Increasing the Efficiency and Level of Environmental Safety of Pro-Environmental City Heat Supply Technologies by Low Power Nuclear Plants

In connection with the current prospect of decarbonization of coal energy through the use of small nuclear power plants (SNPPs) at existing TPPs as heat sources for heat supply to municipal heating networks, there is a technological need to improve heat supply schemes to increase their environmental friendliness and efficiency. The paper proves the feasibility of using the heat-feeding mode of ASHPs for urban heat supply by heating the network water with steam taken from the turbine. The ratio of electric and thermal power of a “nuclear” combined heat and power plant is given. The advantage of using a heat pump, which provides twice as much electrical power with the same heat output, is established. Taking into account that heat in these modes is supplied with different potential, the energy efficiency was used to compare these options. To increase the heat supply capacity, a scheme with the use of a high-pressure heater in the backpressure mode and with the heating of network water with hot steam was proposed. Heat supply from ASHPs is efficient and environmentally friendly even in the case of significant remoteness of heat consumers.

Study on integrated development and hybrid operation mode of nuclear power plant and pumped-storage power station

The nuclear power plant is suitable for base-load operation, while the pumped-storage unit mainly gives play to capacity benefit in the electric power system;hence, the integrated development and hybrid operation mode of the two can better meet the needs of the electric power system. This article first presents an analysis of the necessity and superiority of such mode, then explains its meaning and analyzes the working routes. Finally, it proposes the business modes as follows: low price pumping water electricity plus nuclear power in the near term;nuclear power shifted to pumped storage power participating in market competition in the middle term;and, in the long term, nuclear power shifted to pumped storage power as primary and serving as an electric power system when needed.

Variants of Nuclear Power Plants of Small and Medium Power with Heavy Liquid-Metal Coolants

New design solutions have been proposed for a BRS-GPG type reactor circuit, which are different from transport and stationary low and medium-powered reactor installations cooled with heavy liquid-metal coolants, and which correspond to the evolutionary development of such installations. While developing these solutions, the available experience in creating and operating Soviet pilot and commercial power plants cooled with lead-bismuth coolants was used, including investigations, primarily experimental ones, carried out by team of authors in justification of a capacity range (50 - 250 MW) of low and medium-powered reactor plants with horizontal steam generators (BRS- GPG) proposed and elaborated at the NNSTU.

Present status of pumped hydro storage operations to mitigate renewable energy fluctuations in Japan

This paper focuses on pumped hydro energy storage(PHES)plants’current operations after electricity system reforms and variable renewable energy(VRE)installations in Japan.PHES plants have historically been developed to create electricity demand at night in order to operate base load power plants,such as nuclear power plants,in stable conditions.Therefore,many PHES plants are located midway between nuclear power plants and large demand areas.However,all nuclear power plants had to–at least temporarily–shut down after the Great East Japan Earthquake followed by a nuclear accident at Fukushima Daiichi in 2011,and renewable energy power plants have been deployed rapidly after the introduction of a feed-in-tariff(FIT)scheme.Therefore,PHES plants are being used to mitigate fluctuations of VRE,especially in areas where renewable energy has been significantly installed.The daily highest capacity ratio of PHES plants in Kyushu area has recorded three times higher than it in the other areas where the past operating mode is still conducted.But those operations on PHES plants are simply followed as a dispatch rule of the Organization for Crossregional Coordination of Transmission Operators(OCCTO),market-based operations have not been conducted enough yet.The market design shall be changed to harmonize VRE installation and PHES plants’operations are necessary to make the transition from the past operating mode of PHES plants across Japan.

Technological Perspectives for Propulsion on Nuclear Attack Submarines

This work aims to present the historical context in which the current understanding of the phenomenon of the direct contact condensation started to call the scientific society attention. The development of nuclear power plants Light Water Reactors demanded a safe way to collect and treat the water used to cool the reactor. Some characteristics of this water in a high energetic thermodynamic state made it unsuitable to be directly discharged in the atmosphere. Small room relieves tanks were developed to contain this discharge. These tanks were partially fulfilled with water, and the vapor injection created a vapor plume. In the interface of liquid and vapor, the thermal exchange would be increased by the characteristic turbulence of this region.

