仿真结果,与已通过验证的RELAP5实现的具有相同功能的控制与保护系统的仿真结果进行对比,结果表明二者符合较好。 相似文献
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大亚湾核电站核岛模拟控制系统采用美国Bailey9020模拟平台,计划在30 a大修期间进行数字化升级改造。本项目将SpeedHold-N(SH-N)系统、Bailey9020模拟平台、大亚湾核电站工艺仿真模型通过可编程控制器(PLC)接口搭建最小验证平台,选取稳压器压力控制回路为验证对象,以调试程序为参考,通过对比相同扰动在不同平台的响应差异,验证了大亚湾核电站当前控制参数在目标数字化控制系统(DCS)平台的适用性,发现了Bailey9020模拟平台比例积分(PI)控制器在手自动切换时有比例作用无法响应的固有缺陷,并分别从电路设计和操作程序两方面提出了相应优化措施,达到了模拟平台解析、DCS组态验证、定值转换验证的目的,同时为当前大亚湾核电站机组运行提出了重要反馈。 相似文献
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反应堆瞬态计算程序RELAP5-HD的仿真模型主要采用偏微分方程进行描述,可用于冷却剂温度系统的仿真验证。然而,利用控制理论无法直接对偏微分方程组建立的系统进行稳定性、稳态特性、动态特性分析,从而对冷却剂温度系统的控制器设计缺乏了一种有效的优化手段。为解决上述问题,采用热工水力学第一性原理与空间离散化方法,建立了一套用于分析冷却剂温度系统特性的铅基冷却反应堆热工水力传递函数模型。该模型与RELAP5-HD模型的对比计算结果表明,当控制变量发生阶跃时,传递函数模型与RELAP5-HD模型的输出特性能较好地吻合,准确反映了系统的动力学特性,能够利用控制理论对铅基冷却反应堆冷却剂温度系统的特性进行分析研究。 相似文献
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秦山核电二期工程核仪表系统设计 总被引:2,自引:1,他引:1
对秦山核电二期工程600MW核电站核仪表系统(RPN)的设计、采购和安装调试的基本情况进行分析.秦山核电二期工程RPN的构成和外部接口均参考大亚湾核电站,但系统内部采用了先进的数字化技术.文章首先对系统作简要的描述,然后回顾了系统在初步设计和施工设计阶段的设计,然后介绍了"八五"期间部分设备模拟样机攻关,最后介绍了系统在现场的安装调试期间遇到和解决的一些问题. 相似文献
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RALOC4 code validation against measured data at Ignalina NPP during single Main Safety Valve opening
Although RALOC4 code is validated against many experiments with regard to Western Nuclear Power Plants (NPPs) the code validation problem for the Accident Localization System (ALS) of Ignalina NPP modeling is of special importance because the condensing pools at NPP with RBMK-1500 differ from the pressure suppression systems installed in NPPs with German BWR. The response of Ignalina NPP ALS to the unintentional opening of single Main Safety Valve, which occurred in 1998, is analyzed by employing code RALOC4. The results of post-event calculations compared with the measured data available after the event. The performed analysis showed that RALOC4 code could be applied for the simulation of Ignalina NPP ALS. Nevertheless, the spray modeling in RALOC4 should be improved allowing the simulation of sprays in NON_EQUILIBRIUM zone model and to consider the diameter of water droplet diameter and height of droplet fall. 相似文献
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搞好核电厂安全管理是核电厂安全生产的重要保证.核电厂安全管理包括建立健全安全责任制、安全生产的监督和管理及信息反馈、掌握安全防护规程、用科学态度处理放射性安全工作、保证工业安全及做好安全分析工作。我国核电刚刚起步,学习国外先进的安全规程,吸取他人的经验教训,探讨适合我国国情的安全管理规律是十分必要的. 相似文献
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The analysis of an unintended main safety valve opening at Ignalina NPP was performed with COCOSYS code in order to assess its capability in simulation of the transient processes that occur inside Accident Localisation System of Ignalina NPP. COCOSYS has several user-selected options, e.g. zone model (EQUIL._MOD, NONEQUILIB), water flow model (BAL_DRAIN, DRAIN_BOT), etc for nodalisation development. The influence of a zone model selection, a water overflow model selection and efficiency of heat exchanger in Condenser Tray Cooling System was investigated and presented in the paper. The performed analysis supported introduction of new water overflow model in COCOSYS code and showed that COCOSYS code can be applied for the analysis of Accident Localisation System of Ignalina NPP. 相似文献
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秦山一期核电站未能紧急停堆的预期瞬变导致堆芯熔化的进程及事故缓解措施研究 总被引:1,自引:0,他引:1
选择失去主给水、失去厂外电和正常运行情况下控制棒失控提升3个典型的导致未能紧急停堆的预期瞬变(ATWS)的初因事故,采用自行研制的基于SCDAP/RELAP5/MOD3.1的核反应堆严重事故分析平台,对秦山一期核电站ATWS初因导致堆芯熔化严重事故进程进行了分析研究,对防止ATWS导致堆芯熔化进程的缓解措施的有效性进行了验证。计算分析结果表明,二回路补水和一回路卸压的事故缓解措施能有效地阻止堆芯熔化进程。 相似文献
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Gy. Hegyi A. Keresztúri Cs. Maráczy G. Hordósy E. Temesvári 《Annals of Nuclear Energy》2011,38(2-3):694-704
