Overview of LLCB TBM for ITER Paritosh Chaudhuri Institute for Plasma Research Gandhinagar, INDIA CBBI-16, 8- 10 Sept. 2011, Portland, USA.

Slides:



Advertisements
Similar presentations
Hongjie Zhang Purge gas flow impact on tritium permeation Integrated simulation on tritium permeation in the solid breeder unit FNST, August 18-20, 2009.
Advertisements

Ss Hefei, China July 19, 2011 Nuclear, Plasma, and Radiological Engineering Center for Plasma-Material Interactions Contact: Flowing.
First Wall Heat Loads Mike Ulrickson November 15, 2014.
Test Blanket Module: Steels & Fabrication Technologies
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.
Thermo Fluid Design Analysis of TBM cooling schemes M. Narula with A. Ying, R. Hunt, S. Park ITER-TBM Meeting UCLA Feb 14-15, 2007.
What is Dual Coolant Blanket? Siegfried Malang 2 nd EU-US DCLL Workshop2 nd EU-US DCLL Workshop University of California,University of California, Los.
Presented by: S. Suzuki, Blanket Engineering Lab., Japan Atomic Energy Research Institute, JAERI Contents 1.Outline of blanket development in JAERI 2.Design.
First Wall Thermal Hydraulics Analysis El-Sayed Mogahed Fusion Technology Institute The University of Wisconsin With input from S. Malang, M. Sawan, I.
Thermo-fluid Analysis of Helium cooling solutions for the HCCB TBM Presented By: Manmeet Narula Alice Ying, Manmeet Narula, Ryan Hunt and M. Abdou ITER.
ARIES Meeting General Atomics, February 25 th, 2005 Brad Merrill, Richard Moore Fusion Safety Program Pressurization Accidents in ARIES-CS.
Summary of Current Test Plan for US DCLL TBM in ITER Neil Morley and the US TBM Participants INL, August
The shield block is a modular system made up of austenitic steel SS316 LN-IG whose main function is to provide thermal and nuclear shielding of outer components.
Status of safety analysis for HCPB TBM Susana Reyes TBM Project meeting, UCLA, Los Angeles, CA May 10-11, 2006 Work performed under the auspices of the.
1 Wilfrid Farabolini 23 feb. 04Direction de l’Energie Nucléaire Tritium Control A Major Issue for a Liquid Metal Blanket.
November 8-9, Blanket Design for Large Chamber A. René Raffray UCSD With contributions from M. Sawan (UW), I. Sviatoslavsky (UW) and X. Wang (UCSD)
Ceramic Breeder Blanket Conceptual Design for ARIES-CS Contributors: S. Malang, A.R. Raffray, and L. El-Guebaly ARIES Meeting University of Wisconsin,
Japan considerations on design and qualification of PFC's for near term machines (ITER) Satoshi Suzuki 1, Satoshi Konishi 2 1 Japan Atomic Energy Agency.
The main function of the divertor is minimizing the helium and impurity content in the plasma as well as exhausting part of the plasma thermal power. The.
THERMOFLUID MHD for ITER TBM. CURRENT STATUS By UCLA Thermofluid MHD GROUP Presented by Sergey Smolentsev US ITER TBM Meeting UCLA May 10-11, 2006.