Employing adaptive fuzzy computing for RCP intelligent control and fault diagnosis

Loss of coolant accident(LOCA),loss of fluid accident(LOFA),and loss of vacuum accident(LOVA)are the most severe accidents that can occur in nuclear power reactors(NPRs).These accidents occur when the reactor loses its cooling media,leading to uncontrolled chain reactions akin to a nuclear bomb.This article is focused on exploring methods to prevent such accidents and ensure that the reactor cooling system remains fully controlled.The reactor coolant pump(RCP)has a pivotal role in facilitating heat exchange between the primary cycle,which is connected to the reactor core,and the secondary cycle associated with the steam generator.Furthermore,the RCP is integral to preventing catastrophic events such as LOCA,LOFA,and LOVA accidents.In this study,we discuss the most critical aspects related to the RCP,specifically focusing on RCP control and RCP fault diagnosis.The AI-based adaptive fuzzy method is used to regulate the RCP’s speed and torque,whereas the neural fault diagnosis system(NFDS)is implemented for alarm signaling and fault diagnosis in nuclear reactors.To address the limitations of linguistic and statistical intelligence approaches,an integration of the statistical approach with fuzzy logic has been proposed.This integrated system leverages the strengths of both methods.Adaptive fuzzy control was applied to the VVER 1200 NPR-RCP induction motor,and the NFDS was implemented on the Kori-2 NPR-RCP.

基于数据驱动的核电站主给水泵故障预测研究

目前,核电站设备传感器每天都会产生大量的监测数据,但这些数据的利用程度较低,对于利用监测数据进行核电站重要设备的故障预测研究还处于探索阶段。针对这种情况,本文以核电站主给水泵作为研究对象,将表征主给水泵运行状态的各类监测数据进行预处理和降维,进而通过多个选定的机器学习模型预测设备在未来是否会发生故障。通过对预测模型的效果评估发现,长短时记忆网络模型(LSTM模型)具有较好的预测精度。当模型预测结果超过阈值时发出预警信息,提醒核电站运维人员加强关注,及时采取故障诊断和维修措施,以有效防止因设备的突然故障停运造成较为严重的后果,保证核电站的安全和经济运行。

核电站冷凝水流量功能验证工具开发及试验研究

核电站泄漏监测系统在预防核电站泄漏方面起着重要作用,因此明确泄漏监测系统冷凝水流量测量机理具有重要意义。为了实现冷凝水流量测量机理研究,开发了一套冷凝水流量模拟验证装置,并将该装置成功应用于核电站冷凝水流量测量功能的试验研究。使用该装置对重力疏排水冷凝水流量测量管线的管道布置、管道疏水情况和冷凝水流量计相关参数设置等方面进行验证。同时,对现有设计、安装和调试环节存在的问题给出改进方案。研究结果表明:管线布置不当会导致管道疏水不畅和重力疏水在流量计处无法建立稳定流量等问题。同样地,冷凝水流量计报警阈值设置不当和小流量参数设置不当也会影响系统流量测量功能。该研究可以为重力疏排水流量测量方面的研究和应用提供参考。

核电厂非放射性生产废水处理方案研究

核电厂非放射性生产废水的来源及成分比较复杂,现有核电机组非放射性废水均是排入循环冷却水排水系统后经电厂总排放口排入海中,总排放口排水水质满足《污水综合排放标准》中的排放要求。随着国内环保要求的提高,部分地方环保部门提出废水中的二类污染物含量在排入循环冷却水排水前需满足《污水综合排放标准》中的排放要求,现有废水水质无法满足该要求。为解决上述问题,以CAP1000核电机组为例,通过对非放射性生产废水的组成及现有的处理情况进行分析,明确了超标废水来源主要为凝结水精处理系统树脂再生时排放的酸碱废水,废水中氨氮含量远高于排放标准要求。根据该废水的排放情况及水质特点,先对废水进行了高、低盐分类收集减量处理,然后针对氨氮超标的高盐废水提出了折点氯化法、吹脱+电解制氯氧化联合处理、膜脱氨3种处理方案,并对各方案进行了技术经济比较,为核电厂的非放射性生产废水处理的工艺设计提供参考。

核电厂汽动辅助给水泵转速调节系统鲁棒性提升研究

某核电厂二期工程采用了国产化核级汽动辅助给水泵,调试过程中屡次发生给水泵启动时间超标、启动后转速无法稳定、启动过程中超速跳闸事件。对汽动辅助给水泵调速系统进行了建模及分析,研究了电子调速器、电气转换器、气动定位器、气动增速器等部件的工作特性,分析并通过试验比较了不同流量系数的气动增速器对气动执行机构动态响应特性的影响。采用更换小流量系数气动增速器和调速器参数优化的方法,提升了汽动辅助给水泵调速系统的鲁棒性,解决了给水泵启动失败的问题。该研究的过程和成果可作为国产核级汽动辅助给水泵调速系统的设计参考,对于气动执行机构有关的控制系统参数优化、鲁棒性提升等其他同类问题的处理也有重要的指导意义。

