DEMO Operation Research Articles

Simulation of elevated temperature solid sorbent CO2 capture for pre-combustion applications using computational fluid dynamics

CO2 adsorption is one of the warm gas cleanup technologies under development for integrated gasification combined cycle (IGCC) applications. Computational Fluid Dynamics (CFD) can be a practical and powerful tool for reactor design for and optimization of the CO2 adsorption process. In this present work, CFD simulation was developed in commercially available ANSYS Fluent software and validated for solid sorbent CO2 capture to investigate pressure swing adsorption (PSA) for CO2 separation from syngas with all the auxiliary operating steps. The adsorption equilibrium and kinetics were incorporated into ANSYS by user defined functions (UDFs) for source terms in Navier–Stokes equations. The CFD model well predicted the CO2 breakthrough curve and temperature change. It was shown that in demo reactor operation, at the end of adsorption step, only half of the sorbent was loaded with CO2, while most of the loaded CO2 was released during the desorption step. Further optimization of sorbent packing and cycle operation can be done with assistance of this CFD model to maximize bed loading, and used to design the commercial size reactors.

Progress in high heat flux testing of European DEMO divertor mock-ups

In the framework of the DEMO divertor project of EUROfusion an extensive R&D program has been carried out to develop advanced design concepts for hot water cooled divertor targets. These plasma-facing components made of W blocks as plasma facing material and CuCrZr as cooling tubes should allow a reliable DEMO operation for 2 h long pulses and maximum heat fluxes up to 20 MW/m². Compared to ITER, the operation at the higher coolant temperature of 150 °C, the longer required lifetime, and the significantly higher neutron fluence are the design challenges exceeding the current extent of experience. In this study we present high heat flux test results of a total of 34 tested W monoblock mock-ups. 14 of them where tested with 20 MW/m² heat flux at DEMO hot water cooling conditions up to now. All tests were performed in the HHF test facility GLADIS from 2016 to 2018. Furthermore, we discuss results from the post exposure investigation of selected concepts of the first R&D phase.

Conceptual designs of PHTS, ESS and PCS for DEMO BoP with helium cooled BB concept

The state of EU DEMO Balance of Plant as for the end of 2016 is described in this paper. Starting with 6 outboard and 3 inboard cooling loops, the thermal energy of the helium is transferred by an intermediate heat exchanger to the solar salt, which transports it out of the tokamak building to the energy storage system, which is actually an industrial 2-tanks design. Heat is transferred from solar salt to the feedwater in the steam generator where it boils and becomes superheated before entering the steam turbine. The steam turbine design follows state-of-the-art concept, but it is not finally optimized, since further DEMO BoP design changes still cannot be excluded. The other 3 usable heat sources (two from divertor and one from vacuum vessel) are used to heat-up the feedwater on the return line to the steam generator, thus contributing to the efficient utilization of heat produced during DEMO operation. Taking into account pulse/dwell times of 2.0/0.5 h, DEMO can operate permanently under ∼80% power thus allowing flexible and continuous plant operation. EBSILON® results are presented for steady-state conditions including both pulsed and dwell time periods.

Path-oriented early reaction to approaching disruptions in ASDEX Upgrade and TCV in view of the future needs for ITER and DEMO

