This paper presents the architecture and functional requirements of diesel generator control systems for nuclear power plant units. It proposes a detailed design methodology based on the NASPIC platform for unit diesel generator control systems, including the implementation of control system cabinets. By leveraging the NASPIC platform’s advantages in high-performance computing, real-time response capability, reliability, and information security, this design effectively fulfills the functional requirements of diesel generator control systems. The control cabinets have demonstrated stable operation in nuclear power plants, validating the effectiveness and reliability of the proposed design.

To ensure the safe, reliable, and stable operation of nuclear power plants, the instrumentation and control (I C) systems of nuclear power plants provide various control, protection, and monitoring methods for the plants processes and equipment. This paper conducts extensive research on literature and data related to the development of domestic nuclear power I C systems. By examining the current usage of I C systems in operational and under-construction nuclear power plants in China, it outlines the evolutionary process—from the initial analog instrumentation with relay-based control, to analog control combined with non-safety digital monitoring, and finally to the gradual adoption of fully digital I C systems, including safety-class systems. Additionally, this paper analyzes the characteristics of modern digital I C systems and summarizes the design principles for nuclear power instrumentation and control. Finally, it explores and forecasts future technological advancements in nuclear power I C systems. By clearly outlining the current applications and providing a detailed analysis of development trends, this paper offers valuable insights for the future growth of the nuclear power I C industry.

Image-text retrieval is a widely studied topic in the field of computer vision due to the exponential growth of multimedia data, whose core concept is to measure the similarity between images and text. However, most existing retrieval methods heavily rely on cross-attention mechanisms for cross-modal fine-grained alignment, which takes into account excessive irrelevant regions and treats prominent and non-significant words equally. This paper aims to investigate an alignment approach that reduces the involvement of non-significant fragments in images and text while enhancing the alignment of prominent fragments. For this purpose, we introduce the Cross-Modal Prominent Fragments Enhancement Aligning Network(CPFEAN). In practice, we first design a novel intra-modal fragments relationship reasoning method, and subsequently employ our proposed alignment mechanism to compute the similarity between images and text. Extensive quantitative comparative experiments on MS-COCO and Flickr30K datasets demonstrate that our approach outperforms state-of-the-art methods.

This paper analyzes the current status of the control system of the nuclear diesel generator set. Aiming at the problems of aging and misoperation of the electrical equipment of the original control system of the emergency diesel generator set in a spent fuel receiving and storage facility, the reform plan is proposed from the aspects of functional performance, fault alarm and protection, and seismic resistance of the storage battery. It realizes the stable safe and reliable operation of diesel generator set, and can monitor the data in real time to improve the maintainability.

基于视觉的智能驾驶作为自动化技术的重要应用成为当前研究的热点,但受到实验场景客观限制很难引入课堂的实践教学.为引导学生从理论出发、由实践深入,切实培养敢创新、善实践的复合型人才,构建了基于计算机视觉的智能驾驶虚拟仿真平台,在此基础上提出了一种感知环境深度信息的仿真算法,为解决实际场景标注大量样本的问题提供了解决方案;同时平台提供了多样化的智能驾驶仿真场景、多角度的观察视野以及丰富的功能控制模块.教学实践表明,该平台为学生系统地掌握计算机视觉领域知识,激发学习热情和培养创新实践能力起到了促进作用,平台也为实践教学提供了新的模式.

Image-text retrieval is a complicated and challenging task in the cross-modality area, and lots of experiments have made great progress. Most existing researches process images and text in one pipeline or are highly entangled, which is not practical and human-friendly in the real world. Moreover, the image regions extracted by Faster-RCNN are highly over-sampled in the image pipeline, which causes ambiguities for the extracted visual embeddings. From this point of view, we introduce the Bottom-up Transformer Reasoning Network (BTRN). Our method is built upon the transformer encoders to process the image and text separately. We also embed the tag information generated by Faster-RCNN to strengthen the connection between the two modalities. Recall at K and normalized discounted cumulative gain metric (NDCG) metrics are used to evaluate our model. Through various experiments, we prove our model can reach state-of-the-art results.

Image inpainting aims to restore the missing regions of corrupted images plausibly. Recent techniques of image inpainting handle it as a pure generation problem and utilize large-scale training data to learn semantic and texture information to fill the missing area. However, generative networks frequently distort local structures or introduce visible artifacts while filling large missing regions. To address this issue, we propose SLTFill, an image inpainting method that uses a set of reference images which share scene contents with the corrupted image to guide the synthesis of missing regions. A key challenge addressed by our approach is aligning the transformations of the reference images with the context of the corrupted image and making sure they are consistent in spatial structure and light space. Qualitative and quantitative experiments demonstrate the state-of-the-art performance of our approach on the image inpainting task.

