Control system design for a pressure-tube-type supercritical water-cooled nuclear reactor via a higher order sliding mode method

Nuclear power plants exhibit non-linear and time-variable dynamics.Therefore,designing a control system that sets the reactor power and forces it to follow the desired load is complicated.A supercritical water reactor(SCWR)is a fourth-generation conceptual reactor.In an SCWR,the non-linear dynamics of the reactor require a controller capable of control-ling the nonlinearities.In this study,a pressure-tube-type SCWR was controlled during reactor power maneuvering with a higher order sliding mode,and the reactor outgoing steam temperature and pressure were controlled simultaneously.In an SCWR,the temperature,pressure,and power must be maintained at a setpoint(desired value)during power maneuvering.Reactor point kinetics equations with three groups of delayed neutrons were used in the simulation.Higher-order and classic sliding mode controllers were separately manufactured to control the plant and were compared with the PI controllers speci-fied in previous studies.The controlled parameters were reactor power,steam temperature,and pressure.Notably,for these parameters,the PI controller had certain instabilities in the presence of disturbances.The classic sliding mode controller had a higher accuracy and stability;however its main drawback was the chattering phenomenon.HOSMC was highly accurate and stable and had a small computational cost.In reality,it followed the desired values without oscillations and chattering.

A novel approach for radionuclide diffusion in the enclosed environment of a marine nuclear reactor during a severe accident

A severe accident in a marine nuclear reactor leads to radionuclide leakage,which causes hidden dangers to workers and has adverse effects of environmental pollution.It is necessary to propose a novel approach to radionuclide diffusion in a confined environment after a severe accident in a marine nuclear reactor.Therefore,this study proposes a new method for the severe accident analysis program MELCOR coupled with computational fluid dynamics scSTREAM to study radioactive diffusion in severe accidents.The radionuclide release fraction and temperature calculated by MELCOR were combined with the scSTREAM calculations to study the radionuclide diffusion behavior and the phenomenon of radionuclide diffusion in different space environments of the reactor under the conditions of varying wind velocities of the ventilation system and diffusion speed.The results show that the wind velocity of the ventilation system is very small or zero,and the turbulent diffusion of radionuclides is not obvious and diffuses slowly in the form of condensation sedimentation and gravity settlement.When the wind speed of the ventilation system increases,the flow of radionuclides meets the wall and forms eddy currents,affecting the time variation of radionuclides diffusing into chamber 2.The wind velocity of the ventilation system and the diffusion speed has opposite effects on the time variation trend of radionuclide diffusion into the four chambers.

Development and Application of Maintenance Template in Pressurized Water Reactor Nuclear Power Plant

Good practices of maintenance optimization in nuclear power field need to be effectively consolidated and inherited,and maintenance optimization can provide technology support to create a long-term reliable and economic operation for nuclear power plants( NPPs) especially for a large number of nuclear powers under construction. Based on the development and application of maintenance template in developed countries,and combining with reliability-centered maintenance( RCM) analysis results and maintenance experience data over the past ten years in domestic NPPs, the development process of maintenance template was presented for Chinese pressurized water reactor( PWR) NPP,and the application of maintenance template to maintenance program development and maintenance optimization combined with cases were demonstrated. A shortcut was provided for improving the efficiency of maintenance optimization in domestic PWR NPP,and help to realize a safe,reliable,and economic operation for domestic NPPs.

Influence of the Impeller/Guide Vane Clearance Ratio on the Performances of a Nuclear Reactor Coolant Pump

An AP1000 nuclear reactor coolant pump is considered to assess the influence of the Impeller/Guide vane clearance on the performances of this type of pumps.Experiments and numerical simulations relying on an unidirectional fluid-solid coupling approach are used to investigate the problem(stress,strain and mode of the rotor).The results reveal the relationship existing between the hydraulic performance of the nuclear reactor coolant pump and the clearance ratio.The effect of clearance ratio on the maximum equivalent stress on the back surface of the impeller blade is greater than that on the working surface(the maximum equivalent stress on the back surface of impeller blade is about three times that on the working surface).The clearance ratio has a scarce effect on the first six natural frequencies of the rotor of the nuclear reactor coolant pump.The related vibrational modes have different waveforms.