Routine reaction to approaching disruptions in tokamaks is currently largely limited to machine protection by mitigating an ongoing disruption, which remains a basic requirement for ITER and DEMO []. Nevertheless, a mitigated disruption still generates stress to the device. Additionally, in future fusion devices, high-performance discharge time itself will be very valuable. Instead of reacting only on generic features, occurring shortly before the disruption, the ultimate goal is to actively avoid approaching disruptions at an early stage, sustain the discharges whenever possible and restrict mitigated disruptions to major failures. Knowledge of the most relevant root causes and the corresponding chain of events leading to disruption, the disruption path, is a prerequisite. For each disruption path, physics-based sensors and adequate actuators must be defined and their limitations considered. Early reaction facilitates the efficiency of the actuators and enhances the probability of a full recovery. Thus, sensors that detect potential disruptions in time are to be identified. Once the entrance into a disruption path is detected, we propose a hierarchy of actions consisting of (I) recovery of the discharge to full performance or at least continuation with a less disruption-prone backup scenario, (II) complete avoidance of disruption to sustain the discharge or at least delay it for a controlled termination and, (III), only as last resort, a disruption mitigation. Based on the understanding of disruption paths, a hierarchical and path-specific handling strategy must be developed. Such schemes, testable in present devices, could serve as guidelines for ITER and DEMO operation. For some disruption paths, experiments have been performed at ASDEX Upgrade and TCV. Disruptions were provoked in TCV by impurity injection into ELMy H-mode discharges and in ASDEX Upgrade by forcing a density limit in H-mode discharges. The new approach proposed in this paper is discussed for these cases. For the H-mode density limit sensors used so far react too late. Thus a plasma-state boundary is proposed, that can serve as an adequate early sensor for avoiding density limit disruptions in H-modes and for recovery to full performance.

Plasma–wall interaction studies within the EUROfusion consortium: progress on plasma-facing components development and qualification

The provision of a particle and power exhaust solution which is compatible with first-wall components and edge-plasma conditions is a key area of present-day fusion research and mandatory for a successful operation of ITER and DEMO. The work package plasma-facing components (WP PFC) within the European fusion programme complements with laboratory experiments, i.e. in linear plasma devices, electron and ion beam loading facilities, the studies performed in toroidally confined magnetic devices, such as JET, ASDEX Upgrade, WEST etc. The connection of both groups is done via common physics and engineering studies, including the qualification and specification of plasma-facing components, and by modelling codes that simulate edge-plasma conditions and the plasma–material interaction as well as the study of fundamental processes. WP PFC addresses these critical points in order to ensure reliable and efficient use of conventional, solid PFCs in ITER (Be and W) and DEMO (W and steel) with respect to heat-load capabilities (transient and steady-state heat and particle loads), lifetime estimates (erosion, material mixing and surface morphology), and safety aspects (fuel retention, fuel removal, material migration and dust formation) particularly for quasi-steady-state conditions. Alternative scenarios and concepts (liquid Sn or Li as PFCs) for DEMO are developed and tested in the event that the conventional solution turns out to not be functional. Here, we present an overview of the activities with an emphasis on a few key results: (i) the observed synergistic effects in particle and heat loading of ITER-grade W with the available set of exposition devices on material properties such as roughness, ductility and microstructure; (ii) the progress in understanding of fuel retention, diffusion and outgassing in different W-based materials, including the impact of damage and impurities like N; and (iii), the preferential sputtering of Fe in EUROFER steel providing an in situ W surface and a potential first-wall solution for DEMO.

Analysis of the secondary circuit of the DEMO fusion power plant using GateCycle

Conceptual activities on the DEMO fusion power plant design are progressing in Europe under the lead of the EUROfusion Consortium. According to the current EU DEMO plant design, the Primary Heat Transfer System (PHTS) transfers heat from the nuclear heat sources, i.e. breeding blanket, divertor and (optionally) vacuum vessel, to the Power Conversion System (PCS) responsible for generating electric energy. To mitigate issues related to the pulsed DEMO operation, adding the Energy Storage System (ESS) filled with molten salt, between the PHTS and PCS, has been proposed. One of the four candidate options for the realization of the blanket and the related PHTS is the Water-Cooled Lithium-Lead Breeding Blanket (WCLL). In the present work a detailed GateCycle model of the DEMO PCS, for the option WCLL, with the ESS was created and its operation at the nominal conditions (plasma burn) and at the reduced heating power (dwell period) was studied. It was demonstrated, that during the dwell period the amount of heat provided from the ESS to the considered cycle can be safely reduced down to 50% of its nominal value. This is an important information for the ESS designers which should help in its proper sizing.