后备盘(BUP)主要作为核电厂主控制室数字化技术失效后的备用手段,用于安全停堆并将反应堆维持在冷停堆水平以及事故工况下处理.后备盘实现报警、显示和控制三类功能,其安全、可靠性对于保障反应堆安全停堆十分重要.开展对核电厂后备盘的辐射散热以及自然对流的热分析模拟,定量地对辐射换热以及自然对流在后备盘中所起的作用进行分析说明,掌握后备盘散热的机理.研究结果表明,通过仿真模拟得到的后备盘内部的最高温度在为60℃,满足电源模块最高运行温度70℃的要求;后备盘散热过程中辐射换热过程占据较大的比例,辐射换热与自然对流的比例大概为6∶1,空气自然对流对机柜本身造成的温升可以忽略;真实后备盘内部的结构件与线缆对自然循环散热以及辐射换热所造成的影响可以忽略.该研究结果与经验可以为复杂工况下的核动力仪控设备的散热分析提供参考.

The nuclear reactor 1E DCS priority logic module is one of the very important components of the nuclear reactor I&C (Instrument and Control) system. Its main function is to prioritize different DCS control command, and then drive especial safety facilities and related support systems (pumps, valves and other related equipment). Driven devices such as pumps, valves, etc. need to execute related functions according to the drive signals issued by the priority logic modules. The characteristics and advantages between seven kinds of domestic and international priority logic modules are analyzed by comparing their function and implementation mode. Comparison shows that different manufacturers' products also have their own characteristics. At present, many companies at home and abroad have designed different priority logic modules based on their own ideas. However, the public literature about priority logic modules is very limited, let alone summary analysis about it. Therefore, this paper has important engineering significance and reference value for the design of priority logic module.

安全级DCS优先级逻辑控制模块是核反应堆仪控系统非常重要的组成部分之一,其主要功能是对不同的安全级DCS控制命令进行优先级判断,然后驱动专设安全设施及相关支持系统的泵、阀等相关设备.文中对国外4种优先级逻辑模块进行了介绍,通过仔细分析不同优先级逻辑模块的实现方式与功能,对比出它们之间的特点与优势.对比结果表明,TXS的AV42板卡以及Common Q的CIM板卡是最为复杂的.AV42板卡能够实现以开环以及闭环两种方式控制就地设备,并且在就地控制中还增加了诊断技术.此外,AV42板卡还增加了对设备卡顿情形的处理机制.为了防止误触发,Common Q的CIM板卡版面上增加了停止功能.此外,为了防止软件共因故障,CIM上还提供了手动闭锁功能(独立于软件系统的指令).其他厂家的产品也有他们各自的优点.目前,国内外的许多厂家都针对他们各自的想法设计出不同的优先级逻辑模块.然而,目前针对优先级逻辑模块的公共资料十分有限.因此,本文对优先级逻辑模块的设计拥有重要的工程应用以及参考价值.

基于数字孪生的理论,分析了核仪控DCS领域内的应用需求,提出了数字化的全生命周期设计框架,阐述了在各阶段的建设任务.此外,还阐述了基于数字孪生的核仪控产品从设计到生产制造阶段的关键技术.

核电厂安全级仪控系统的T2试验一般是指系统的逻辑功能试验,即包括紧急停堆系统和专设安全设施驱动系统的功能.依托核电厂数字化安全级仪控系统(NASPIC平台),基于概率安全分析(PSA)模型,提出了一种针对于T2定期试验测试用例集的生成策略.经验证,本文提出的T2定期试验测试用例集生成策略可以覆盖全部的系统逻辑功能,满足了国内外相关标准的要求.

安全级数字化仪控系统(Digital Control System, DCS)优先级逻辑模块是核反应堆仪控系统重要的组成部分之一,其主要功能是针对不同的DCS控制命令进行优先级判断,然后驱动专设安全设施及支持系统的泵、阀等相关设备.文中对国内3种优先级逻辑模块进行了介绍,通过分析不同优先级逻辑模块的实现方式与功能,对比其特点与优势.结果表明NASPIC的ACM模块在防止软件共因故障方面表现更为优异.这种纯硬件结构不仅能降低模块响应的时间,还能有效地避免软件共因故障,但这也导致了其能够实现的功能相对有限.Firmsys平台CIM的功能更为复杂.CIM中既含有MCU又含有CPLD,能够对设备实现分类化、精细化控制.NASPIC的PLM是完全基于CPLD建立起来的,PLM可以实现的功能相对丰富,为了避免软件共因故障,PLM中没有设计微处理器结构.目前,国内外基于自身理念设计的不同优先级逻辑模块的相关文献十分有限,因此,本文具有重要的工程意义与参考价值. 