核电厂主给水泵仪表可靠性提升措施

为进一步提升核电厂主给水泵仪表的运行可靠性,以某核电厂为例,指出主给水泵仪表的运行缺陷,包括异常波动、密封失效、元件老化、计量偏差、备用故障报警逻辑问题,详细阐述了可靠性提升措施,并对提升效果及未来改进方向进行了总结。

核电厂冷却水泵叶轮裂纹缺陷的成因及解决方法

针对冷却水泵叶轮出现裂纹缺陷的问题,以某工厂生产冷却水泵为例,从材料铸造工艺、力学性能、裂纹形貌等方面对裂纹缺陷产生的原因进行了分析,并对比了两种冷却水泵叶轮铸造工艺的区别。研究结果表明,该工厂生产叶轮的力学性能较差,在裂纹周围有沿晶面断裂特征,且其金属结构不均匀,造成叶轮裂纹缺陷的主要原因为铸造工艺存在问题。针对工艺方面存在的问题,分别从铸造工艺和热处理工艺两个方面提出改进方案,以期为相关人员提供参考。

A New Algorithm for a Condenser Design for Large-Scale Nuclear Power Plants in Tropical Region

This work presents a new velocity search algorithm for designing a condenser of a 1200 MWe large-scale nuclear power plant situated in tropical region.For this,the condenser pressure was considered in the range of 7.5-15 kPa while its tube inner diameter was taken as 28 mm with 1 mm tube wall thickness.Both longitudinal and transverse condensers with multiple shell tanks and varied shell tank lengths from 8-14 m have been considered in this work.Three different tertiary coolant temperature rises were chosen as 4°C,8°C and 12°C by considering tropical region average reservoir water temperature range of 28°C to 32°C during summer.Velocity of tertiary coolant was kept within 0.75-1.5 m/s to ensure sufficient turbulence to avoid erosion-corrosion of the tubes.Numerical simulation has been employed to obtain tube-side pressure drop and convection heat transfer coefficient directly from tertiary coolant inlet velocity using κ-ω turbulent flow model.A new iterative“Velocity-search algorithm”has been developed that focuses on finding the correct tertiary coolant velocity instead of overall heat transfer coefficient.Results revealed that velocity-search algorithm yielded very close to the important physical and thermal parameters of condenser compared to the existing design data in large scale nuclear power plants.Velocity-search algorithm has given less number of condenser design physical parameters that meets the velocity acceptance criteria for longitudinal condenser compared to the transverse condenser.Finally,velocity-search algorithm is found to be more reliable,robust,and consistent for condenser design compared to the conventional design algorithm used in Log-Mean Temperature Difference method.

CPR1000机组凝汽器半侧安全运行评价方法研究

我国CPR1000压水堆核电机组半侧运行期间凝汽器多次发生冷却管断裂事件,造成严重的安全影响和经济损失,半侧运行时支撑板跨距偏大引起的弹性汽流激振是主要可能原因之一。美国热交换协会等相关设计规范中并没有给出半侧运行时的计算方法。结合冷却管振动基本理论,推导出凝汽器支撑板跨距计算的数学模型,可以适用于压水堆核电机组凝汽器半侧运行的跨距计算。该文以某压水堆核电项目为例,对比不同凝汽器的冷却管评价模型,利用某核电站半侧运行的实际工况数据对评价方法进行验证,提出Connors系数(Kc)值2.3及风险因子0.8可以适用于压水堆核电站凝汽器钛管安全运行的评价准则。基于振动碰磨理论计算出的凝汽器安全跨距最小为556mm,现场应变初步测量结果也证实该方法是足够保守的,可利用该方法进行支撑板跨距的设计校核。