Effect of steady state and ELMy heat loads on the PFC of DEMO

It was shown that the edge localized modes (ELM) in steady state DEMO operation can pose a severe tread to the first wall water cooled blanket module due to misalignment Igitkhanov and Bazylev [1]. In this work we estimate the ELM effect on the FW blanket module with helium as a coolant. The characteristics of ELMs in DEMO are derived by using the scaling arguments from experiments, the prediction of peeling-ballooning mode theory and by extrapolating results made for ITER. The temperatures of the module materials are calculated for different inlet coolant conditions by using the MEMOS code Bazylev et. al [2]. Calculations show the evolution of the maximum temperatures in the W, EUROFER, Cu alloy and the stainless-steel tube for different level of surface inclination to the magnetic field line. The level of surface inclination to the magnetic field is varied to find when the FW blanket module materials remain within a proper temperature range, limited by crystallization and ductility.

Dynamic thermal-hydraulic modelling of the EU DEMO HCPB breeding blanket cooling loops

A global, system-level thermal-hydraulic model of the EU DEMO tokamak fusion reactor is currently under development and implementation in a suitable software at Politecnico di Torino, including the relevant heat transfer and fluid dynamics phenomena, which affect the performance of the different cooling circuits and components and their integration in a consistent design. The model is based on an object-oriented approach using the Modelica language, which easily allows to preserve the high modularity required at this stage of the design. The first module of the global model will simulate the blanket cooling system and will be able to investigate different coolant options and different cooling schemes, to be adapted to the different blanket systems currently under development in the Breeding Blanket (BB) project. The paper presents the Helium-Cooled Pebble Bed (HCPB) module of the EU DEMO blanket cooling loops system model. The model is used to compare different schemes for the cooling of the different components of the HCPB BB, and to suggest improvements aimed at optimizing the pumping power required by the cooling system. The model is then used to analyse a pulsed scenario, characteristic of the EU DEMO operation.

Effect of heat loads on the plasma facing components of demo

In this paper we analyse a thermo-hydraulic performance of the first wall blanket module during the stationary DEMO operation with the edge localized mode (ELM). Heat loads are estimated based on scaling arguments and predictions from the peeling-ballooning ELM model. Effect of parallel heat fluxes intersecting with the first wall panels and avoidance of overheating by inclination of the panels are considered. The material temperatures of the W/EUROFER sandwich type module with water cooling stainless stell tube and Cu alloy compliance embedded into EUROFER is calculated by using the MEMOS code. The calculations were carried out indicating the required geometric parameters as well as the cooling conditions which allow keeping materials temperatures within allowable engineering limits. Effect of inclination of the first wall plates to avoid the misalignment problems is considered.

Effect of Transient Thermal Loads on Tungsten Monoblock Module in DEMO

Thermo-hydraulic analyses of the tungsten mono-block divertor module with a water cooling tube made from a diamond/copper composite (DCC) as a laminate and a martensitic steel EUROFER against the power loadings expecting in DEMO operation is presented. Thermal analysis is carried out by using the code MEMOS, which simulates W armor damage under the repetitive edge localized modes (ELM) heat impact. Heat transfer to the water coolant is studied for various coolant conditions which allow one to keep the material temperatures within the allowable design limits under neutron irradiation. The thermal performance is analyzed for the DEMO I and DEMO II reactor conditions for un-mitigated and mitigated ELMs. The importance of W vapor shielding effect is discussed.