Neutronic design investigation of a liquid injection-based second shutdown system for a typical research reactor using MCNPX

Safety systems, built on state-of-the-art technology, are essential for achieving acceptable levels of plant safety to minimize hazards to the reactor and the general public. The second shutdown system(SSS) as an engineered safety feature and a part of the reactor protection system(RPS) is a means for rapidly shutting down a nuclear reactor, keeping it in a subcritical state and serving as a backup to the first shutdown system(FSS). In this research, one SSS with two types of optimum chamber designs is proposed that take into account the main current characteristic features of the Tehran research reactor with improvements over earlier designs. They are based on a liquid neutron absorber injection that is preferably different, diverse, and independent from the FSS based on the rod drop mechanism. The major design characteristics of this SSS with two different chambers were investigated using MCNPX 2.6.0 code. The performed calculations showed that the designed SSS is a reliable shutdown system, assuring an appropriate shutdown margin and injection time, with no significant effects on the effective delayed neutron fraction while causing minimal variations to the core structure. Further, the reasonable financial cost and the prolongation of the operation cycle are additional advantages of this design.

Artificial Intelligence Driven Nuclear Power Reactors(A Technical Memorandum)

The 21st Century era and new modern technologies surrounding us day-in and day-out have opened a new door to“Pandora Box”,that we do know it as AI(artificial intelligence)and its two essential integrated components namely ML(machine learning)and DL(deep learning).However,the strive and progress in AI,ML,and DL pretty much has taken over any industry that we can think of,when it comes to dealing with cloud of structured data in form of BD(big data).A NPP(nuclear power plant)has multiple complicated dynamic system-of-components that have nonlinear behaviors.For controlling the plant operation under both normal and abnormal conditions,the different systems in NPPs(e.g.,the reactor core components,primary and secondary coolant systems)are usually monitored continuously,which leads to very huge amounts of data.Of course Nuclear Power Industry in form of GEN-IV(Generation IV)has not been left behind in this 21st century era by moving out of GEN-III(Generation III)to more modulars form of GEN-IV,known as SMRs(small modular reactors),with a lot of electronic gadgets and electronics that read data and information from it to support safety of these reactor,while in operation with a built in PRA(probabilistic risk assessment),which requires augmentation of AI in them to enhance performance of human operators that are engaged with day-to-day smooth operation of these reactors to make them safe and safer as well as resilience against any natural or man-made disasters by obtaining information through ML from DL that is collecting massive stream of data coming via omni-direction.Integration of AI with HI(human intelligence)is not separable,when it comes to operation of these smart SMRs with state of the art and smart control rooms with human in them as actors.This TM(technical memorandum)is describing the necessity of AI playing with nuclear reactor power plant of GEN-IV being in operation within near term sooner than later,when specially we are facing today’s cyber-attacks with their smart malware agents at work.

Wide Range Neutron Monitoring(WRNM)System in Boiling Water Reactors(A Short Communication&Memorandum)

The WRNM(wide range neutron monitoring)is a newly developed neutron monitoring channel which was initially conceived as a means to meet Regulatory Guide 1.97 requirements for post-accident neutron monitoring.The scope was expanded to include the startup monitoring function with the aim of replacing both the source and IRMs(intermediate range monitors)in BWRs(boiling water reactors).The WRNMs,consisting of a newly designed fixed incore regenerative sensor and new electronics,which include both counting and MSV(mean square voltage)channels,have been tested in several reactors and its capabilities have been confirmed.The channel will cover the neutron flux range from 103 nv to 1.5×103 nv;it has greater than 1 decade overlap between the counting and MSV channels.Because of the regenerative fissile coating the sensor,even though fixed incore,has a life of approximately 6.0 full power years in a 51 kW/L BWR and similar situation has been proposed for newly designed small modular reactor such as BWRX-300 of General Electric Hitachi reactor.

Dynamic evaluation of a scaled-down heat pipe-cooled system during start-up/shut-down processes using a hardware-in-the-loop test approach

Micro-mobile heat pipe-cooled nuclear power plants are promising candidates for distributed energy resource power genera-tors and can be flexibly deployed in remote places to meet increasing electric power demands.However,previous steady-state simulations and experiments have deviated significantly from actual micronuclear system operations.Hence,a transient analysis is required for performance optimization and safety assessment.In this study,a hardware-in-the-loop(HIL)approach was used to investigate the dynamic behavior of scaled-down heat pipe-cooled systems.The real-time features of the HIL architecture were interpreted and validated,and an optimal time step of 500 ms was selected for the thermal transient.The power transient was modeled using point kinetic equations,and a scaled-down thermal prototype was set up to avoid mod-eling unpredictable heat transfer behaviors and feeding temperature samples into the main program running on a desktop PC.A series of dynamic test results showed significant power and temperature oscillations during the transient process,owing to the inconsistency of the rapid nuclear reaction rate and large thermal inertia.The proposed HIL approach is stable and effective for further studying of the dynamic characteristics and control optimization of solid-state small nuclear-powered systems at an early prototyping stage.

