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.

Comissioning of the lEA-R1 Nuclear Reactor New Heat Exchanger

This work presents results on the commissioning of the new heat exchanger of the IEA-R1 nuclear reactor in the occasion of its operational power upgrade from 2 MW to 5 MW, in comparison to the values calculated in the project of IESA Design and Equipments Company. This reactor is a swimming pool type, light water moderated and with graphite reflectors, used for research purposes and medical radioisotopes production. During monitoring procedures, issues were observed on the reactor operation at 5 MW mainly due to the ageing of the reactor＇s oldest heat exchanger （TC-A） and excessive vibrations at high flow rates on the other installed heat exchanger （TC-B）. So it was decided to provide a new IESA heat exchanger with 5 MW capacity to definitely substitute the TC-A heat exchanger. The results show that the IEA-R1 nuclear reactor can be operated safely and continuously at 5 MW with the new IESA heat exchanger.

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.

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.

Study and Evaluation of Aluminum Capsules to Irradiation of Gaseous Samples in Nuclear Reactor

Gas irradiation in research nuclear reactors is an important way to produce radionuclides. Although some nuclear reactors centers offer this type of service, there are few publications about capsules to irradiation of gaseous samples. This paper describes a method to fabricate and evaluate aluminum capsules to irradiate gaseous samples in nuclear reactor. A semi-circular slotted die from a hydraulic presshead was modified to seal aluminum tubes. The aluminum capsules were subjected to leak detection tests, which demonstrated the accordance with standard ISO 9978.

Thermal–hydraulic analysis of space nuclear reactor TOPAZ-Ⅱ with modified RELAP5

With the advantages of high reliability, power density, and long life, nuclear power reactors have become a promising option for space power. In this study, the Reactor Excursion and Leak Analysis Program 5(RELAP5), with the implementation of sodium–potassium eutectic alloy(NaK-78) properties and heat transfer correlations, is adopted to analyze the thermal–hydraulic characteristics of the space nuclear reactor TOPAZ-Ⅱ.A RELAP5 model including thermionic fuel elements(TFEs), reactor core, radiator, coolant loop, and volume accumulator is established. The temperature reactivity feedback effects of the fuel, TFE emitter, TFE collector,moderator, and reactivity insertion effects of the control drums and safety drums are considered. To benchmark the integrated TOPAZ-Ⅱ system model, an electrical ground test of the fully integrated TOPAZ-Ⅱ system, the V-71 unit,is simulated and analyzed. The calculated coolant temperature and system pressure are in acceptable agreement with the experimental data for the maximum relative errors of 8 and 10%, respectively. The detailed thermal–hydraulic characteristics of TOPAZ-Ⅱ are then simulated and analyzed at the steady state. The calculation results agree well with the design values. The current work provides a solid foundation for space reactor design and transient analysis in the future.

Computational Tools for the Integrated Design of Advanced Nuclear Reactors

Advanced nuclear reactors offer safe, clean, and reliable energy at the global scale. The development of such devices relies heavily upon computational models, from the pre-conceptual stages through detailed design, licensing, and operation. An integrated reactor modeling framework that enables seamless communication, coupling, automation, and continuous development brings significant new capabilities and efficiencies to the practice of reactor design. In such a system, key performance metrics （e.g., optimal fuel management, peak cladding temperature in design-basis accidents, levelized cost of electricity） can be explicitly linked to design inputs （e.g., assembly duct thickness, tolerances）, enabling an exceptional level of design consistency. Coupled with high-performance computing, thousands of integrated cases can be executed simultaneously to analyze the full system, perform complete sensitivity studies, and efficiently and robustly evaluate various design tradeoffs. TerraPower has developed such a tool-the Advanced Reactor Modeling Interface （ARMI） code system-and has deployed it to support the TerraPower Traveling Wave Reactor design and other innovative energy products currently under development. The ARMI code system employs pre-existing tools with strong pedigrees alongside many new physics and data management modules necessary for innovative design. Verification and validation against previous and new physical measurements, which remain an essential element of any sound design, are being carried out. This paper summarizes the integrated core engineering tools and practices in production at TerraPower.

Ultrasound Imaging in Nuclear Reactors Cooled by Liquid Metals

In nuclear reactors cooled by liquid metals, ultrasound is the only type of field that allows obtaining images of the reactor cores and diagnostics of the integrity of the fuel assemblies. The article discusses the features of the practical realization of ultrasonic imaging systems based on phased arrays and offers an alternative solution of imaging on the basis of the acoustic lenses of refractive and diffraction types. Using lenses eliminates many of the technical and technological problems associated with the development of multi-element phased arrays. It is shown that lens systems allow using traditional methods of transformation of acoustic fields into the visible images by 2D piezo matrix and a more promising way of acoustooptical transformation based on coherent optical interferometry.