Blanket/Materials Testing Strategy for FNSF and Its Breeding Potential

In the U.S., the Fusion Nuclear Science Facility (FNSF) is viewed as an essential element of the fusion developmental roadmap. The tritium self-sufficiency, blanket testing, and materials testing are of particular interest since they define a critical element of the FNSF mission. There is a definitive need to breed the majority of, if not all, the tritium required for operation. A staged blanket testing strategy has been developed to test and enhance the blanket performance during each phase of operation. A materials testing module is critically important to include in FNSF to test large specimens of future generations of materials (for blanket, divertor, magnets, etc.) in relevant fusion environment. In this strategy, the test modules play a pivotal role and serve as “forerunners” for more advanced versions of blanket and materials that will validate their characteristics and features to assure the successful operation of DEMO and advanced power plants.

First disruption studies and simulations in view of the development of the DEMO Physics Basis

In the development of the DEMO Physics Basis an important role is played by the prediction of the plasma disruption features and by the evaluation of the electro-magnetic (EM) and thermal loads associated with these events. Indeed, the kind and number of foreseen plasma disruptions drive the development of the DEMO operation scenarios and the design of vessel and in-vessel components.To characterize a plausible macroscopic plasma dynamics during these events, we will carry out an extrapolation from present-day machines of the main parameters characterizing the disruptions: thermal and current quench time, evolution of plasma current, β and li, safety factor limits, halo current fraction and width, radiated heat fraction. In particular, we will focus on extrapolations for the thermal and current quench characteristic times, due to their importance for the subsequent simulations aimed at the evaluation of the EM and thermal loads. The different options for DEMO design will be taken into account and the possible range of variation of the parameters will be estimated.The 2D axysimmetric MAXFEA and the 3D CarMa0NL codes will be used to evaluate the effects of the induced currents and the EM loads during a disruptive event and to analyze the various design options obtained by the PROCESS code. The results of these simulations, modeled as worst expected events, will be used as input for the system level analysis and design of the vessel and relevant in-vessel components. First simulations with CarMa0NL code including 3D structures are here presented to show the significant effects due to the large access ports.

Model improvements for tritium transport in DEMO fuel cycle

DEMO operation requires a large amount of tritium, which is directly produced inside the reactor by means of Li-based breeders. During its production, recovering and purification, tritium comes in contact with large surfaces of hot metallic walls, therefore it can permeate through the blanket cooling structure, reach the steam generator and finally the environment.The development of dedicated simulation tools able to predict tritium losses and inventories is necessary to verify the accomplishment of the accepted tritium environmental releases as well as to guarantee a correct machine operation.In this work, the FUS-TPC code is improved by including the possibility to operate in pulsed regime: results in terms of tritium inventory and losses for three pulsed scenarios are shown. Moreover, the development of a 1-D model considering the radial profile of the tritium generation is described. By referring to the inboard segment on the equatorial axis of the helium-cooled lithium–lead (HCLL) blanket, preliminary results of the 1-D model are illustrated: tritium partial pressure in Li–Pb and tritium permeation in the cooling and stiffening plates by assuming several permeation reduction factor (PRF) values. Future improvements will consider the application of the model to all segments of different blanket concepts.

원격 제어 기능을 포함한 교육용 모바일 로봇 시스템의 설계 및 구현

This paper presents the design and implementation of the educational remote controlled robot system including remote sensing in the embedded environment. The design of sensing information processing, software design and template design mechanism for the programming practice are introduced. LPC1769 ∙제1저자 : 정중수 ∙교신저자 : 정광욱 ∙투고일 : 2015. 2. 13, 심사일 : 2015. 3. 18, 게재확정일 : 2015. 4. 21. * 국립안동대학교 공과대학 정보통신공학과 교수(Dept. of Information & Communication Engineering, Andong National University) ** 구미1대학 정보통신공학과 교수/(주)맨엔텔 대표이사(Dept. of Information & Communication Engineering, Kumi University / CEO of Co, Ltd) ※ “이 논문은 2013년도 안동대학교 특별연구비에 의하여 연구되었음” 34 Journal of The Korea Society of Computer and Information April 2015 using Cortex-M3 core as CPU, LPCXPRESSO as debugging environment, C language as firmware development language and FreeRTOS as OS are used in development environment. The control command is received via RF communication by the server and the robot system which is operated by driving the various sensors. The educational procedure is from robot demo operation program as hands-on practice and then compiling, loading of the basic robot operation program, already supplied. Thereafter the verification is checked by using the basic robot operation to allow demo operation such as hands-on-training procedure. The original protocol is designed via RF communication between server and robot system, and the satisfied performance result is presented by analyzing the robot sensing data processing. ▸