Studies on Production Planning of Dispersion Type U3Si2-Al Fuel in Plate-Type Fuel Elements for Nuclear Research Reactors

Several fuel plants that supply nuclear research reactors need to increase their production capacity in order to meet the growing demand for this kind of nuclear fuel. After the enlargement of the production capacity of such plants, there will be the need of managing the new production level. That level is usually the industrial one, which poses challenges to the managerial staff. Such challenges come from the fact that several of those plants operate today on a laboratorial basis and do not carry inventory. The change to the industrial production pace asks for new actions regarding planning and control. The production process based on the hydrolysis of UF6 is not a frequent production route for nuclear fuel. Production planning and control of the industrial level of fuel production on that production route is a new field of studies. The approach of the paper consists in the creation of a mathematical linear model for minimization of costs. We also carried out a sensitivity analysis of the model. The results help in minimizing costs in different production schemes and show the need of inventory. The mathematical model is dynamic, so that it issues better results if performed monthly. The management team will therefore have a clearer view of the costs and of the new, necessary production and inventory levels.

Prospects of Technological Improvement of Nuclear and Environmental Safety of World Energy

Today, the most urgent problem of the existing and future nuclear power industry is to ensure the nuclear and environmental safety of the operation of nuclear power reactor units (NPPs) and nuclear power plants (NPPs). It is solved thanks to the application of deeply echeloned protection and an anti-accident complex of methods and means for effective control of the operation of active reactor zones (AZR). However, the danger of existing NPPs in the world from time to time manifests itself in the form of severe post-project accidents and catastrophes with the release into the environment of a significant amount of radioactive materials dangerous for all living things. The results of the analysis show that the unconditional fulfillment of the main requirements of nuclear environmental safety and biocompatibility is possible only in the so-called wave nuclear reactor of the G-V generation, which, unlike reactors of the previous generations III, II+ and IV, does not require supercritical loading of the core with nuclear fuel. In the active zone of this reactor, nuclear-physical processes governed by physical law are implemented, which exclude the operator’s participation in regulating the reactivity of the reactor’s active zone, which makes it the reactor with the highest level of nuclear and environmental safety today, which is based on the principles of so-called internal safety, free from the human factor. The possibility of burning nuclear fuel based on U238 and Th232 in it expands the reserves of energetic nuclear fuel almost to inexhaustibility. The technology of nuclear reactors of the G5 generation through the secondary use of spent irradiated nuclear fuel (SNF) for the production of energy and energy raw materials with simultaneous burning of it to an environmentally safe state is able to quickly reduce the available stocks and further production of dangerous SNF, guarantee the nuclear and environmental safety of NPPs with reactors G5 and to technologically make nuclear post-project accidents and disasters impossible at the level of physical law with the complete elimination of the human factor.

Assessment of Axial Power Peaking Factors in GHARR-1 LEU Core: A Decadal Simulation Analysis

This study aims to thoroughly investigate the axial power peaking factors (PPF) within the low-enriched uranium (LEU) core of the Ghana Research Reactor-1 (GHARR-1). This study uses advanced simulation tools, like the MCNPX code for analysing neutron behavior and the PARET/ANL code for understanding power variations, to get a clearer picture of the reactor’s performance. The analysis covers the initial six years of GHARR-1’s operation and includes projections for its whole 60-year lifespan. We closely observed the patterns of both the highest and average PPFs at 21 axial nodes, with measurements taken every ten years. The findings of this study reveal important patterns in power distribution within the core, which are essential for improving the safety regulations and fuel management techniques of the reactor. We provide a meticulous approach, extensive data, and an analysis of the findings, highlighting the significance of continuous monitoring and analysis for proactive management of nuclear reactors. The findings of this study not only enhance our comprehension of nuclear reactor safety but also carry significant ramifications for sustainable energy progress in Ghana and the wider global context. Nuclear engineering is essential in tackling global concerns, such as the demand for clean and dependable energy sources. Research on optimising nuclear reactors, particularly in terms of safety and efficiency, is crucial for the ongoing advancement and acceptance of nuclear energy.