A Survey of New Methods for Production of Some Radionuclides, at Laboratory Scale, through Secondary Reactions in Nuclear Reactors

The studies performed in the frame of a project destined for the search of new (t,n) and (p,n) reactions of interest in nuclear reactors are described. Experimental evidences of the observations of the reactions: 46Ti(t,n)48V, 48Ti(p,n)48V, 52Cr(t,n)54Mn, 56Fe(p,n)56Co, 72Ge(t,n)74As and 74Ge(p,n)74As, are presented. Additional data on some secondary reactions, already characterised for the production of 7Be, 56Co, 58Co, 65Zn and 88Y, were also obtained. The significance of these data is discussed.

Comparison of Small Modular Reactor and Large Nuclear Reactor Fuel Cost

Small modular reactors (SMRs) offer simple, standardized, and safe modular designs for new nuclear reactor construction. They are factory built, requiring smaller initial capital investment and facilitating shorter construction times. SMRs also promise competitive economy when compared with the current reactor fleet. Construction cost of a majority of the projects, which are mostly in their design stages, is not publicly available, but variable costs can be determined from fuel enrichment, average burn-up, and plant thermal efficiency, which are public parameters for many near-term SMR projects. The fuel cost of electricity generation for selected SMRs and large reactors is simulated, including calculation of optimal tails assay in the uranium enrichment process. The results are compared between one another and with current generation large reactor designs providing a rough comparison of the long-term economics of a new nuclear reactor project. SMRs are predicted to have higher fuel costs than large reactors. Particularly, integral pressurized water reactors (iPWRs) are shown to have from 15% to 70% higher fuel costs than large light water reactors using 2014 nuclear fuels market data. Fuel cost sensitivities to reactor design parameters are presented.

Reactor field reconstruction from sparse and movable sensors using Voronoi tessellation-assisted convolutional neural networks

The aging of operational reactors leads to increased mechanical vibrations in the reactor interior.The vibration of the incore sensors near their nominal locations is a new problem for neutronic field reconstruction.Current field-reconstruction methods fail to handle spatially moving sensors.In this study,we propose a Voronoi tessellation technique in combination with convolutional neural networks to handle this challenge.Observations from movable in-core sensors were projected onto the same global field structure using Voronoi tessellation,holding the magnitude and location information of the sensors.General convolutional neural networks were used to learn maps from observations to the global field.The proposed method reconstructed multi-physics fields(including fast flux,thermal flux,and power rate)using observations from a single field(such as thermal flux).Numerical tests based on the IAEA benchmark demonstrated the potential of the proposed method in practical engineering applications,particularly within an amplitude of 5 cm around the nominal locations,which led to average relative errors below 5% and 10% in the L_(2) and L_(∞)norms,respectively.

Proposal of a Deuterium-Deuterium Fusion/PWR Fission Hybrid Reactor

This article proposes to associate a Deuterium-Deuterium (D-D) fusion reactor with a PWR (fission Pressurized Water Reactor) in a hybrid reactor. Even if the mechanical gain (Q factor) of the D-D fusion reactor is below the unity and consequently consumes more energy than it supplies, due to the high energy amplification factor of the PWR fission reactor, the global yield is widely superior to 1. As the energy supplied by the fusion reactor is relatively low and as the neutrons supplied are mainly issued from D-D fusions (at 2.45 MeV), the problems of heat flux and neutrons damage connected with materials, as with D-T fusion reactors are reduced. Of course, there is no need to produce Tritium with this D-D fusion reactor. This type of reactor is able to incinerate any mixture of natural Uranium, natural Thorium and depleted Uranium (waste issued from enrichment plants), with natural Thorium being the best choice. No enriched fuel is needed. So, this type of reactor could constitute a source of energy for several thousands of years because it is about 90 more efficient than a standard fission reactor, such as a PWR or a Candu one, by extracting almost completely the energy from the fertile materials U238 and Th232. For the fission part, PWR technology is mature. For the fusion part, it is based on a reasonable hypothesis done on present Stellarators projects. The working of this reactor is continuous, 24 hours a day. In this paper, it will be targeted a reactor able to provide net electric power of about 1400 MWe, as a big fission power plant.