In the passive residual heat removal system of a molten salt reactor, one of the residual heat removal methods is to use the thimble-type heat transfer elements of the drain salt tank to remove the residual heat of fuel salts. An experimental loop is designed and built with a single heat transfer element to analyze the heat transfer and flow characteristics. In this research, the influence of the size of a three-layer thimble-type heat transfer element on the heat transfer rate is analyzed. Two methods are used to obtain the heat transfer rate, and a difference of results between methods is approximately 5%. The gas gap width between the thimble and the bayonet has a large effect on the heat transfer rate. As the gas gap width increases from 1.0 mm to 11.0 mm, the heat transfer rate decreases from 5.2 kW to 1.6 kW. In addition, a natural circulation startup process is described in this paper. Finally, flashing natural circulation instability has been observed in this thimble-type heat transfer element.

Background: In contrast with pressure water reactor, molten salt reactors (MSR) have low pressure, small size, high multiplication efficiency, and little residual radioactivity. The residual heat removal system is an important part of nuclear power reactor to ensure the reactor safety. However, little attention has been paid to the passive residual heat removal system (PRHRS) of the molten salt reactor in the literature so far. Purpose: This study aims to make some preliminary work for the design of the passive residual heat removal system and to make sure the safety and stability of residual heat removal of MSRs. Methods: The heat transfers and flow characteristics are analyzed for the cooling thimble of MSRs. An experimental loop is designed and built with a single heat transfer element to simulate the sensible heat and decay heat removal process in drain salt tank when the reactor is shut down. The changes of pressure and temperature are analyzed through which the characteristics of flashing flow instability are studied. Moreover, the changes of heat transfer coefficients are analyzed. Results: The heat transfer rate increases from 550 W to 1900 W when the thimble outer wall temperature increases from 440 oC to 700 oC. And the heat transfer coefficient changes from 800 W·(m2·K)?1 to 1700 W·(m2·K)?1 during the flashing process. Conclusion: The flashing phenomenon will cause the periodical change of natural circulation flow rate and temperature of pipe wall, resulting in the periodic fluctuation of wall temperature and periodic mechanical vibration of the system. In the process of designing the molten salt reactor passive heat removal system, the flashing flow instability phenomenon should be taken into consideration to ensure the safety and stability of heat removal.

The passive residual heat removal system of molten salt reactor consists of a boiling and a condensing loop, which are connected by a steam drum. A numerical model based on two-fluid model in RELAP5 is developed to demonstrate the use of the passive system and investigate the interaction between the boiling loop and the condensing loop at steady state. The numerical work is partially validated by experimental results obtained in a single-tube natural circulation loop. Heat flux, initial water inventory, the height of the condensing loop and the heat transfer capacity of the condenser are changed separately to study how the system responds. The results show that the mass flow rate in the boiling loop indeed has a maximum. Both the maximum value and the system pressure are mainly affected by heat flux and heat transfer capacity of the condenser. The mass flow rate in the condensing loop is controlled by heat flux instead of other parameters, thus it seems that the condensing loop is a follower of the boiling loop. The initial water inventory and the height of the condensing loop have little influence on the system performance. Additionally, Bubbly flow, slug flow, vertical stratified flow, horizontal stratified flow and annular-mist flow appear in the natural circulation system. (C) 2016 Published by Elsevier Ltd.

To get an insight into the operating characteristics of the passive residual heat removal system of molten salt reactors, a two-phase natural circulation test facility was constructed. The system consists of a boiling loop absorbing the heat from the drain tank, a condensing loop consuming the heat, and a steam drum. A steady-state experiment was carried out, in which the thimble temperature ranged from 450°C to 700°C and the system pressure was controlled at levels below 150 kPa. When reaching a steady state, the system was operated under saturated conditions. Some important parameters, including heat power, system resistance, and water level in the steam drum and water tank were investigated. The experimental results showed that the natural circulation system is feasible in removing the decay heat, even though some fluctuations may occur in the operation. The uneven temperature distribution in the water tank may be inevitable because convection occurs on the outside of the condensing tube besides boiling with decreasing the decay power. The instabilities in the natural circulation loop are sensitive to heat flux and system resistance rather than the water level in the steam drum and water tank. RELAP5 code shows reasonable results compared with experimental data.

An experimental setup has been designed and fabricated for the analysis of the passive residual heat removal system as applied to molten salt reactor. The main characteristic of this test facility is the presence of two natural circulation loops and a cooling thimble with double-wall barrier. The objective is to investigate the thermal performance of the drain tank cooling system, as well as further understand the heat transfer process. It is observed that the passive cooling system with single cooling thimble is capable of removing 2175 W at the desired temperature of 699 degrees C. Due to the large temperature difference, radiation heat transfer plays a significant role between the thimble and the bayonet tube. The contributions of several constituent thermal resistances in the heat flow path, together with the overall thermal resistance, have been obtained at various boundary conditions. It is also found that the operating mode is quite unstable with periodical oscillations at relative low temperature. From this study, valuable reference data and useful information are provided for practical applications. (C) 2016 Elsevier Ltd. All rights reserved.