June19-21, 2000Finalizing the ARIES-AT Blanket and Divertor Designs, ARIES Project Meeting/ARR ARIES-AT Blanket and Divertor Design (The Final Stretch)
A design for the DCLL inboard blanket S. Smolentsev, M. Abdou, M. Dagher - UCLA S. Malang – Consultant, Germany 2d EU-US DCLL Workshop University of California,
US ITER TBM Meeting Idaho Fall, Idaho, Aug M Dagher P Fogarty 1.TBM/ITER General Arrangement 2.Equatorial Test port Configuration 3.Test Port.
Status of the ARIES-CS Power Core Configuration and Maintenance Presented by X.R. Wang Contributors: S. Malang, A.R. Raffray ARIES Meeting PPPL, NJ Sept.
Thermofluid MHD issues for liquid breeder blankets and first walls Neil B. Morley and Sergey Smolentsev MAE Dept., UCLA APEX/TBM Meeting November 3, 2003.
March 20-21, 2000ARIES-AT Blanket and Divertor Design, ARIES Project Meeting/ARR Status ARIES-AT Blanket and Divertor Design The ARIES Team Presented.
The effect of the orientations of pebble bed in Indian HCSB Module Paritosh Chaudhuri Institute for Plasma Research Gandhinagar, INDIA CBBI-16, Sept.
ITER Test Blanket Module and the Need for Coordination Outline What ITER means for the World Technology / Chamber / Blanket Community ITER Plans for TBM.
By S. Saeidi Contribution from: S. Smolentsev, S. Malang University of Los Angeles August, 2009.
1 Recent Progress in Helium-Cooled Ceramic Breeder (HCCB) Blanket Module R&D and Design Analysis Ying, Alice With contributions from M. Narula, H. Zhang,
ASIPP EAST Overview Of The EAST In Vessel Components Upgraded Presented by Damao Yao.
Iván Fernández CIEMAT 2 nd EU-US DCLL Workshop, University of California, Los Angeles, Nov th, 2014.
Finite Elements Modelling of Tritium Permeation
High-Power Density Target Design and Analyses for Accelerator Production of Isotopes W. David Pointer Argonne National Laboratory Nuclear Engineering Division.
Neutronics Parameters for Preferred Chamber Configuration with Magnetic Intervention Mohamed Sawan Ed Marriott, Carol Aplin UW Fusion Technology Inst.
0 Laser Flash Method for Effective Thermal Diffusivity Measurement of Pebble Beds CBBI-16 Portland, OR, USA Sept. 9, 2011 Mu-Young Ahn 1, Duck Young Ku.
October 27-28, 2004 HAPL meeting, PPPL 1 Thermal-Hydraulic Analysis of Ceramic Breeder Blanket and Plan for Future Effort A. René Raffray UCSD With contributions.
HCCB TBM Mechanical Design R. Hunt, A. Ying, M. Abdou Fusion Science & Technology Center University of California Los Angeles May 11, 2006 Presented by.
1 Solid Breeder Blanket Design Concepts for HAPL Igor. N. Sviatoslavsky Fusion Technology Institute, University of Wisconsin, Madison, WI With contributions.
1 Preliminary Design of China ITER TBM with Helium-Cooled and Solid Breeder Concept Preliminary Design of China ITER TBM with Helium-Cooled and Solid Breeder.
Neutronics Analysis for K-DEMO Blanket Module with Helium coolant June 26, 2013 Presented by Kihak IM Prepared by Y.S. Lee Fusion Engineering Center DEMO.
