Assessment and comparison of pulsed and steady-state tokamak power plants
|
|
- Juniper Lamb
- 6 years ago
- Views:
Transcription
1 Assessment and comparison of pulsed and steady-state tokamak power plants Farrokh Najmabadi UC San Diego 21 st International Toki Conference, 28 Novemeber-1 December 2011 Toki, Japan
2 Choice between steady-state and pulsed operation is purely an economic consideration A widely-held belief is that steady-state operation of a tokamak needs a high bootstrap fraction (e.g., > 85%). It requires operation in reverse-shear mode with high β N and a high degree of control of plasma profiles. Thus, steady-state operation requires a major extrapolation from present data base. However, the first steady-state power plant proposals (ARIES-I and SSTR) operated in the 1 st stability regime (monotonic q profile) Both designs had bootstrap fraction ~60-70% Required current-drive powers of 70 MW (SSTR) to MW (ARIES-I & ARIES-I versions). In fact, ARIES-I plasma profiles are very similar to Hybrid mode (sans pedestal) and a high-degree of profile control is NOT required. Thus, the trade-off is between the cost of additional currentdrive power vs issues associated with pulsed operation.
3 Outline I. System-level issues which are generic to any pulsed power plant (e.g., thermal energy storage). II. Tokamak-specific issues: operating points and magnets. III. Engineering design of power components Recent work on high-heat flux components
4 System Level Issues Thermal Energy Storage
5 A pulsed-power plant requires thermal energy storage Connecting a power plant to the grid is NOT a trivial issue: Utilities require a minimum electric power for a plant to stay on the grid. Load balancing requires a slow rate of change in introducing electric power into the grid. Overall, it is extremely expensive to attach an intermittent electric power source to the grid, a steady electric power is required. Large thermal power equipments such as pumps and heat exchangers cannot operate in a pulse mode. For example, the rate of change of temperature in a steam-generator is < 2 o C/min in order to avoid induced stress and boiling instabilities. Overall, a thermal energy storage is needed to ensure a constant thermal power flow to the balance of the plant.
6 The thermal energy storage system is quite massive. During the dwell time (no fusion power), thermal energy storage should supply thermal energy to the power cycle. Stored energy = M c p (T charge -T discharge ) Rate of change of storage temperature, T/ t, is set by the power cycle. Small T/ t leads to a large mass for the storage system with a complicated design to ensure a relatively uniform storage temperature. During the dwell time, fusion core temperature will follow the storage temperature. At the start of the burn phase, fusion core components see a large temperature change from T discharge to operating temperature (> T charge ) which could result in large strains. There is substantial benefit in minimizing (T charge -T discharge ) or the dwell time. Other critical issues include tritium extraction and permeation to energy storage system, power needed for plasma start-up,
7 Pulsar thermal energy storage system Energy accumulated in the outer shield D=during the burn phase Thermal power is extracted from shield and is regulated by mass-flow-rate control during dwell phase Limited storage capability (limited by shield mass and temperature limit) means limited dwell time (< 200 s). This approach requires precise mass flow rate controlled and assumes good coolant mixing and temperature uniformity. Judged by industrial people to be beyond current capabilities. Extension to modern blanket design (such as DCLL)?
8 Thermal energy storage dictates design choices. Thermal energy storage dictates many aspects of the design (including thermal conversion efficiency). In principle, it would be best to produce a credible storage design/power cycle before optimizing the tokamak. Cost of thermal energy storage scales linearly with the dwell time. Minimizing dwell time is important. Efforts to increase pulse length beyond ~20 X dwell time have little benefits. Average plant power already close to burn value, Impact of reducing number of cycle by a factor of two on fatigue issues are small. Allowable stress for 316LN
9 Tokamak-specific Issues
10 Pulsed and steady-state devices optimize in different regimes Steady-state, 1 st stability tokamaks (monotonic q profiles) Require minimization of current drive power Operate at high aspect ratio (to reduce I), maximize bootstrap fraction (εβ p 1) and raise on-axis q Can achieve 60%-70% bootstrap fraction with β N Current-drive power ~ MW. Typically optimizes at A ~ 4-6. Pulsed plasma Pressure (density/temperature) profile sets the achievable plasma β (no control of current profile). Can achieve 30%-40% bootstrap fraction with β N Optimizes at larger plasma current, medium aspect ratio, and higher β.
11 Magnet systems for steady-state devices can be quite simpler For steady-state devices (assuming a long start-up with current-drive assist), TF system can be substantially simpler Typical ARIES magnets consists of TF coils bucked against a bucking cylinder. The overturning forces are reacted against each other through structural caps on the top and bottom of TF coils. Pulsed plasma Lower allowable stress on the structure and lower current-density in the conductor. Torridly continuous structures are avoided as much as possible in order to minimize large eddy currents during start-up o Large Joule losses in cryogenic structures o Reduced coupling of PF coils to the plasma o Impact on plasma equilibrium and position. For the same magnet technology, we found that the field in the coil is lower and magnet cost are substantially higher.
12 Even with shield-storage, we found the steady-state system to be superior. Major Parameters of ARIES and PULSAR Power Plants PULSAR ARIES-I Aspect ratio Plasma major radius (m) Plasma minor radius (m) Toroidal field on axis (T) Toroidal field on the coil (T) Plasma beta 2.8% 1.9% 1.9% Plasma current (MA) Bootstrap fraction Neutron wall loading (MW/m2) Cost of electricity (mills/kwh) Assuming the same plant availability and unit cost for components.
