Probabilistic Approach of Coastal Defense Against Typhoon Attacks for Nuclear Power Plant

With the global warming and sea level rising, it is widely recognized that there is an increasing tendency of typhoon occurrence frequency and intensity. The defenses code against typhoon attacks for nuclear power plant should be calibrated because of the increasing threat of typhoon disaster and severe consequences. This paper discusses the probabilistic approach of definitions about ＂probable maximum typhoon＂ and ＂probable maximum storm surge＂ in nuclear safety regulations of China and has made some design code calibrations by use of a newly proposed Double Layer Nested Mtdti-objective Probability Model （DLNMPM）.

Transient Analysis of Steam Generator in PWR Nuclear Power Plant

The water level control system of steam generator in a pressurized water reactor of nuchear power plant plays an important role which effects the water level control of the steam generator are due the reverse dynamics behavior,so the transient analysis of the steam generator should firstly solve their mathematical models.For determination of dynamic behavior and design and testing of the control system, a nonlinear math model is developed using one dimensional conservation equations of mass,momentum and energy of primary and secondary sides of the steam generator. The nonlinear model is verified with standard power plant data available in the references, then the steady states and transient calculations are performed for full power to 5% power reactor operation of the steam generator of Chinese Qinshan Nuclear Power Plant.

Atmospheric Dispersion and Deposition of Radionuclides (¹³⁷Cs and ¹³¹I) Released from the Fukushima Dai-ichi Nuclear Power Plant

The Lagrangian Particle Dispersion Model (LPDM) in the 594 km× 594 km model domain with the horizontal grid scale of 3 km×3 km centered at a power plant and the Eulerian Transport Model (ETM) modified from the Asian Dust Aerosol Model 2 (ADAM2) in the domain of 70° LAT × 140° LON with the horizontal grid scale of 27 km×27 km have been developed. These models have been implemented to simulate the concentration and deposition of radionuclides (137Cs and 131I) released from the accident of the Fukushima Dai-ichi nuclear power plant. It is found that both models are able to simulate quite reasonably the observed concentrations of 137Cs and 131I near the power plant. However, the LPDM model is more useful for the estimation of concentration near the power plant site in details whereas the ETM model is good for the long-range transport processes of the radionuclide plume. The estimated maximum mean surface concentration, column integrated mean concentration and the total deposition (wet+dry) by LPDM for the period from 12 March to 30 April 2011 are, respectively found to be 2.975 × 102 Bq m-3, 3.7 × 107 Bq m-2, and 1.78 × 1014 Bq m-2 for 137Cs and 1.96 × 104 Bq m-3, 2.24 × 109 Bq m-2 and 5.96 × 1014 Bq m-2 for 131I. The radionuclide plumes released from the accident power plant are found to spread wide regions not only the whole model domain of downwind regions but the upwind regions of Russia, Mongolia, Korea, eastern China, Philippines and Vietnam within the analysis period.

Tsunami Impacts on Nuclear Power Plants along Western Coast of India Due to a Great Makran Earthquake: A Numerical Simulation Approach

The major seismicity source in the northern Arabian Sea is the Makran Subduction Zone (MSZ) that defines the tectonic boundary between the Arabian plate and the Eurasian plate, located offshore Iran and Pakistan over which an instrumentally registered earthquake (Mw 8.1) generated a tsunami on 27 November, 1945. It has caused severe cataclysm to a vulnerable population along the surrounding coastlines, including India. It has been on a long seismic quiescence since this last event. The population and industrialization have exponentially increased along the coastal areas in last half decade. The highly exposed coastal locations to the tsunamis are the areas where the nuclear power plants are located. In the present work, a numerical simulation of a great tsunamigenic earthquake (M 9) is presented that predicts the generation, propagation, run-up and travel time using TUNAMI N2 for estimating tsunami impacts along the nuclear power plants of the western coast of India. TUNAMI N2 code was designed for shallow water wave equations, which uses the finite-difference method based on staggered-leap frog scheme. Thus, it has potential to simulate a far-field tsunami with much more accuracy than other methods. It is observed that the tsunami will strike along the coast of Jaitapur Nuclear Power Plant (Maharashtra), Tarapur Nuclear Power Plant (Maharashtra), Kaiga Nuclear Power Plant (Karnataka) and Mithi-Virdi Nuclear Power Plant (Gujarat) after 210, 215, 225 and 230 minutes, respectively. Results show that the tsunami run-up is highest for Jaitapur coast (2.32 m). The Mithi-Virdi coast is the least effected (0.93 m) while Kaiga (2.15 m) and Tarapur coast (2.12 m) might have faced quite intense tsunami consequences. The arrival times and run-ups of the tsunami along the coast of different power plants have been calculated since these parameters are of vital importance in mitigation of the coastal hazard, evacuation planning and installation of early warning system in order to save the inhabited communities from the disaster.