Design study of advanced blanket for DEMO reactor US/JP Workshop on Fusion Power Plants and Related Advanced Technologies 23 th -24 th Feb at UCSD,
ITER test plan for the solid breeder TBM Presented by P. Calderoni March 3, 2004 UCLA.
Pacific Northwest National Laboratory U.S. Department of Energy TBM Structure, Materials and Fabrication Collaboration Issues R.J. Kurtz 1, and A.F. Rowcliffe.
David Rapisarda CIEMAT 2 nd EU-US DCLL Workshop University of California, Los Angeles, Nov th, 2014.
1 Parametric Thermal-Hydraulic Analysis of TBM Primary Helium Loop Greg Sviatoslavsky Fusion Technology Institute, University of Wisconsin, Madison, WI.
1 Neutronics Assessment of Self-Cooled Li Blanket Concept Mohamed Sawan Fusion Technology Institute University of Wisconsin, Madison, WI With contributions.
Background information of Party(EU)’s R&D on TBM and breeding blankets Compiled and Presented by Alice Ying TBM Costing Kickoff Meeting INL August 10-12,
Development of tritium breeder monitoring for Lead-Lithium cooled ceramic breeder (LLCB) module of ITER presented V.K. Kapyshev CBBI-16 Portland, Oregon,
Status of MHD/Heat Transfer Analysis for DCLL US-ITER TBM Meeting February 14-15, 2007 Rice Room, Boelter Hall 6764, UCLA Thermofluid / MHD group Presented.
DCLL ½ port Test Blanket Module thermal-hydraulic analysis Presented by P. Calderoni March 3, 2004 UCLA.
Association Euratom-Cea IEA Tritium and Safety Issues in LL Breeders, June 2007, Idaho Falls J-F. Salavy 1 IEA Implementing Agreement on Nuclear.
Review of Thermofluid / MHD activities for DCLL Sergey Smolentsev & US TBM Thermofluid/MHD Group 2006 US-Japan Workshop on FUSION HIGH POWER DENSITY COMPONENTS.
Required Dimensions of HAPL Core System with Magnetic Intervention Mohamed Sawan Carol Aplin UW Fusion Technology Inst. Rene Raffray UCSD HAPL Project.
March 3-4, 2005 HAPL meeting, NRL 1 Assessment of Blanket Options for Magnetic Diversion Concept A. René Raffray UCSD With contributions from M. Sawan.
ITER TBM MEETING March UCLA DCLL Module Design Prepared by M. Dagher UCLA P. Fogarty ORNL.
1 A Self-Cooled Lithium Blanket Concept for HAPL I. N. Sviatoslavsky Fusion Technology Institute, University of Wisconsin, Madison, WI With contributions.
704 MHz cavity design based on 704MHZ_v7.stp C. Pai
Thermal-hydraulic analysis of unit cell for solid breeder TBM
ISIS TS1 Project: Target Design and Analysis
DCLL TBM Reference Design
Integrated Design: APEX-Solid Wall FW-Blanket
University of California, San Diego
DCLL TBM Design Status, Current and future activities
Status of ARIES-CS Power Core Engineering
VLT Meeting, Washington DC, August 25, 2005
Updated DCLL TBM Neutronics Analysis
DCLL TBM Design Status FNST Meeting, August 12-14, 2008, UCLA
TBM Design Meeting UCLA
Nuclear Analyses for two “Look-alike” HCPB Blanket Sub-modules for Testing in ITER Mahmoud Z Youssef UCLA Presented at ITER-TBM4 Meeting, UCLA, March 2-4,
Presentation transcript:

Overview of LLCB TBM for ITER Paritosh Chaudhuri Institute for Plasma Research Gandhinagar, INDIA CBBI-16, Sept. 2011, Portland, USA

Intruduction In India, development of Lead-Lithium Ceramic Breeder (LLCB) blanket is being performed as the primary candidate of Test Blanket Module (TBM) towards DEMO reactor. The LLCB TBM will be tested from the first phase of ITER operation (H-H phase) in one-half of a ITER port no-2. The Indian TBM R&D program is focused on the development and characterization of materials: structural (IN-RAFMS), breeding materials (Pb–Li, Li2TiO3) and development of technologies for Lead-Lithium Systems, Helium Cooling Systems, Tritium Extraction Systems, TBM manufacturing and coatings.

Indian TBM Program Lead-Lithium cooled Ceramic Breeder (LLCB)  Tritium Breeder: Lithium Titanate;  Coolant: Pb-Li eutectic alloy (multiplier and breeder)  FW coolant: Helium Gas;  Structural Material : Reduced Activation FMS  Helium purge gas for T extraction from CB India is developing Lead-Lithium cooled Ceramic Breeder (LLCB) TBM for testing in ITER Port No-2 with position of TBM Leader.

LLCB TBM Internal Arrangements Dimensions~1.66 m (P) x.484 m (T)x.496 m (R) Structural Material FMS (IN-RAFMS) 28 mm thick BreederPbLi, Li 2 TiO 3 Total Power Deposition ~ MW NWL~ 0.78 MW/m 2 CoolantPbLi and Helium

5 TBM SYSTEM - Schematic

LLCB TBM Neutronic Analysis

 Fusion Power : 500 MW, Neutron Wall load : 0.78 MW/m^2  Pulse duration is 400 sec with pulse repetition time 1800 sec  Major and Minor radius of plasma: 6.3/ 2.1 m  Structural Material: IN-RAFMS (Eurofer as a reference material)  Shield Material:- SS-316 (65%) and water (35%)  Vacuum Vessel: Borated SS-316 cooled with water  Blanket materials: -Pb-Li Eutectic (90 % Li-6) (Breeder, Multiplier and coolant) -Li2TiO3 (60 % Li-6, Packing fraction 60%)(Breeder Material) -He (Coolant for the First wall) Input parameters for LLCB TBM neutronic calculations

Poloidal Radial view of ITER neutronic model (with LLCB TBM) LLCB TBM in ITER sector model  A 15 degree sector of the ITER machine has been constructed. The shield and vacuum vessel has been made using the intersection of concentric tori and concentric cylinders.  TBM dimensions are taken as 1.66 x0.484x0.518 m^3 (pol x tor x rad ). TBM geometry has been modeled using the parallel planes.  A 20 cm thick water jacket has been placed around the TBM. 100 cm thick shield plug has been put after the TBM.

Total Power deposited in LLCB TBM is 0.62 MW Power density profile & Total Power Deposition in LLCB TBM

Preliminary LLCB TBM shield block design  The TBM in ITER will attenuate the flux ~1 order magnitude  To reduce dose rates this flux should be further reduced (~10 6 n/cm 2 /s) for safety and maintenance purpose  The shield block consists of 60 % SS 316 LN and 40 % water (DM) LLCB TBM + Shield Block in Frame assembly LLCB TBM Shield block Shield Flange with lip seal Frame Assembly Schematic Reference Solid TBM Sub-modules

LLCB TBM Thermal Hydraulic

Thermal Hydraulics of ITER TBM Main Objectives: To optimize a suitable design of FW cooling w.r.t. Neutron wall load and heat flux. To estimate the temperature and thermal stresses of all materials used in the blanket module and ensure these values are within design limits. To keep the temperatures of ceramic breeder zones within the temperature window for effective Tritium release To optimize the flow parameters (velocity, pressure)

X20 20X11 R2.5 Circuit-1 Circuit-2 28 LLCB First Wall Structure Total number of channels - 64 Number of circuits (counter flow) – 2 Number of passes per circuit – 4 Number of channels per passes – 8 Pitch: 25.5 mm (typical to all channels) Rib thickness between channels = 5.5 mm

Input Parameters - Total Heat Load: MW - He inlet Temp: 300  C - He inlet pressure: 8 MPa - He velocity : 45 m/s - PbLi inlet Temp: 325  C - PbLi inlet pressure: 1.2 MPa - PbLi velocity: 0.1, m/s Assumptions - Flow is Steady and incompressible - Flow is turbulent Output: - Radial temp. profile in all zones; - PbLi velocity, Pressure Drop profile, outlet temp. Power Deposition on LLCB TBM Analytical Model

Radial Temperature Plot for LLCB TBM Temperature Distribution of LLCB TBM (for different V= 0.1, 0.2, 0.3, 0.5 m/s)

Radial temperature profile in different CB zones Radial temperature profile in different CB zones for PbLi velocity of 0.1 m/s

Peak Temperature at different zones in LLCB TBM Peak Temperature at different zones in TBM (for PbLi velocity = 0.1 m/s) Zone Name Temperature (  C) Analytical Analysis) CFD Analysis Be Surface Be-FMS Interface FMS-Insulator Interface Insulator-PbLi Interface Insulator-FMS Interface FMS-CB Interface CB zone489 (5 th Breeder) 491 (5 th Breeder)