13 Engineering Design of Power Components
14 Engineering design of components in fusion is mostly based on elastic analysis. Conservative design rules allow elastic analysis to be used, e.g. no ratcheting requires P L +P B <3S m where S m =min(1/3 S u, 2/3 S y ). There are many design rules accounting for primary & secondary stress, fracture, fatigue, Design rules for high-temperature operation are incomplete (e.g., interaction of different failure mechanism such as creep & fatigue).
15 Plastic analysis may yield a significantly larger design window for steady-state For plasma-facing components (first wall, divertors) relaxation from local plasticity can significantly expand the design window, enabling operation at a higher heat flux. Pulsed operation reduces the benefit significantly. High temperature creep and creep-fatigue interaction will restrict the operating space even further. More analysis (and data) is needed.
16 We have performed plasto-elastic analysis of several components. Three components were considered: Finger-type divertor Joint between W and Steel for the divertor First wall (high heat flux and transients due to convective SOL). 3D elastic-plastic analysis with thermal stress relaxation (yield) Application of accumulated strain limit (0.5 e ue ) instead of 3S m Birth-to-death modeling (Fabrication steps, operating scenarios, off-normal events) Plans to analyze high temperature creep and creep-fatigue interaction (which will restrict the operating space further).
17 Examples of birth-to-death thermal cycles. Fabrication Cycle fabrication normal operation with shutdowns transients Temperature Heat Flux (gradients) Time FW Operating Cycle with warm shutdown
18 He-cooled W divertor explored in the ARIES Designs T-tube Plates with jet and/or pin-fin cooling Finger Finger/plate combinations
19 Inclusion of yield extends finger divertor limits Elastic analysis,15 MW/m 2 Elasto-plastic analysis,15 MW/m 2 SF= Allowable (3S m ) / Maximum stress SF > 1 to meet the ASME 3S m criterion The minimum elastic safety factor is 0.3 in the armor and 0.9 in the thimble But plastic strain (one cycle) is well within the 1% strain limit (ε ue /2)
20 External transition joints help alleviate one of the more challenging aspects of HHFC s W Ta ODS steel coolant mat l ε 2d ε allowable ODS 0.77% ~1% Ta 0.54% 5-15% W ~0 % ~1% Cu braze
21 Ratcheting leads to strain (damage) accumulation Cold shutdown Warm shutdown (4 time steps per cycle) Design does not meet 3S m criterion. Cold shutdown is the most severe condition (considering 1050 C stress-free temperature). In our case, ratcheting saturates after ~100 cycles. Creep, fatigue, and creep-fatigue interaction are all expected to be more severe under cyclic loading
22 A modified first wall concept using W pins was proposed to better resist transients Goal of 1 MW/m 2 normal, 2 MW/m 2 transient W pins are brazed into ODS steel plates, which are brazed to RAFS cooling channels Pins help resist thermal transients and erosion Similar to micro brush concept developed for the ITER divertor Minor impact on neutronics
23 Inclusion of thermal stress relaxation also extends the first wall performance Maximum ODS XY shear stress at: Room temperature: 20 C Coolant temperature: 385 C Peak temperature: 582 C 3S m ~ 600 / 550 / 400 MPa Elastic analysis σ xy = 885 / 600 / 450 MPa Plastic analysis σ xy = 460 / 200 / 90 MPa
24 Highlights The trade-off is between the cost of additional current-drive power vs issues associated with pulsed operation. Thermal energy storage is needed. It dictates many aspects of the design. It would be best to produce a credible storage design/power cycle before optimizing the tokamak. Efforts to increase pulse length beyond ~20 X dwell time have little benefits. Pulsed-plasma and steady-state plants operate at different plasma operating regimes. Substantial simplification in TF design and capabilities for long, non-inductive start-up Plasto-elastic analysis of plasma-facing components indicate a larger operating window for steady-state operation.