Systematic Approach to Training for Design of Nuclear Power Plant Decommissioning Training in South Korea

In 1979, unavailability of MFWS （Main Feedwater System） in TMI （Three Mile Island） Unit-2 happened in the United States. To make it worse, due to malfunction of Isolation Control Valves in AFWS （Auxiliary Feedwater System）, the supply of cooling water to SGs （Steam Generators） was delayed approximately 8 minutes compared to AOP （Abnormal Operating Procedure）. In the long run, on account of deferred heat sink provision to SGs, the reactor core was melted partially. It was the first critical accident in the US commercial NPP （Nuclear Power Plant） history. Hence, after TMI Accident, US NRC （Nuclear Regulatory Committee） suggested more than one hundred alternatives to improve safety and reliability of NPP. Among these countermeasures, one proposal was related to training area. It was SAT （Systematic Approach to Training） methodology. Therefore, the goal of SAT is the enhancement of NPP stability through training point of view. Since the appearance of SAT in the nuclear industry, it has acquired the unwavering position in the US NPP training field. Meanwhile, significance of NPP decommissioning has been soared up in South Korea since the announcement of Kori Unit-1 decommissioning decision. According to the proclaimed plan from Korean government, Kori Unit-1 is scheduled to be decommissioned from June, 2017. Under this circumstance, nurturing sufficient number of NPP decommissioning engineers is one of the most urgent issues in South Korean nuclear industry. Hence, to elevate efficiency and consistency of training quality, SAT methodology can be a reliable solution for the training of decommissioning engineers. For this reason, the present study aims to design SAT based NPP decommissioning engineer training in South Korea.

Twin model-based fault detection and tolerance approach for in-core self-powered neutron detectors

The in-core self-powered neutron detector(SPND)acts as a key measuring device for the monitoring of parameters and evaluation of the operating conditions of nuclear reactors.Prompt detection and tolerance of faulty SPNDs are indispensable for reliable reactor management.To completely extract the correlated state information of SPNDs,we constructed a twin model based on a generalized regression neural network(GRNN)that represents the common relationships among overall signals.Faulty SPNDs were determined because of the functional concordance of the twin model and real monitoring sys-tems,which calculated the error probability distribution between the model outputs and real values.Fault detection follows a tolerance phase to reinforce the stability of the twin model in the case of massive failures.A weighted K-nearest neighbor model was employed to reasonably reconstruct the values of the faulty signals and guarantee data purity.The experimental evaluation of the proposed method showed promising results,with excellent output consistency and high detection accuracy for both single-and multiple-point faulty SPNDs.For unexpected excessive failures,the proposed tolerance approach can efficiently repair fault behaviors and enhance the prediction performance of the twin model.

Violation of the Convention on the Protection of the Black Sea against Pollution,Bucharest’s Convention of 1992 and the Sofia Protocol of 2018 by Turkish Government’s Sinop Nuclear Power Complex Project

On September 12th,2020,the Turkish Ministry of Environment and Urban Planning issued a final approved version of an EIA(Environmental Impact Assessment)report for the Sinop(4,800 MWe)nuclear power plant and nuclear fuel fabrication complex project,located on Turkey’s Southern Black Sea coastline,in the Sinop Providence,which will be built on a BOO(Build-Own-Operate)basis by an offshore company known as General Directorate of Electricity Generation Inc.(EUAS)International ICC,along with unknown shadowy partners and investors.This project violates the Convention on the Protection of the Black Sea against Pollution,namely,Bucharest’s Convention of 1992,and the Sofia Protocol of 2018 which are established to preserve the uniqueness of the Black Sea,sustain the fisheries,and protect marine life.The Black Sea is the largest anoxic water basin in the world with oxygen rich surface waters supporting marine life which constitute only about 13% of the Black Sea volume.For the rest of the entire Black Sea at a depth greater than 150-200 m,there is a permanent hydrogen sulphide zone devoid of life,the oxygen is completely absent after this level.This unique bio-hydrological characteristic has been regulating the preconditions of its following distinctive biodiversity for thousands of years.The biologically rich regions are only limited to only oxygen rich shelf zones,with depths of up to 50-100 m in the southern coast line,and in the northern Black Sea shallow-water areas with depths of up to 5-10 m.This water body,bordering the hydrogen sulphide zone,is approximately 200-300 m wide and averages 5-50 m deep,in which high concentrations of fish eggs and larvae strive,and circulate counterclockwise along 4340 km coastline of the Black Sea.