水润滑推力轴承摩擦学性能的分析和试验研究

针对某堆型核电机组主泵水润滑推力轴承极端工况及超长服役周期的特点,从分析和试验两个方面进行了摩擦学的相关研究工作。在分析方面,考虑了平衡支撑下的润滑性能分析,仿真分析了推力轴承在不同转速下的最小润滑膜厚度h_(min)、摩擦力矩M_(f)和温升ΔT,并根据起飞转速建立了基于启停次数的轴承磨损预测模型;在试验方面,进行了与轴承材料相关的摩擦磨损试验,获得了石墨-不锈钢摩擦副在边界润滑和干摩擦下的摩擦因数f和体积磨损量ΔW,通过缩比的台架试验测试了石墨轴承的润滑膜厚度h、瓦面温度T及磨损质量Δm。结果表明:工作转速对h_(min)影响较大,对ΔT影响较小,石墨材料对干摩擦下的f和ΔW影响较大;从h和T来看,试验结果与计算分析的结果较为接近,说明半尺寸缩比试验与实际使用工况相似性较佳。

凝汽器管道壁面泄漏流场数值模拟研究

凝汽器作为核电厂热力循环中的冷却设备,乏汽通过与其内的冷却水进行热交换形成冷凝水以进行再次热力循环。凝汽器所用的海水冷却水一旦泄漏就会对二回路的蒸汽发生器造成污染,进而影响核电站的安全运行。本文针对凝汽器液相泄漏过程的流动特性进行了数值模拟研究,采用Euler-Euler两相流模型和k-ε湍流模型分析了不同流动工况下两相泄漏过程的流场,在常温常压情况下通过粒子图像测速(PIV)对凝汽器冷凝管的泄漏过程进行了可视化测量分析,并与数值模拟作了对比。结果表明:气液两相同时存在于泄漏过程中时,泄漏量与管道内液相流动工况无关,只与管道内外侧压差相关;泄漏的检测时间可设定在发生泄漏后0.003s内,此时泄漏射流附近存在的两个回流涡会导致示踪气体的回流,可能会对示踪气体的传质和对流扩散产生影响;管道内外侧压差为100000Pa情况下,冷却剂泄漏的质量流量约为3.925×10^(-3)kg/s至9.813×10^(-3)kg/s。

火电厂凝结水泵变频电源配置方案选择与经济性分析

目前,变频调节技术已经被广泛应用在各个行业及领域,这是因为该技术的应用能够起到节约资源、减少浪费的效果,适合应用在火电厂等资源型企业运行中,但是火电厂需要做好对凝结水泵变频电源的配置工作,才能够确保变频调节技术的应用效果。为此,首先分析了火电厂凝结水泵及变频器运行原理,其次总结了火电厂凝结水泵变频电源配置方案的设计流程,最后以1000 MW超超临界机组为例,分析了凝结水泵配置方案、凝结水泵变频电源配置方案及选择建议,以期为火电厂凝结水泵变频电源配置工作提供参考。

核电站水工构筑物结构保护优化与耐久性监测探讨

核电站PX泵房长期处于半海水且加药的环境下,盐雾浓度大,服役环境严苛,易产生钢筋锈蚀从而造成钢筋混凝土结构的劣化。为此,本文全方面阐述了核电站水工构筑物常见的腐蚀类型、结构保护优化的措施及相关的耐久性监测手段,为后续类似核电工程的建设提供一些理论参考。

PISAA软件中含不凝性气体的蒸汽冷凝模型研究与验证

针对核电厂事故状态下含不凝性气体的蒸汽在安全壳内壁面上的冷凝传热问题,本文以严重事故一体化分析软件的开发为背景,建立了基于热质比拟方法的安全壳内含不凝性气体蒸汽冷凝计算模型,用于求解安全壳内壁面由于冷凝造成的传热量以及产生的凝结水的量,并通过Wisconsin实验以及ISP47-TOSQAN实验对该模型进行了验证。结果表明:严重事故一体化分析软件中的冷凝模型针对Wisconsin实验各工况下的冷凝传热系数计算结果与实验测量值基本一致,冷凝传热系数平均偏差在10%左右,最大偏差不超过30%,整体与实验测量值吻合较好,均在工程可接受偏差范围内;严重事故一体化分析软件针对ISP47-TOSQAN实验中的阶段A和阶段B的计算结果与实验测量值吻合,稳态1/2/3/4保持一致,准确地模拟了实验过程。

核电站CRF001循环水泵电机非驱动端的振动问题分析

以某核电站CRF001循环水泵电机为研究对象,通过模态分析方法提出非驱动端加固方案。具体研究中通过模态分析方法获取核电站CRF001循环水泵电机非驱动端前六阶固有振型,确认非驱动端振动问题,进而提出非驱动端加固方案。通过有限元仿真分析方法对加固后非驱动端关键部件进行应力分析,初步检验加固方案可行性,另采用工程应用分析实施进一步应用检验,最终确认非驱动端加固方案具有较强应用价值。