25 Thank you!
CONCLUSIONS OF THE ARIES AND PULSAR STUDIES: DIRECTIONS FOR AN ATTRACTIVE TOKAMAK POWER PLANT
CONCLUSIONS OF THE ARIES AND PULSAR STUDIES: DIRECTIONS FOR AN ATTRACTIVE TOKAMAK POWER PLANT R. W. Conn, F. Najmabadi for The ARIES Team DOE Headquarters, Germantown May 18, 1994 ARIES Is a Community-Wide
More informationCOMPARISON OF STEADY-STATE AND PULSED-PLASMA TOKAMAK POWER PLANTS
COMPARISON OF STEADY-STATE AND PULSED-PLASMA TOKAMAK POWER PLANTS F. Najmabadi, University of California, San Diego and The ARIES Team IEA Workshop on Technological Aspects of Steady State Devices Max-Planck-Institut
More informationOVERVIEW OF THE ARIES AND PULSAR STUDIES
OVERVIEW OF THE ARIES AND PULSAR STUDIES F. Najmabadi, R. W. Conn, University of California, San Diego and The ARIES Team ISFNT-3 University of California, Los Angeles June 27 July 1, 1994 ARIES Is a Community-Wide
More informationOverview of ARIES ACT-1 Study
Overview of ARIES ACT-1 Study Farrokh Najmabadi Professor of Electrical & Computer Engineering Director, Center for Energy Research UC San Diego and the ARIES Team Japan-US Workshop on Fusion Power Plants
More informationTHE ARIES-I TOKAMAK REACTOR STUDY
THE ARIES-I TOKAMAK REACTOR STUDY Farrokh Najmabadi, Robert W. Conn, and The ARIES Team 16th SOFT London, September 3-7, 1990 ARIES Is a Community-Wide Study U. W. UCLA ANL U. IL. FEDC ORNL RPI ARIES GA
More informationTHE ARIES TOKAMAK REACTOR STUDIES
THE ARIES TOKAMAK REACTOR STUDIES Farrokh Najmabadi for The ARIES Team Fusion Power Associates Symposium Pleasanton, CA, April 9-10, 1992 ARIES Is a Community-Wide Study ANL UCLA GA MIT LANL PPPL ARIES
More informationOverview of the ARIES Fusion Power Plant Studies
Overview of the ARIES Fusion Power Plant Studies Farrokh Najmabadi IAEA Technical Committee Meeting on Fusion Power Plant Studies March 24-28, 1998 Culham, United Kingdom The ARIES Team Has Examined Several
More informationAdvanced Study of a Tokamak Transmutation System
Abstract Advanced Study of a Tokamak Transmutation System L. J. Qiu, Y. C. Wu, B. Wu, X.P. Liu, Y.P. Chen, W.N. Xu, Q.Y. Huang Institute of Plasma Physics, Chinese Academy of Sciences P.O. Box 1126, Hefei,
More informationImproved Design of a Helium- Cooled Divertor Target Plate
Improved Design of a Helium- Cooled Divertor Target Plate By X.R. Wang, S. Malang, R. Raffray and the ARIES Team ARIES-Pathway Meeting University of Wisconsin, Madison April 23-24, 2009 Critical Design
More informationDevelopment Scenario of Tokamak Reactor for Early Demonstration of Electric Power Generation
Development Scenario of Tokamak Reactor for Early Demonstration of Electric Power Generation US/Japan Workshop on Power Plant Studies and Related Advanced Technologies With EU Participation 24-25 January
More informationTOROIDAL REACTOR DESIGNS AS A FUNCTION OF ASPECT RATIO
TOROIDAL REACTOR DESIGNS AS A FUNCTION OF ASPECT RATIO C.P.C. Wong, J.C. Wesley, R.D. Stambaugh, E.T. Cheng General Atomics, San Diego, California TSI Research Inc., Solana Beach, California e-mail contact
More informationConceptual design of a demonstration reactor for electric power generation
1 FT/P7-4 Conceptual design of a demonstration reactor for electric power generation Y. Asaoka 1), R. Hiwatari 1), K. Okano 1), Y. Ogawa 2), H. Ise 3), Y. Nomoto 3), T. Kuroda 3), S. Mori 3), K. Shinya
More informationToroidal Reactor Designs as a Function of Aspect Ratio
Toroidal Reactor Designs as a Function of C.P.C. Wong ), J.C. Wesley ), R.D. Stambaugh ), E.T. Cheng ) ) General Atomics, San Diego, California ) TSI Research Inc., Solana Beach, California e-mail contact
More informationDEMO Concept Development and Assessment of Relevant Technologies
1 FIP/3-4Rb DEMO Concept Development and Assessment of Relevant Technologies Y. Sakamoto, K. Tobita, H. Utoh, N. Asakura, Y. Someya, K. Hoshino, M. Nakamura, S. Tokunaga and the DEMO Design Team Japan
More informationIntegrated System Level Simulation and Analysis of DEMO with Apros. Sami Kiviluoto
Integrated System Level Simulation and Analysis of DEMO with Apros Sami Kiviluoto 3.11.2016 DEMO modelling project Fortum joined FinnFusion consortium in the fall 2015 EUROfusion WPPMI project (Plant Level
More informationOverview of the ARIES Program
Overview of the ARIES Program Farrokh Najmabadi University of California San Diego Presentation to: ARIES Program Peer Review August 18, 2000 UC San Diego Electronic copy: http://aries.ucsd.edu/najmabadi/talks/
More informationStatus of Fusion Research
Status of Fusion Research Farrokh Najmabadi Prof. of Electrical Engineering Director of Center for Energy Research UC San Diego NCSU Seminar North Carolina September 2, 2010 World uses (& needs) a lot
More informationStatus Report and Documentation of DCLL Design