核热推进系统分析程序模型与计算方法初步研究

核热推进(NTP)系统具有高比冲、大推力和工作时间长等特点,在深空探测和轨道机动等方面具有明显的优势。系统性能分析是NTP系统研发与设计的重要内容。结合对国际历史上已开发程序的分析以及现阶段的研发需求,将系统性能分析划分为稳态设计点性能分析与优化、稳态非设计点性能分析以及瞬态性能分析3个主要环节。在清华大学核能与新能源技术研究院自主开发的核动力发动机系统分析程序PANES基础上,提出了基于“流网-热网”的系统分析程序框架,并建立了反应堆中子动力学与涡轮泵动态特性等数学模型,提出了对应的计算分析方法,拓展了原程序的功能。该工作为NTP系统设计方法的进一步研究和应用提供了重要基础。

核电厂反应堆冷却剂系统抗震阻尼比研究

基于2023年国内新建核电厂安全审评中核安全监管部门对抗震物项阻尼比取值提出关注的背景。介绍核电工程中抗震分析阻尼比取值依据,指出标准与工程实践之间存在的差异,以及由此产生的困惑;阐述阻尼比在动态分析中的作用原理;对比核电领域不同标准与导则文件对于机械设备阻尼比的要求,指出当前标准的相关要求对于由多种部件组成的组合设备或系统过于保守;重点对压水堆核电厂反应堆冷却剂系统与设备阻尼比进行研究,给出国内外核电工程实践中该系统与设备的阻尼比取值依据,并针对核电工程实践中组合设备或系统阻尼比取值依据不足的问题提出建议。

堆外核测量系统输出高压纹波测试准确性研究

为解决堆外核测量系统输出高压纹波测试过程极易受到外部环境干扰,导致测试数据波动大、不稳定的问题,基于当前高压纹波测试现状,对高压纹波测试准确性进行研究。从数据仿真、结构设计、信号线缆优化、安全防护与验证、测试流程标准化等五个方面入手,通过总结其他行业高压纹波测试方法和结合堆外核测系统产生高压的频率特点,提出一种基于堆外核测量系统输出高压纹波测试准确性设计方法。利用数据仿真技术计算出符合堆外核测量系统高压纹波测试的隔直滤波电容容值,优化整个高压纹波测试外部环境以减少外部电磁干扰。通过验证测试环境的安全性、可靠性以及制定相关测试标准,确保堆外核测量系统输出高压纹波测试的准确性。该研究能够有效提升堆外核测量系统中输出高压纹波测试的准确性,解决了现存问题。

Nuclear geyser model of the origin of life:Driving force to promote the synthesis of building blocks of life

We propose the nuclear geyser model to elucidate an optimal site to bear the first life.Our model overcomes the difficulties that previously proposed models have encountered.Nuclear geyser is a geyser driven by a natural nuclear reactor,which was likely common in the Hadean Earth,because of a much higher abundance of 235U as nuclear fuel.The nuclear geyser supplies the following：（1）high-density ionizing radiation to promote chemical chain reactions that even tar can be used for intermediate material to restart chemical reactions,（2）a system to maintain the circulation of material and energy,which includes cyclic environmental conditions（warm/cool,dry/wet,etc.）to enable to produce complex organic compounds,（3）a lower temperature than 100℃ as not to break down macromolecular organic compounds,（4）a locally reductive environment depending on rock types exposed along the geyser wall,and（5）a container to confine and accumulate volatile chemicals.These five factors are the necessary conditions that the birth place of life must satisfy.Only the nuclear geyser can meet all five,in contrast to the previously proposed birth sites,such as tidal flat,submarine hydrothermal vent,and outer space.The nuclear reactor and associated geyser,which maintain the circulations of material and energy with its surrounding environment,are regarded as the nuclear geyser system that enables numerous kinds of chemical reactions to synthesize complex organic compounds,and where the most primitive metabolism could be generated.

Novel Technological Developments with Impacts on Perspectives for Mobile Nuclear Power Plants

New research developments suggest that nuclear reactors using fusion may enter the market sooner than imagined even for mobile applications, like merchant ship propulsion and remote power generation. This article aims at pointing such developments and how they could affect nuclear fusion. The method is enumerating the main nuclear reactors concepts, identifying new technological or theoretical developments useful to nuclear field, and analysing how new recombination could affect feasibility of nuclear fusion. New technologies or experimental results do not always work the way people imagine, being better or worse for intended effects or even bringing completely unforeseen effects. Results point the following designs could be successful, in descending order of potential: aneutronic nuclear reactions using lattice confinement, aneutronic nuclear reactions using inertial along magnetic confinement, hybrid fission-lattice confinement fusion, and fission reactions.