An old issue and a new challenge for nuclear reactor safety

Nuclear reactor safety(NRS)and the branch accident analysis(AA)constitute proven technologies:these are based on,among the other things,long lasting research and operational experience in the area of water cooled nuclear reactors(WCNR).Large break loss of coolant accident(LBLOCA)has been,so far,the orienting scenario within AA and a basis for the design of reactors.An incomplete vision for those technologies during the last few years is as follows:Progress in fundamentals was stagnant,namely in those countries where the WCNR were designed.Weaknesses became evident,noticeably in relation to nuclear fuel under high burn-up.Best estimate plus uncertainty(BEPU)techniques were perfected and available for application.Electronic and informatics systems were in extensive use and their impact in case of accident becomes more and more un-checked(however,quite irrelevant in case of LBLOCA).The time delay between technological discoveries and applications was becoming longer.The present paper deals with the LBLOCA that is inserted into the above context.Key conclusion is that regulations need suitable modification,rather than lowering the importance and the role of LBLOCA.Moreover,strengths of emergency core cooling system(ECCS)and containment need a tight link.

Mechanical and fatigue properties of SA508-Ⅳ steel used for nuclear reactor pressure vessels

The mechanical and fatigue properties of SA508-Ⅳ steel with martensite and granular bainite, respectively, were studied. The mechanical tests results showed that the ultimate tensile strength and impact toughness of the specimen with martensite were 830 MPa and 158 J, respectively, and those of the specimen with granular bainite were 811 MPa and 115 J, respectively. The former had higher tensile strength and impact toughness than the latter. The impact tests results showed that the former belonged to typical dimple fracture, while the latter belonged to brittle fracture. The fatigue tests results showed that the fatigue life of the specimen with martensite was 2717 cycles, and that of the specimen with granular bainite was 1545 cycles under the strain amplitude of ± 0.45%. The specimen with martensite had fewer crack initiation points, narrower fatigue striations separation, and larger volume fraction of high-angle grain boundaries than the latter. The fewer crack initiation points meant fewer fatigue cracks, the narrower fatigue striations separation meant slower crack propagation rate, and the larger volume fraction of high-angle grain boundaries could more effectively hinder fatigue crack propagation. Based on these facts, the fatigue life of the specimen with martensite was higher than that of the specimen with granular bainite.

Design and Thermal Analysis of the Large Fire Door for AP1000 Nuclear Reactor

The large fire door is the key component to ensure the effectiveness of fire zone in AP1000 nuclear reactor.According to the fire design requirements and design criteria,the global structure of the large fire door is designed.Based on the designed structure,the thermal mathematical model of the large fire door is established.Based on the solid heat transfer theory,the multi-layer heat transfer theory and integrated heat transfer theory,the differential equations of heat conduction,initial conditions,and boundary conditions are determined.Thermal analysis for the fire door leaf and the closure is carried out by using the method of numerical simulation.Results show that:considering the thermal load,the whole structure of the large fire door can meet the fire resistance limit of 3 hours and the design is reasonable and feasible.This study provides theory basis for the design of the large fire door.

Turbulent vortex trains in narrow square arrayed rod bundles of a dual-cooled nuclear reactor

The dual-cooled nuclear reactor is currently considered for improving the designs of current/future nuclear reactors. Investigation of the thermal-hydraulic characteristics of the nuclear reactor via experiments is essential for commercializing the dual-cooled nuclear reactor. In this paper, the turbulent flow in square arrayed six-rod bundles in the form of magnified copies of the dual-cooled and current OPR-1000 nuclear reactor is experimentally investigated by means of hot-wire anemometry and smoke-wire generation methods. Vortex trains which do not exist in an ordinary reactor subchannel are presented in the subchannel of the dual-cooled reactor. The vortices are induced by a span-wise velocity gradient. This flow pulsation phenomenon increases the inter-channel mixing of the subchannel. To understand the periodic feature of the pulsation, axial/cross velocities are measured and the periodic characteristic frequencies are obtained by a Fast Fourier Transform (FFT) analysis. The peak frequency that represents the quasi-periodic pulsation of the flow is increased with an increase in the axial velocity while the wavelength of the pulsation remains constant within a tested range of the Reynolds number (9000 51000). The vortex trains are highly synchronized with each other, as confirmed by means of visualization.

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.

A Semilinear Parabolic System Arising in the Nuclear Reactors

This note studies the following initial boundary value problem, which arises in the nuclear

GLOBAL SOLUTION,ASYMPTOTIC BEHAVIOR AND BLOW-UP PROBLEMS FOR A MODEL OF NUCLEAR REACTORS

The existence of global solution and the blow-up problem for a model of nuclear reactorsare discussed by using the upper-lower solution and energy estimate methods; asymptoticbehavior of global solution is also discussed with the aid of L_p estimate and semigroupmethod for this model. Nice results, which explain the phenomenon of nuclear reactorsbetter, are obtained.