Status Report and Documentation of DCLL Design He primary and secondary loops footprint at TCWS DCLL design evolution DCLL, DEMO inboard routing assessment Documentation of DCLL design Clement Wong, Dick
More informationPOSSIBILITY OF THE HE-COOLED SIC-COMPOSITE DIVERTOR
POSSIBILITY OF THE HE-COOLED SIC-COMPOSITE DIVERTOR X.R. Wang 1, S. Malang 2, M. S. Tillack 1 1 University of California, San Diego, CA 2 Fusion Nuclear Technology Consulting, Germany ARIES-Pathways Project
More informationARIES Systems Code Development, Visualization and Application
ARIES Systems Code Development, Visualization and Application Lane Carlson 1, Mark Tillack 1, Farrokh Najmabadi 1 1 University of California San Diego Center for Energy Research La Jolla, CA, USA lcarlson@ucsd.edu
More informationThe ARIES-AT advanced tokamak, Advanced technology fusion power plant
Fusion Engineering and Design 80 (2006) 3 23 The ARIES-AT advanced tokamak, Advanced technology fusion power plant Farrokh Najmabadi a,, The ARIES Team: A. Abdou b, L. Bromberg c, T. Brown d, V.C. Chan
More informationEU DEMO Design Point Studies
EU DEMO Design Point Studies R. Kemp 1, D. J. Ward 1, G. Federici 2, R. Wenninger 2,3 and J. Morris 1 1 CCFE, Culham Science Centre, Oxfordshire OX14 3DB, United Kingdom 2 EFDA PPPT, Boltzmannstr.2, Garching
More informationARIES-ST: A Spherical Torus Fusion Power Plant
ARIES-ST: A Spherical Torus Fusion Power Plant Farrokh Najmabadi University of California, San Diego, La Jolla, CA, United States of America 9 th Course on Technology of Fusion Reactors 26 July 1 August
More informationMagnetohydrodynamics (MHD) III
Magnetohydrodynamics (MHD) III Yong-Su Na National Fusion Research Center POSTECH, Korea, 8-10 May, 2006 Review II 1. What is Stability? 2. MHD Instability Interchange Mode Flux Tube Instabilities 3. Formulation
More informationA Pilot Plant as the Next Step toward an MFE Demo, )
A Pilot Plant as the Next Step toward an MFE Demo, ) George H. NEILSON, David A. GATES, Charles E. KESSEL, Jonathan E. MENARD, Stewart C. PRAGER, Steven D. SCOTT, James R. WILSON and Michael C. ZARNSTORFF
More informationDisruptions in ITER: Major Catastrophe or Minor Annoyance? Sarah Angelini April 21, 2011
Disruptions in ITER: Major Catastrophe or Minor Annoyance? Sarah Angelini April 21, 2011 Disruptions Types of Disruptions Time Scales from the IDDB Thermal Quench Current Quench DINA Simulation Results
More informationDesign and Technology Development of Solid Breeder Blanket Cooled by Supercritical Water in Japan
Design and Technology Development of Solid Breeder Blanket Cooled by Supercritical Water in Japan M. Enoeda, Y. Kosaku, T. Hatano, T. Kuroda, N. Miki, T. Honma and M. Akiba Japan Atomic Energy Research
More informationDESIGN OPTIMIZATION OF HIGH-PERFORMANCE HELIUM-COOLED DIVERTOR PLATE CONCEPT
DESIGN OPTIMIZATION OF HIGH-PERFORMANCE HELIUM-COOLED DIVERTOR PLATE CONCEPT X.R. Wang a, S. Malang b, A.R. Raffray a and the ARIES Team a Center for Energy Research, University of California, San Diego,
More informationDesign and Development of Lower Divertor for JT-60SA
1 FTP/P1-29 Design and Development of Lower Divertor for JT-60SA S. Sakurai, H. Higashijima, H. Kawashima, Y. K. Shibama, T. Hayashi, H. Ozaki, K. Shimizu, K. Masaki, K. Hoshino, S. Ide, K. Shibanuma,
More informationHigh performance blanket for ARIES-AT power plant
Fusion Engineering and Design 58 59 (2001) 549 553 www.elsevier.com/locate/fusengdes High performance blanket for ARIES-AT power plant A.R. Raffray a, *, L. El-Guebaly b, S. Gordeev c, S. Malang c, E.
More informationDESIGN, FABRICATION, INSTALLATION AND TESTING OF IN-VESSEL CONTROL COILS FOR DIII D
GA A24056 DESIGN, FABRICATION, INSTALLATION AND TESTING OF IN-VESSEL CONTROL COILS FOR DIII D by P.M. ANDERSON, C.B. BAXI, A.G. KELLMAN, E.E. REIS, and J.I. ROBINSON OCTOBER 2002 DISCLAIMER This report
More informationSimulation of Power Exhaust in Edge and Divertor of the SlimCS Tokamak Demo Reactor
J. Plasma Fusion Res. SERIES, Vol. 9 (2010) Simulation of Power Exhaust in Edge and Divertor of the SlimCS Tokamak Demo Reactor Nobuyuki ASAKURA, Katsuhiro SHIMIZU, Hisato KAWASHIMA, Kenji TOBITA and Tomonori
More informationARIES: Fusion Power Core and Power Cycle Engineering
ARIES: Fusion Power Core and Power Cycle Engineering The ARIES Team Presented by A. René Raffray ARIES Peer Review Meeting University of California, San Diego ARIES: Fusion Power Core and Power Cycle Engineering/ARR
More informationLOSS OF COOLANT ACCIDENT AND LOSS OF FLOW ACCIDENT ANALYSIS OF THE ARIES-AT DESIGN
LOSS OF COOLANT ACCIDENT AND LOSS OF FLOW ACCIDENT ANALYSIS OF THE ARIES-AT DESIGN E. A. Mogahed, L. El-Guebaly, A. Abdou, P. Wilson, D. Henderson and the ARIES Team Fusion Technology Institute University
More informationCERAMIC BREEDER BLANKET FOR ARIES-CS