The reasons for large discrepancies between the computed and measured values of the efficiency of control rods observed during start-up experiments on the Russian pressurized water type VVER reactors are discussed. The numerical simulation of the measurements including the prediction of the ex-core detector signals was used to resolve the discrepancies. The time and space dependent neutron flux in the core during these measurements have been calculated by the KIKO3D nodal kinetic code. For calculating the ionization chamber signals the Green function technique has been applied. The Green functions of ionization chambers have been evaluated via solving the neutron transport equation in the reflector regions with the MCNP Monte Carlo code. The detector signals have been calculated and compared with measured ones using the inverse point kinetics transformation. Large number of asymmetric rod drop measurements (with one rod stuck) and some differential rod worth measurements from the Zero Power Physics Tests were provided by the Paks NPP for validation. The experiments cover different fuels (without and with enrichment zoning) and loading patterns. The intermediate range ionization chambers have been used during the scram measurements. The newly developed method provides fairly sufficient match of measured and calculated results. The time behavior of the detector readings observed in the measurements are described by the code in a consistent manner.As a further application the uncertainty of scram rod worth of the KARATE-440 code system was determined by static calculations and subsequent simulation of rod drop with the KIKO3D code. The calculated results were compared to measurements carried out by the Paks NPP. The uncertainty of scram rod worth is established by statistical analysis. 相似文献
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LinAo Nuclear Power Plant (NPP) Phase II is a newly-built CPR1000 reactor in China, and many new technologies including the incorporation of digital control system (DCS) substituting traditional analog control systems have been applied. This is the first time for Chinese engineers to setup and adjust the DCS configurations. Both the lack of the operating experiences and the plant safety requirements from the government make a necessity of the closed-loop DCS test before commercial plant operation. The most practical way is to build a digital plant as the controlled target and this digital plant is used to provide the plant thermal–hydraulic parameters and feedbacks for the DCS. Though the RELAP5 code has been developed for the best-estimate transient simulation of light water reactor coolant systems and is used worldwide, its functionality is too limited to implement a digital plant, such as the simulation of the complicated plant control and protection systems, the 3-dimensional neutron kinetics and the fluid network for the plant auxiliary systems. To overcome these drawbacks, a RELAP5-based extensible simulator has been built to satisfy the new requirements for the implementation of a digital plant. Any simulation code of desired functionality can be integrated into this simulator as a simulation module once it applies a set of well-defined data exchange interfaces. At the present stage, a RELAP5 module, a control system modeling module, a software–hardware data bridge module and some other auxiliary modules have been integrated into the simulator. There are more than 60 systems that need to be tested with the DCS in LinAo Phase II, and the whole testing work is separated into several phases. In this paper, we take the testing of the pressure control system and water level control system of pressurizer as example. A typical transient of 10% load step change from 100%FP (full power) to 90%FP was performed for the closed-loop DCS test. The necessity and capability of this RELAP5-based engineering simulator has been demonstrated. 相似文献
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应用一体化严重事故分析程序MELCOR1.8.5进行模拟分析,研究了由西屋公司制定、经美国NRC(NuclearRegulatoryCommission)认证的“堆芯损伤评价导则(CDAG)”应用于中国百万千瓦级核电站在严重事故初期评价堆芯损伤状态和程度的有效性。初步分析结果表明,CDAG可较好地评价百万千瓦级核电站无缓解措施的冷却剂丧失事故(LOCA)堆芯损伤状况和损伤程度,对进一步研究和验证CDAG的综合评价能力和适用性、推进现有核电厂建立严重事故管理导则具有重要的参考价值。 相似文献
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反应堆控制系统是核电厂重要仪控系统之一,对保障核电厂的正常运行起着重要作用。为确保控制系统在核电厂运行过程中的良好控制品质和减少现场调试时间,有必要在设计阶段通过仿真研究对控制系统参数进行优化设计。分析了三代核电华龙一号(HPR 1000)海外首堆的反应堆控制系统功能,对各控制系统被控变量进行了说明;在此基础上,对控制系统参数优化流程进行说明;利用核电厂数字化仿真工具,通过系统建模仿真对控制系统参数进行敏感性分析,根据不同参数取值下的系统静态和动态响应特性得到较优的控制系统参数,经性能验证满足设计要求。所获得的反应堆控制系统参数已用于海外华龙一号首堆反应堆控制系统设计,并用于指导核电厂现场调试和核电厂运行。 相似文献