Neutronics requirements for a DEMO fusion power plant

This paper addresses the neutronic requirements a DEMO fusion power plant needs to fulfil for a reliable and safe operation. The major requirement is to ensure Tritium self-sufficiency taking into account the various uncertainties and plant-internal losses that occur during DEMO operation. A further major requirement is to ensure sufficient protection of the superconducting magnets against the radiation penetrating in-vessel components and vessel. Reliable criteria for the radiation loads need to be defined and verified to ensure the reliable operation of the magnets over the lifetime of DEMO. Other issues include radiation induced effects on structural materials such as the accumulated displacement damage, the generation of gases such as helium which may deteriorate the material performance. The paper discusses these issues and their impact on design options for DEMO taking into account results obtained in the frame of European Power Plant Physics and Technology (PPPT) 2013 programme activities with DEMO models employing the helium cooled pebble bed (HCPB), the helium cooled lithium lead (HCLL), and the water-cooled (WCLL) blanket concepts.

DEMO divertor limitations during and in between ELMs

Operation of DEMO in comparison to ITER will be significantly more demanding, as various additional limitations of physical and technical nature have to be respected. In particular a set of extremely restrictive boundary conditions on divertor operation during and in between ELMs will have to be respected. It is of high importance to describe these limitations in order to consider them as early as possible in the ongoing development of the DEMO concept design. This paper extrapolates the existing physics basis on power and particle exhaust to DEMO.In phases between ELMs or with mitigated ELMs surface overheating and W sputtering pose challenging boundary conditions. For attached divertor conditions at 90% total radiation fraction a peak power density of about 15 MW m−2 convected or radiated to the outer divertor is estimated. As this clearly exceeds the tolerable limit, some degree of divertor detachment is regarded as essential for the operation of DEMO. A loss of detachment with a peak power density of more than 30 MW m−2 cannot be tolerated for more than a second before the divertor would suffer from a destructive event. The combination of the limitations on the peak power flux density and W sputtering rate necessitates divertor temperatures less than 4 eV.For uncontrolled ELMs sizes in the order of 100 MJ are estimated. Results on ELM broadening from JET suggest that in DEMO an energy density limit of 0.5 MJ m−2 per ELM is exceeded by a factor of about 8 for a large range of relative ELM sizes. This highlights the necessity of a reactor-relevant ELM control technique for DEMO, which is capable of reducing the maximum size of the energy loss per ELM to the divertor by more than an order of magnitude without a strong reduction of confinement.

Damage and fatigue crack growth of Eurofer steel first wall mock-up under cyclic heat flux loads. Part 1: Electron beam irradiation tests

Recently, the idea of bare steel first wall (FW) is drawing attention, where the surface of the steel is to be directly exposed to high heat flux loads. Hence, the thermo-mechanical impacts on the bare steel FW will be different from those of the tungsten-coated one. There are several previous works on the thermal fatigue tests of bare steel FW made of austenitic steel with regard to the ITER application. In the case of reduced-activation steel Eurofer97, a candidate structural material for the DEMO FW, there is no report on high heat flux tests yet. The aim of the present study is to investigate the thermal fatigue behavior of the Eurofer-based bare steel FW under cyclic heat flux loads relevant to DEMO operation. To this end, we conducted a series of electron beam irradiation tests with heat flux load of 3.5MW/m2 on water-cooled mock-ups with an engraved thin notch on the surface. It was found that the notch root region exhibited a marked development of damage and fatigue cracks whereas the notch-free surface manifested no sign of crack formation up to 800 load cycles. Results of extensive microscopic investigation are reported.