CERAMIC BREEDER BLANKET FOR ARIES-CS A.R. Raffray 1, S. Malang 2, L. El-Guebaly 3, X. Wang 4, and the ARIES Team 1 Mechanical and Aerospace Engineering Department and Center for Energy Research, 460 EBU-II,
More informationEX/9-1 Progress in Preparing Scenarios for ITER Operation
EX/9-1 Progress in Preparing Scenarios for ITER Operation George Sips (JET-EFDA, UK) G. Giruzzi, S. Ide, C. Kessel, T. Luce, J. Snipes, J. Stober For the IOS-TG of the ITPA FEC 2014, St Petersburg, Russia
More informationGA A22436 CREEP-FATIGUE DAMAGE IN OFHC COOLANT TUBES FOR PLASMA FACING COMPONENTS
GA A22436 CREEP-FATIGUE DAMAGE IN OFHC COOLANT TUBES FOR PLASMA FACING COMPONENTS by E.E. REIS and R.H. RYDER OCTOBER 1996 GA A22436 CREEP-FATIGUE DAMAGE IN OFHC COOLANT TUBES FOR PLASMA FACING COMPONENTS
More informationTHE ARIES research team continues to develop integrated
552 IEEE TRANSACTIONS ON PLASMA SCIENCE, VOL. 40, NO. 3, MARCH 2012 Development, Visualization, and Application of the ARIES Systems Code Lane Carlson, Mark Tillack, Farrokh Najmabadi, and Charles Kessel
More informationASSESSMENT OF TOKAMAK PLASMA OPERATION MODES AS FUSION POWER PLANTS: THESTARLITESTUDY
ASSESSMENT OF TOKAMAK PLASMA OPERATION MODES AS FUSION POWER PLANTS: THESTARLITESTUDY Farrokh Najmabadi Dept. of Electrical & Computer Eng. and Fusion Energy Research Program, University of California,
More informationDesign of Solid Breeder Test Blanket Modules in JAERI
Design of Solid Breeder Test Blanket Modules in JAERI Presented by: S. Suzuki, Blanket Engineering Lab., Japan Atomic Energy Research Institute, JAERI Contents 1. Outline of blanket development in JAERI
More informationStudies of Impurity Seeding and Divertor Power Handling in Fusion Reactor
1 FIP/P8-11 Studies of Impurity Seeding and Divertor Power Handling in Fusion Reactor K. Hoshino 1, N. Asakura 1, K. Shimizu 2 and S. Tokunaga 1 1 Japan Atomic Energy Agency, Rokkasho, Aomori 039-3212
More informationSummary of Major Features of ARIES- ST and ARIES-AT Blanket Designs
Summary of Major Features of ARIES- ST and ARIES-AT Blanket Designs Presented by A. René Raffray University of California, San Diego with the Contribution of the ARIES Team and S. Malang APEX Meeting,
More informationARIES-AT BLANKET AND DIVERTOR
ARIES-AT BLANKET AND DIVERTOR A. R. Raffray, M.S.Tillack, X. Wang L. El-Guebaly, I. Sviatoslavsky S. Malang University of California, San Diego University of Wisconsin Forschungszentrum Karlsruhe EBU-II,
More informationResearch and Development Status of Reduced Activation Ferritic/Martensitic Steels Corresponding to DEMO Design Requirement
Research and Development Status of Reduced Activation Ferritic/Martensitic Steels Corresponding to DEMO Design Requirement Hiroyasu Tanigawa 1, Hisashi Tanigawa 1, M. Ando 1, S. Nogami 2, T. Hirose 1,
More informationDCLL Blanket for ARIES-AT: Major Changes to Radial Build and Design Implications
DCLL Blanket for ARIES-AT: Major Changes to Radial Build and Design Implications L. El-Guebaly Fusion Technology Institute UW - Madison ARIES-Pathways Project Meeting December 12-13, 2007 Georgia Tech
More informationPlasma Scenarios and Control
Plasma Scenarios and Control R. J. Hawryluk Presented at the SECOND IAEA DEMO PROGRAMME WORKSHOP December 19, 2013 What is the Difference in the Scenario and Control Requirements Between ITER and DEMO?
More information5 th INTERNATIONAL CONFERENCE ON THE FRONTIERS OF PLASMA PHYSICS AND TECHNOLOGY April 21, 2011, Singapore
5 th INTERNATIONAL CONFERENCE ON THE FRONTIERS OF PLASMA PHYSICS AND TECHNOLOGY April 21, 2011, Singapore Outline Early history the underpinnings in Basic Studies Tokamak program Aditya and SST-1 - Some
More informationDEVELOPMENT OF PHYSICS AND ENGNEERING DESIGNS FOR JAPAN S DEMO CONCEPT
Y. SAKAMOTO et al. DEVELOPMENT OF PHYSICS AND ENGNEERING DESIGNS FOR JAPAN S DEMO CONCEPT Y. SAKAMOTO National Institutes for Quantum and Radiological Science and Technology, Fusion Energy Research Development
More information2. Fusion Engineering Research Project
2. Fusion Engineering Research Project Fusion Engineering Research Project (FERP) started in FY2010 at NIFS. Along with a conceptual design of the helical fusion reactor FFHR-d1, the project is conducting
More informationEUROPEAN FUSION DEVELOPMENT AGREEMENT. PPCS Reactor Models. 9 th Course on Technology of Fusion Tokamak Reactors
PPCS Reactor Models 9 th Course on Technology of Fusion Tokamak Reactors International School of Fusion Reactor Technology - 2004 David Maisonnier EFDA CSU Garching (david.maisonnier@tech.efda.org) PPCS
More informationHAPL Blanket Strategy
HAPL Blanket Strategy A. René Raffray UCSD With contributions from M. Sawan and I. Sviatoslavsky UW HAPL Meeting Georgia Institute of Technology Atlanta, GA HAPL meeting, G.Tech. 1 Outline Background Strategy