Alternative Concept

Indian RAFMS Development: - Composition Achieved - Melts are under characterization (Microstructure & Mechanical properties) Lead-Lithium Technologies development: - Lead-Lithium production - Pb-Li Corrosion experiments - Full scale Pb-Li loop development R&D Activities under progress (1/2) Ceramic Pebbles Fabrication: -Lab scale pebbles (Lithium Titanate) successfully fabricated - Pebble bed characteristics are under investigation -Large scale production plans are under progress

Helium Loop development -¼ the size loop development plan - Small scale TBM testing Tritium Extraction Systems development -H/D extraction from Helium purge gas -H/D extraction from Pb-Li - Permeation Analysis R&D Activities under progress (2/2)

RAFM steel is a structural material under development for fusion reactor applications. Several fabrication processes for the production of TBM sub- components and assembly need to be investigated in the developmental program. For the fabrication of sub-components (first wall, stiffening plates, cooling plates and caps) using the Indian RAFMS different options are being considered to investigate its fabricability: HIP process, EB welding, Laser and Narrow Gap TIG welding. Work under Progress

EB welding on 6 mm thick austenitic stainless steel plates. Welds of this structure cleared both ultrasonic examination and radiography. Virtually no distortion was observed in this structure. It is planned to make similar structures using RAFM steel plates to gain experience in fabrication. Initial trials for First wall small scale mock-up by HIP process were carried out using stainless steel discs with pre-machined slots for making internal channels. Fig.6 shows the hipped part.

Cutting of straight square channel by wire EDM followed by hot bending. advantage of this approach is that there will be no mating surface and therefore, no lack of bonding. However, challenges remain in cutting such a long square channel by wire EDM. Alternative approach for FW fabrication

Summary  The major areas of R&D activities are: - Development of technologies of circulators, heat exchangers and diagnostics for Lead-Lithium Systems, HCS, LL, TES - Lead-Lithium Loop developments: - MHD studies: - Li2TiO3 Ceramic Pebbles development  The RpRs report on safety analysis have been submitted, further detail work is in progress.  Institute for Plasma Research (IPR), collaboration with Bhaba Atomic Research Centre (BARC), Indira Gandhi Centre for Atomic Research (IGCAR) and other research institutions and universities within India involve the R&D activities focusing on ITER-TBM systems development.

26 Thank you

MHD pressure drop in LLCB TBM Channel Parameters Channel Height (2a) = 50 mm Channel Width (2b) = 484 mm Insulation layer (A2B2) thickness = 0.2 mm Outer wall (A1B1) thickness = 5 mm Channel length = 10 m Electrical conductivity : Pb-Li ~ 0.7e6, Fe wall ~1.4e6, Alumina ~ 1e-9 Hartmann number ~ Mean velocity is 0.1 m/s MHD Pressure Drop (1)Without coating : kPa (2)With Alumina coating : kPa With Flow dividers Considerably small MHD pressure drop inside TBM The MHD pressure drop is calculated with an analytical expression for fully developed laminar flow in a rectangular duct.

2D MHD Code: Velocity & induced field With aluminaWithout alumina Further work: MHD effects in Manifolds & in fringe field Heat Transfer with MHD effects Crack analysis More details in Poster NO: PO1-10, K.S. Goswami Surface plot velocity profile contour plot of induced magnetic field Velocity profile

Analysis with 2-D MHD code Fully developed MHD flow in a rectangular channel Surrounded by various layers with different electrical properties Similar to S.Smolentsev et al. (FED, 2005) Velocity and induced magnetic field are calculated self consistently No-slip condition at the liquid-solid interface No induced magnetic field at the outer boundary of the computational domain

Validation of 2-D MHD code Ha & grid sizeJ.R Hunt (1965)MHD code 100 (81x81,12) x x (81x81,12)0.4195x x (81x81,12) x x (161x161,24) x x (81x81,12) 0.310x x Comparison with analytical results

Velocity profile for higher Ha