Nuclear Energy and Its History: Past Consequences, Present Inadequacies and a Perspective for Success

An attempt is made to locate nuclear technology within a logical context considering history, risks, societal catastrophes and perspectives: the need is identified for a new direction in the exploitation in order to restore the role in energy production. We depict the situation coming from a marvelous history of discoveries started at the beginning of the XX century;heroes are recalled who made possible something that is inconceivable today: design, construction and production of electricity in a few years;that history was tainted by intentional nuclear explosions, i.e. the original sin that we are now paying. Then, we attempt to show that the societal risk is an inherent part of the civilization. Restoring the public trust (towards nuclear fission technology) by matching nuclear safety with the current technological status and advancers in risk assessment is the key objective. The “independent assessment”, or a principle for the exploitation of nuclear energy already stated in the 50’s of the previous century, shall then re-appear. This is used to erect the signpost for a “dynamic barricade” to further reduce the risk of operation of nuclear reactors and to match the design with current technological capabilities and with the frontiers of the research.

超临界二氧化碳核能动力系统的兴起和发展

超临界二氧化碳(S-CO_(2))核能动力系统以S-CO_(2)为工质,通过直接或间接循环将核释热转换为电能或机械能。本文总结了国际上S-CO_(2)核能动力系统的概念初创、研究重启、协同创新3个历史阶段近60年的发展历程,分析了S-CO_(2)核能动力系统在核反应堆设计、材料工艺、热工流体力学、换热器、涡轮发电系统、系统运行、控制及安全策略等方面的研究现状,并提出了S-CO_(2)核能动力系统在基础研究和工程攻关中面临的技术挑战和攻关方向。目前,以中美为代表的能源强国已初步完成S-CO_(2)动力转换系统的实验室级测试,预计可在5~10年内实现中等规模工程示范甚至规模化应用。S-CO_(2)核能动力系统基本具备了走向工程的前提条件,有望引领先进核能技术变革。

Nanomaterials Driven Magnetic Nuclear Fusion Confinement Approaches(A Technical Memorandum)

Nuclear energy driven magnetic confinement via donut shape device known as Tokamak,a toroidal apparatus,for producing controlled fusion reactions in hot plasma,was originally suggested as a basic yet more promising fusion reactor.Today the more innovative version of this apparatus that is known as an ITER(international thermonuclear experimental reactor)shows a way toward MCF(magnetic confinement fusion)of hot plasma goal by satisfying Lawson’s Criteria to some degree of achievement.However,since this fusion driven reactor of hot plasma needs to operate at almost 150 million Celsius,the internal material of this reactor is a matter of concern for scientists that are involved with its fabrication.Uniqueness of nanomaterials from the point of view of physical and chemical properties is suggested as a possible potential application for this special and innovative reactor for a nuclear fusion device.Convergence of nanotechnology in study of new generation of materials of this kind can shape the path for various technological developments and a large variety of disciplines,including MCF driven plasma of hot fusion as well.This short TM(technical memorandum)written by these two authors will cover this aspect of technology in a holistic way and the more granular level is left to the reader of this TM to investigate further.

反应堆系统遇水下爆炸载荷环境与关键设备陆地冲击试验载荷匹配研究

船用反应堆系统的抗冲击性能是决定核安全的重要设计维度。在核安全审评活动中,对反应堆系统抗冲击的主要关注点有两项:能够表征实际条件下平台遭遇水下爆炸时反应堆系统与设备的冲击设计载荷;抗冲击的设计载荷与根据陆上抗冲击试验载荷的匹配问题。本文从上述两个问题出发,首先建立了反应堆系统遇水下爆炸冲击环境预报的计算手段,开发了基于国产有限元平台的载荷预报程序并进行了缩比模型的试验验证。应用该程序,对某型反应堆系统在考虑舱体、基座、筏架、重型设备耦合作用情况下的冲击载荷传递机理进行了仿真,获得了反应堆系统关键设备接口位置的冲击设计环境。此外,本文建立了中型摆锤冲击机的虚拟试验模型并进行了台架试验验证。应用虚拟试验技术对燃料组件设计-试验载荷环境匹配性进行研究,得到了能够匹配燃料组件设计载荷环境下的陆地冲击机试验参数设置。本文所述研究成果统一了核级设备抗冲击设计和试验的载荷环境,为后续产品研制中充分考虑实际条件下的冲击载荷提供了技术支撑。