More informationDesign and analysis of the ARIES-ACT1 fusion core
Design and analysis of the ARIES-ACT1 fusion core M. S. Tillack, X. R. Wang, D. Navaei, H. H. Toudeshki, A. F. Rowcliffe, F. Najmabadi and the ARIES Team University of California San Diego 9500 Gilman
More informationTechnical Challenges on the Path to DEMO and the Strategy of EFDA on the Power Plant Physics and Technology
Technical Challenges on the Path to DEMO and the Strategy of EFDA on the Power Plant Physics and Technology Gianfranco Federici Head of PPPT Department HTS Fusion Conductor Workshop, KIT Karlsruhe 26 27.05.2011
More informationThe ITER Blanket System Design Challenge
The ITER Blanket System Design Challenge Presented by A. René Raffray Blanket Section Leader; Blanket Integrated Product Team Leader ITER Organization, Cadarache, France With contributions from B. Calcagno
More informationDevelopment and Application of System Analysis Program for Parameters Optimization and Economic Assessment of Fusion Reactor (SYSCODE)
Development and Application of System Analysis Program for Parameters Optimization and Economic Assessment of Fusion Reactor (SYSCODE) Presented By Dehong Chen Contributed by FDS Team Key Laboratory of
More informationEU Designs and Efforts on ITER HCPB TBM
EU Designs and Efforts on ITER HCPB TBM L.V. Boccaccini Contribution: S. Hermsmeyer and R. Meyder ITER TBM Project Meeting at UCLA February 23-25, 2004 UCLA, February 23rd, 2004 EU DEMO and TBM L.V. Boccaccini
More informationConcept Development of DEMO in Japan
US-JPN WS on Fusion Power Reactor 2014.3.13, UCSD Concept Development of DEMO in Japan JAEA Kenji Tobita 2/22 OUTLINE 1. Circumstances Surrounding DEMO 2. DEMO Design Activity 3. Safety Study 1. Circumstances
More informationImpact of Advanced Technologies on Fusion Power Plant Characteristics: The ARIES-AT Study
Impact of Advanced Technologies on Fusion Power Plant Characteristics: The ARIES-AT Study Farrokh Najmabadi University of California, San Diego, La Jolla, CA, United States of America ANS 14 th Topical
More informationWe can describe a simple power balance for a fusion power plant in steady state with the following definitions,
Producing Electricity in a Fusion Nuclear Science Facility or Similar C. E. Kessel, PPPL 1. Introduction The ultimate goal of fusion plasma and fusion nuclear science is the construction and operation
More informationEU considerations on Design and Qualification of Plasma Facing Components for ITER
EU considerations on Design and Qualification of Plasma Facing Components for ITER Patrick Lorenzetto, F4E Barcelona with inputs from B. Riccardi (F4E), V. Barabash and M. Merola (ITER IO) on Readiness
More informationFusion Nuclear Science & Technology. PFC Concepts and R&D towards DEMO Divertor Technology
Fusion Nuclear Science & Technology PFC Concepts and R&D towards DEMO Divertor Technology D.L. Youchison & R.E. Nygren Sandia National Laboratories A.R. Raffray Univ. of California San Diego Los Angeles,
More informationDevelopment of High Heat Flux Components in JAERI
US-Japan Workshop on Fusion High Power Density Device and Design, UCLA, February, 16-19, 1999 Development of High Heat Flux Components in JAERI K. Ezato, NBI Heating Lab., Dept. Of Fusion Engineering,
More informationConcept of power core components of the SlimCS fusion DEMO reactor
Concept of power core components of the SlimCS fusion DEMO reactor K. Tobita, H. Utoh, Y. Someya, H. Takase, N. Asakura, C. Liu and the DEMO Design Team Japan Atomic Energy Agency, Naka, Ibaraki-ken, 311-0193
More informationFUSION TECHNOLOGY INSTITUTE
FUSION TECHNOLOGY INSTITUTE Apollo-L2, An Advanced Fuel Tokamak Reactor Utilizing Direct Conversion W I S C O N S I N G.A. Emmert, G.L. Kulcinski, J.P. Blanchard, L.A. El-Guebaly, H.Y. Khater, J.F. Santarius,
More informationCyclic Stress-Strain Curve for Low Cycle Fatigue Design and Development of Small Specimen Technology
1 PD/P8-2 Cyclic Stress-Strain Curve for Low Cycle Fatigue Design and Development of Small Specimen Technology A. Nishimura 1, S. Nogami 2, E. Wakai 3 1 National Institute for Fusion Science (NIFS), Toki,
More informationReflections on Fusion Chamber Technology and SiC/SiC Applications Mohamed Abdou UCLA
Reflections on Fusion Chamber Technology and SiC/SiC Applications Mohamed Abdou UCLA Presented at CREST Conference, Kyoto, Japan, May 21, 2002 The Region Immediately Surrounding the Plasma Divertor / First
More informationAN OVERVIEW OF DUAL COOLANT Pb-17Li BREEDER FIRST WALL AND BLANKET CONCEPT DEVELOPMENT FOR THE US ITER-TBM DESIGN
GA A24985 AN OVERVIEW OF DUAL COOLANT Pb-17Li BREEDER FIRST WALL AND BLANKET CONCEPT by C.P.C. WONG, S. MALANG, M. SAWAN, M. DAGHER, S. SMOLENTSEV, B. MERRILL, M. YOUSSEF, S. REYES, D.K. SZE, N.B. MORLEY,