基于SEM-AHP方法的核电厂定量人为因素评价研究

文章基于人为因素分析与分类系统,分析了115起核电厂执照运行事件报告(Licensed Operation Event Report,LOER),并对影响核电厂事件发生的人为因素进行分类,采用结构方程模型方法得出各影响因素的路径系数,两两比较得到层次分析模型的判断矩阵,进而确定各因素的权重。同时对核电厂人为因素的重要性做排序分析,找到核电厂运行过程中的安全管理薄弱环节,从而促进核电厂系统平稳运行。

基于FMEA分析和可靠性建模计算的核电厂楔形闸阀预防性解体维修策略优化研究

以核电厂某核岛系统楔形闸阀为例,结合其使用环境、结构特点和运维检修数据分析,基于FMEA分析、风险管控评估和可靠性建模计算,对其预防性解体维修周期进行了优化评估。研究结果表明:该楔形闸阀具备解体周期延长条件,阀门解体周期延长对缩减核电厂不必要的维修资源投入、降低维修辐射剂量,以及缩短大修关键路径有直接贡献。

核电厂海水循环泵肘形进水流道优化设计

为提高核电厂海水循环泵运行的稳定性,基于CFD数值模拟对海水循环泵肘形进水流道进行优化设计。以国内某核电厂的进水流道为研究对象,建立全新坐标系,通过理论推导得出肘形流道控制参数之间的函数关系,根据流道性能评判指标等参数,通过数值模拟研究肘形进水流道4个重要控制参数对流道水力性能的影响,优选出最佳方案并进行试验验证。结果表明,在设计流量下优化后的流道出口平均流速提高了0.026 m/s,流速分布均匀度降低0.13%,水流平均角度提高0.034°,水力损失降低11.2%。本文的研究结果显示提出的肘形流道优化设计方法快速、可靠,具有出口流态较好、水力损失较小等优点,可为工程建立提供一定的参考。

基于遗传算法的核电备件库存模型仿真优化

备品备件是核电厂设备检修的物质基础,然而持有过量的备件库存将增加核电厂运营成本,通过设置合理的备件库存参数,在保障供应和控制库存之间寻找到一个合理的平衡点。针对核电备件库存参数设置问题,将遗传算法引入备件库存参数优化领域。通过建立核电备件库存参数评价方法,将核电备件库存参数设置转换为多变量优化问题,使用遗传算法计算最优库存参数。选用大亚湾核电厂的备件进行仿真测试,当给备件增加多种领用波动时,使用遗传算法计算的备件库存参数均优于当前数据库中的库存参数,结果表明遗传算法应用到备件库存参数优化领域具有重要的实用价值。

高温高压水管道热疲劳试验设计与数值模拟

核电厂压力容器启停堆和正常运行时,温度压力会在一定范围内发生波动,随着波动次数的增加,会对设备和管道造成严重疲劳损伤。依托北京某高校高温高压腐蚀实验站亚临界水汽环境结构材料试验装置双回路,设计了一种承压管道高通量热疲劳试验方法。基于该试验方法,完成了瞬态水温变化对管道疲劳裂纹扩展的影响研究。结果表明,疲劳裂纹扩展速率随冷却速率降低而逐渐降低。结合有限元数值模拟和三维裂纹扩展分析软件建立了阶梯管道结构的热疲劳裂纹扩展速率模型,并用试验结果对模型进行了修正。

基于霍尔三维结构理论的化学控制计算软件的开发

在核电厂中化学控制计算与机组反应性控制、设备管道腐蚀以及机组运行寿命等方面有着重要联系,通过开发并使用程序化计算软件可有效提高计算准确率、缩短工时以及降低人因失误,进一步提升化学精细化控制水平。根据霍尔三维结构模型,从逻辑维、时间维和知识维角度分析软件开发过程,将化学科学专业知识与软件开发和系统管理相融合,形成一套完整的软件开发方法。软件开发大大提升了化学控制智能化水平,有效保证机组长期、稳定、安全地运行。

核电厂高压蒸汽管道焊缝切割装置的研制

为了修复某核电厂GPV高压缸排汽管膨胀节导流筒焊缝因热处理不当导致的开裂,研发了一款焊缝切割装置,用于将开裂的焊缝切除并加工坡口以进行下一步的自动焊接修复。首先利用SolidWorks对焊缝切割设备进行了机械设计并三维建模,对驱动电机进行了选型计算,设计了控制系统架构,随后制作了设备样机和膨胀节导流筒模拟体,最后在试验室利用模拟体对设备样机的性能进行了连续切削试验,验证了该装备能够高效完成膨胀节导流筒开裂焊缝的切除和坡口加工,且坡口加工质量优良,装置性能稳定。