Dynamic Simulation of a 1MWe Concentrated Solar Power Tower Plant System with Dymola®

This paper presents a first vertion of a dynamic model of the 1 MWe Central Receiver System demo plant at Badaling in Beijing with the purpose of improving the performance and reliability of the CSP technology deployed at this plant using dynamic modelling by Dymola® software in the frame of cooperation between EDF R&D and IEECAS (Institute of Electrical Engineering of the Chinese Academy of Sciences). This dynamic model is capable of simulating the dynamic behavior of the entire CSP plant, covering the heliostat field, the superheated steam cavity receiver, the thermal storage system and the power block. In order to validate the model, the demo plant operation data is used. The recorded evolution of the different process parameters during the operation is compared to the results predicted by the dynamic model. According to the results obtained from the operation, the simulation predicts fairly well the main process parameters. The predicted trend of the dynamic behavior of the system is also satisfactory. The results show that the model could be used to support the operation of the entire solar thermal power tower system and help improve the performance of the CSP technology deployed at Badaling.

Plasma Facing Materials Lifetime in Steady-State DEMO Operation

In the steady-state operation the life-time performance of functional and structural materials in fusion reactor DEMO will be limited by several processes such as a sputtering erosion, transient events and neutron irradiation. The design strategy is to determine the structure and coating thicknesses which maximize component lifetime against all lifetime limitations. The sputtering erosion of the first wall tungsten armor layer due to the plasma impact during the steady state DEMO operation is considered here. It is shown that for DEMO conditions the total sputtering erosion of W armor by the charge-exchange DT neutrals could at least reach~1mm during one year of steady-state operation.

임베디드 교육용 라우터 실습장비의 구현

본 논문에서는 교육용 임베디드 라우터 실습장비 설계를 제시하였다. 교육용 임베디드 라우터 실습장비는 사용자가 인터넷 환경에서네트워크 구성과 임베디드 라우터 프로그래밍 실습 교육을 할 수 있도록 설계 및 구현되었다. 정적 및 동적 라우팅 프로토콜과 방화벽을 지원하는 라우터 기술 교육용 시스템 설계를 임베디드 환경에서 이더넷 인터페이스를 바탕으로 제시하였다. 개발 환경으로는 CPU는 PXA255, 디버깅 환경은 ADS 1.2, RTOS는 uC/OS-ii, 개발 언어는 C 언어를 사용하였다. 교육 과정으로는 제공된 정적 및 동적 라우팅 프로토콜과 방화벽 프로그램을 컴파일 및 로딩이후 데모 동작의 ping 처리과정으로 이들 기능의 점검 후 ping 패킷 처리의 성능도 살펴보았다. 이후 데모 기능과 유사하게 동작하는 프로그래밍을 하도록 기술 교육을 단계별로 진행시키도록 하였다. 즉, 정적 및 동적 라우팅 프로토콜과 패킷 필터링에 따른 프로그래밍의 완성 및 이의 검증을 수행하도록 하였다. This paper presents the design of the educational router system. This system is designed and implemented to support network configuration and embedded programming technology of the user on Internet. Not only Static routing protocol but also a kind of dynamic routing protocols such as OSPF and RIP and firewall have been programmed for education based on ethernet interface. ADS 1.2 as debugging environment, uC/OS-ii as RTOS and C language as development language are used. The educational procedures is compile, loading of static routing protocol, a kind of dynamic routing protocols such as OSPF and RIP and firewall program already supplied. Thereafter the verification is checked by using "ping" test to allow for demo operation such as hands-on training procedure. Finally programming procedure similar with demo operation of static routing protocol, a kind of dynamic routing protocols such as OSPF and RIP and packet filtering function is educated step by step.