More informationThermal-Hydraulic Study of ARIES-CS Ceramic Breeder Blanket Coupled with a Brayton Cycle
Thermal-Hydraulic Study of ARIES-CS Ceramic Breeder Blanket Coupled with a Brayton Cycle Presented by A. R. Raffray With contributions from L. El-Guebaly, S. Malang, X. Wang and the ARIES team ARIES Meeting
More informationDevelopment of Advanced Operation Scenarios in Weak Magnetic-Shear Regime on JT-60U
EX/1-4Rc Development of Advanced Operation Scenarios in Weak Magnetic-Shear Regime on JT-60U T. Suzuki 1), N. Oyama 1), A. Isayama 1), Y. Sakamoto 1), T. Fujita 1), S. Ide 1), Y. Kamada 1), O. Naito 1),
More informationExperimental Study of Plasma Confinement on EAST
IAEA-F1-CN-180- EXC/P4-06 Experimental Study of Plasma Confinement on EAST Xiang Gao, Yao Yang, Zixi Liu, Long Zeng, Shoubiao Zhang, Nan Shi, Yinxian Jie, Wei Liao, Yumin Wang, Jingliang Bu, Baonian Wan,
More informationEAST(HT-7U ) Physics and Experimental Plan
EAST(HT-7U ) Physics and Experimental Plan EAST team, presented by Jiangang Li Institute of Plasma Physics, CAS 1 st EAST IAC meeting Hefei, Oct.10-11 The mission of EAST To explore the methods to achieve
More informationAn overview of dual coolant Pb 17Li breeder first wall and blanket concept development for the US ITER-TBM design
Fusion Engineering and Design 81 (2006) 461 467 An overview of dual coolant Pb 17Li breeder first wall and blanket concept development for the US ITER-TBM design C.P.C. Wong a,, S. Malang b,m.sawan c,
More informationHe-cooled Divertor Development in the EU: The Helium Jet cooled Divertor HEMJ
He-cooled Divertor Development in the EU: The Helium Jet cooled Divertor HEMJ Presented by Thomas Ihli * Contributors: Divertor Group at, Germany and Efremov Institute, Russia ARIES Meeting General Atomics,
More informationARIES-ACT-DCLL NWL Distribution and Revised Radial Build
ARIES-ACT-DCLL NWL Distribution and Revised Radial Build L. El-Guebaly and A. Jaber Fusion Technology Institute University of Wisconsin-Madison http://fti.neep.wisc.edu/uwneutronicscenterofexcellence Contributors:
More informationFUSION TECHNOLOGY INSTITUTE
FUSION TECHNOLOGY INSTITUTE An Improved First Stability Advanced Fuel Tokamak, Apollo-L3 W I S C O N S I N G.A. Emmert, G.L. Kulcinski, J.P. Blanchard, L.A. El-Guebaly, H.Y. Khater, C.W. Maynard, E.A.
More informationTaming Plasma-Materials Interface for Steady-State Fusion
Taming Plasma-Materials Interface for Steady-State Fusion by H.Y. Guo, with H. Wang, J.G. Watkins, A.L. Moser, J. Boedo, L. Casali, B. Covele, B. Grierson, M. Groth, D.N. Hill, A.W. Hyatt, L.L. Lao, A.W.
More informationW. M. Stacey, J. Mandrekas, E. A. Hoffman, G. P. Kessler, C. M. Kirby, A.N. Mauer, J. J. Noble, D. M. Stopp and D. S. Ulevich
A FUSION TRANSMUTATION OF WASTE REACTOR W. M. Stacey, J. Mandrekas, E. A. Hoffman, G. P. Kessler, C. M. Kirby, A.N. Mauer, J. J. Noble, D. M. Stopp and D. S. Ulevich Nuclear & Radiological Engineering
More informationFeedback Stabilization of Vertical Instabilities. Gratefully Acknowledge Chuck Kessel for assistance with ARIES comparisons
Feedback Stabilization of Vertical Instabilities Mike Kotschenreuther University of Texas Gratefully Acknowledge Chuck Kessel for assistance with ARIES comparisons Introduction The conducting shell needed
More informationCritical Physics Issues for Tokamak Power Plants
Critical Physics Issues for Tokamak Power Plants D J Campbell 1, F De Marco 2, G Giruzzi 3, G T Hoang 3, L D Horton 4, G Janeschitz 5, J Johner 3, K Lackner 4, D C McDonald 6, D Maisonnier 1, G Pereverzev
More informationRecent Results and Plans for the Advanced Tokamak Program
Recent Results and Plans for the Advanced Tokamak Program Program Advisory Committee Review February 6-7, 2002 MIT PSFC Presented by A. Hubbard Outline Overview Results from 2001and plans for 2002 - Internal
More informationWendelstein 7-X A technology step towards DEMO
A technology step towards DEMO Hans-Stephan Bosch Max-Planck Institute for Plasma Physics Greifswald, Germany 18th Internatinal Toki Conference, December 9 13, 2008, Toki-City, Japan 1-YKA06-Y0001.0 The
More informationR&D required to place a test module on FNF (how does it compare to ITER TBM?) R&D required for base blanket
Testing Strategy, Implications for R&D and Design What are the preferred blankets options for testing on FNF and what are the implications for R&D? Comparison of strategies for testing space allocation
More informationFusion power core engineering for the ARIES-ST power plant
Fusion Engineering and Design 65 (2003) 215/261 www.elsevier.com/locate/fusengdes Fusion power core engineering for the ARIES-ST power plant M.S. Tillack a, *, X.R. Wang a, J. Pulsifer a, S. Malang b,
More informationAN ADVANCED COMPUTATIONAL APPROACH TO SYSTEM MODELING OF TOKAMAK POWER PLANTS