热核聚变发电厂储能系统运行模式研究

[目的]核聚变作为一种新型核能利用形式,其清洁绿色特性被视为人类未来的“终极能源”,也是我国在能源发展道路上的必经之路。在热核聚变发电厂中,中国聚变工程实验堆(CFETR)聚变反应堆具有一定周期性、脉冲式的输出特性,而汽轮发电机组的运行是稳定、连续的,因此核岛和常规岛之间需配有储能岛以进行储能缓冲;为实现其储能缓冲的功能,储能系统需选择合适的运行模式。[方法]核岛与常规岛耦合运行或解耦运行,对应着储能系统不同的运行模式;文章基于CFETR核岛侧的功率输出特性,从系统配置、设备选型和运行控制等方面对储能系统运行模式的不同方案进行分析和对比。[结果]研究结果表明,核岛与常规岛的解耦运行模式在常规岛发电效率、设备设计技术成熟性和机组运行控制上都更具优势,因此推荐热核聚变发电厂储能系统采用解耦运行模式。[结论]储能系统解耦的运行模式方案大都采用成熟技术,可进行规模化的商业应用,因此可为实现热核聚变发电厂的商业化设计提供支撑。

核电站常用管道材料J-C本构模型参数识别及验证

针对核电站管道在受到地震等灾害产生的偶然冲击载荷作用时的应力流动与变形情况,采用电子万能试验机和高速拉伸试验机,分别对碳钢A106(A-106 GR.B)、不锈钢304L(A-312 TP304L)、合金钢A335(A-335 GR.P11)3种核电站常用的管道材料进行准静态拉伸与动态拉伸试验,并根据试验所得结果拟合建立了3种材料的Johnson-Cook(J-C)本构模型。利用Abaqus有限元分析软件建立管道的冲击模型,并通过冲击试验台架对仿真结果进行验证。结果表明,3种材料都出现了明显的应变率强化效应,A335的抗冲击能力更好一些,304L与A106抗冲击能力差别不大。试验结果与仿真结果吻合度良好,拟合得到的J-C本构模型参数可靠性较高,能够准确地描述材料在冲击载荷下的应力流动与变形行为,对核电站常用管道受冲击载荷作用仿真分析研究具有十分重要的意义。

核电厂核岛消防应急照明与疏散指示系统设计

对核电厂核岛应急照明设计特点进行了介绍,包括应急照明分类与运行方式、系统供电、照明控制与灯具选型等。从“系统型式、设备选型、系统配电”等方面指出当前核电厂核岛消防应急照明与疏散指示系统与新标准规范的差异和不足。结合核环境A型消防应急照明灯市场供应情况,给出近期采用“自带电源非集中控制型系统”和远期采用“集中电源集中控制型系统”的设计建议。

基于MBSE的核电厂应急预案系统的顶层需求分析与接口设计研究

在核电领域,任何意外情况都可能引发严重的后果,因此,制定科学、有效的应急预案系统是保障核电安全不可或缺的一环。作为一种先进的系统工程方法,基于模型的系统工程(MBSE)以其系统化、模型化、可视化的优势,为核电应急预案系统的设计与实施提供了新的思路和方法。探讨了MBSE在核电厂应急预案系统设计中的初步应用,分析了核电应急体系的关键要素以确定其与MBSE方法论的契合度,利用MBSE方法论进行了顶层需求的承接。针对应急体系接口问题,构建了核电应急体系各组成部分之间的交互模型。结果表明,MBSE可以提高核电厂应急预案系统设计便利性和开发效率。

核电厂动力机器模型基础动力参数测试的优化

核电厂汽轮机厂房多位于坚硬的岩石地基上,模型基础动力参数测试共振频率达到100 Hz以上,土质地基上常用的测试方法由于其机械式激振器扰频上限低的缺陷,对核电厂岩石地基适用性很差。在激振器、垫层、激振器与模型基础的连接方式、测试方式等方面进行研究,优化、改进了测试设备和测试方法,开发了全自动常扰力测试控制系统,使之更适合核电厂岩石地基,测试结果更准确地代表岩石地基的动力特性。

核电厂地下管网结构化数据建模研究

在数字化设计平台建设过程中,为了解决缺少专业设计软件的情况下少量地下管网模型的生产和集成问题,需要建立一种与设计软件解绑的建模方式。通过研究使用结构化的管网数据驱动生成三维模型的方法,实现核电厂已有地下管道和管道附属物快速逆向建模,进一步实现核电厂室外工程设计模型和施工模型的集成与应用。