AN ADVANCED COMPUTATIONAL APPROACH TO SYSTEM MODELING OF TOKAMAK POWER PLANTS Zoran Dragojlovic 1, Charles Kessel 2, Rene Raffray 1, Farrokh Najmabadi 1, Lester Waganer 3, Laila El-Guebaly 4, Leslie Bromberg
More informationProspects for a next-step ST
Prospects for a next-step ST Jon Menard (PPPL) with contributions from: Ray Fonck (UW Madison) Stan Kaye (PPPL) Dick Majeski (PPPL) Masa Ono (PPPL) Steve Sabbagh (Columbia U.) October 24, 2009 Comprehensive
More informationIntegrated Scenarios: Advanced Regimes
Integrated Scenarios: Advanced Regimes Program Advisory Committee Meeting February 6-8, 2008 MIT PSFC Presented by A. Hubbard, for the Advanced Scenarios thrust group Outline Scope and niche of Advanced
More informationMagnetic Confinement Fusion: Progress and Recent Developments
Magnetic Confinement Fusion: Progress and Recent Developments Howard Wilson, Dept Physics, University of York, Heslington, York YO10 5DD With thanks to A Field, K Gibson and A Kirk howard.wilson@york.ac.uk
More informationGA A FUSION TECHNOLOGY FACILITY KEY ATTRIBUTES AND INTERFACES TO TECHNOLOGY AND MATERIALS by C.P.C. WONG
GA A27273 FUSION TECHNOLOGY FACILITY KEY ATTRIBUTES AND INTERFACES TO TECHNOLOGY AND MATERIALS by C.P.C. WONG MARCH 2012 DISCLAIMER This report was prepared as an account of work sponsored by an agency
More informationRole of Fusion Energy in the 21 st Century
Role of Fusion Energy in the 21 st Century Farrokh Najmabadi Prof. of Electrical Engineering Director of Center for Energy Research UC San Diego Lehigh University Physics Department Colloquium April 26,
More informationCompeted ITER Task Agreements (ITAs, for information and expression of interest)
Competed ITER Task Agreements (ITAs, for information and expression of interest) ONGOING CALLS Ref F4E-CITA-002: "Evaluation of edge MHD stability and uncontrolled ELM energy losses for ITER H-mode plasmas
More informationLithium and Liquid Metal Studies at PPPL
Lithium and Liquid Metal Studies at PPPL LTX R. Maingi, M. Jaworski, R. Kaita, R. Majeski, J. Menard, M. Ono PPPL High Power Devices NSTX-U Surface Analysis EAST Test stands IAEA TM on Divertor Concepts
More informationR & D of the Fabrication Technology for ITER Magnet Supports
ITR/P1-45 R & D of the Fabrication Technology for ITER Magnet Supports 1 P. Y. Lee*, 1 C.J.Pan, 1. L. Hou, 1 S.L.Han, 1 Z.C.Sun, 1 X. R. Duan, 1 Y. Liu, 2 F. Savary, 2 Y. K. Fu and 2 R. Gallix, 2 N. Mitchell
More informationDevelopment of Low Activation Structural Materials
Materials Challenge for Clean Nuclear Fusion Energy Development of Low Activation Structural Materials T. Muroga National Institute for Fusion Science, Oroshi, Toki, Gifu 509-5292, Japan Symposium on Materials
More informationITER R&D Needs, Challenges, and the Way Forward
ITER R&D Needs, Challenges, and the Way Forward Bernard Bigot Director General ITER Organization, Cadarache, France Fusion Power Co-ordinating Committee Mtg, IO Headquarters, 24 Jan 2018 1 Context of ITER
More informationFusion Transmutation Reactor -Feasible early use of fusion fast neutron
Fusion Transmutation Reactor -Feasible early use of fusion fast neutron Jung-Hoon HAN, CARFRE, SNU October 2, 2009 Fusion-Fission Hybrid workshop, Gaithersburg, MD 1 contents 1. prologue, atmosphere and
More informationA feasible DEMO blanket concept based on water cooled solid breeder
1 FTP/P7-33 A feasible DEMO blanket concept based on water cooled solid breeder Y. Someya 1, K. Tobita 1, H. Utoh 1, K. Hoshino 1, N. Asakura 1, M. Nakamura 1, Hisashi Tanigawa 2, M. Enoeda 2, Hiroyasu
More informationPreliminary Design of ITER Component Cooling Water System and Heat Rejection System
Preliminary Design of ITER Component Cooling Water System and Heat Rejection System A.G.A. Kumar 1, D.K. Gupta 1, N. Patel 1, G. Gohil 1, H. Patel 1, J. Dangi 1, L. Sharma 1, M. Jadhav 1, L. Teodoros 2,
More informationMaintenance Concept for Modular Blankets in Compact Stellarator Power Plants
Maintenance Concept for Modular Blankets in Compact Stellarator Power Plants Siegfried Malang With contributions of Laila A. EL-Guebaly Xueren Wang ARIES Meeting UCSD, San Diego, January 8-10, 2003 Overview
More informationReNeW PMI Theme PFC Panel Report
ReNeW PMI Theme PFC Panel Report Organization First question at the beginning: What are we doing? Technologists to physicists: What heat fluxes will the DEMO have? Physicists to technologists: What are
More informationHigh heat flux components for a DEMO fusion reactor: material and technology development
High heat flux components for a DEMO fusion reactor: material and technology development Matti Coleman Power Plant Physics and Technology Department EUROfusion G. Federici, J-H. You, T. Barrett, C. Bachmann,
More information