ITER Test Blanket Module (TBM) Program Briefing to DOE/OFES and US ITER Project Management Mohamed Abdou Video Conference, September 9, 2004
I. What is ITER TBM? II. Why is ITER TBM Critical? III. History of ITER TBM, current TBWG Activities, and other Parties’ TBMs IV. US Strategy for ITER TBM and description of effort and selected concepts Programmatic Issues requiring DOE and US ITER management attention V. Outline Why must the US participate in ITER TBM? (includes tritium supply issue, impact on physics, focus for materials and technology) - Part of US ITER Organization - Focus for Plasma Chamber, Materials, and other programs - Resources - Formal International Agreements
What is the ITER TBM Program? • Breeding Blankets will be tested in ITER, starting on Day One, by inserting Test Blanket Modules (TBMs) in specially designed ports. • Each TBM will have its own dedicated systems for tritium recovery and processing, heat extraction, etc. Each TBM will also need new diagnostics for the nuclear-electromagnetic environment. • Each ITER Party is allocated limited space for testing two TBMs. (Number of Ports reduced to 3. Number of Parties increased to 6). • ITER’s construction plan includes specifications for TBMs because of impacts on space, vacuum vessel, remote maintenance, ancillary equipment, safety, availability, etc. • The ITER Test Program is managed by the ITER Test Blanket Working Group (TBWG) with participants from the ITER International Team and representatives of the Parties.
First DT plasma phase H-Plasma Phase D Phase Accumulated fluence = 0.09 MWa/m2 BlanketTest ITER Operational Plan Calls for Testing Breeding Blankets from Day 1 of Operation
TBM Roll Back from ITER 1st Plasma Shows R&D must beaccelerated now for TBM Selection in 2005 EU schedule for Helium-Cooled Pebble Bed TBM (1 of 4 TBMs Planned) ITER First Plasma A final decision on blanket test module selection by 2005 in order to initiate design, fabrication and out-of-pile testing. (Reference: S. Malang, L.V. Boccaccini,ANNEX 2, "EFDA Technology Work programme 2002 Field: Tritium Breeding and Materials 2002 activities- Task Area: Breeding Blanket (HCPB), Sep. 2000)
ITER’s Principal Objectives Have Always Included Testing Tritium Breeding Blankets • “The ITER should serve as a test facility for neutronics, blanket modules, tritium production and advanced plasma technologies. The important objectives will be the extraction of high-grade heat from reactor relevant blanket modules appropriate for generation of electricity.”—The ITER Quadripartite Initiative Committee (QIC), IEA Vienna 18–19 October 1987 • “ITER should test design concepts of tritium breeding blankets relevant to a reactor. The tests foreseen in modules include the demonstration of a breeding capability that would lead to tritium self sufficiency in a reactor, the extraction of high-grade heat and electricity generation.”—SWG1, reaffirmed by ITER Council, IC-7 Records (14–15 December 1994), and stated again in forming the Test Blanket Working Group (TBWG)
The US has been a leader in the science and engineering of blanket testing in “ITER” for 20 years • The US has made major contributions to the ITER Test Program since early CDA (and, even earlier, INTOR) • The US is widely acknowledged as the primary “intellectual power” in fusion testing because: – Many US studies in the 80’s and 90’s such as FINESSE, VNS, etc, defined the FNT issues and testing requirements – The Fusion Engineering Science for “effective testing” in fusion facilities has been developed mostly by the US (search scholarly journals on the subject, it is dominated by US authors)
II. Why is ITER TBM Critical? Why is it necessary for the US to participate in ITER TBM?
Tritium Self-SufficiencyTritium self-sufficiency condition: Λa > Λr Λr = Required tritium breeding ratio Λr is 1 + G, where G is the margin required to account for tritium losses, radioactive decay, tritium inventory in plant components, and supplying inventory for start-up of other plants. Λr is dependent on many system physics and technology parameters: • plasma edge recycling, tritium fractional burn-up in the plasma • tritium inventories (release/retention) in components • efficiency/capacity/reliability of the tritium processing system, etc. Λa = Achievable tritium breeding ratio Λa is a function of technology, material and physics. • FW thickness, amount of structure in the blanket, blanket concept (ITER detailed engineering is showing FW may have to be much thicker than we want for T self sufficiency) • Presence of stabilizing/conducting shell materials/coils for plasma control and attaining advanced plasma physics modes • Plasma heating/fueling/exhaust, PFC coating/materials/geometry • Plasma configuration (tokamak, stellerator, etc.)
ITER Blanket Testing is Essential to: • Achieve a key element of the “ITER Mission” • Establish the conditions governing the scientific feasibility of the D-T cycle, i.e., determine the “phase-space” of plasma, nuclear, material, and technological conditions in which tritium self-sufficiency can be attained - The D-T cycle is the basis of the current world plasma physics and technology program. There is only a “window” of physics and technology parameters in which the D-T cycle is feasible. We need to determine this “window.” (If the D-T cycle is not feasible the plasma physics and technology research would be very different.) • Examples of questions to be answered: • Can we allow low plasma-edge recycling? • Is high plasma-edge recycling necessary for T self sufficiency? • Are advanced physics modes acceptable? • Is the “temperature window” for tritium release from solid breeders sufficient for adequate TBR? • Is there a blanket/material system that can exist in this phase-space?
ITER Blanket Testing is Essential to (cont’d): • Achieve the most critical milestone in blanket and material research: testing in the integrated fusion environment (ITER construction and operation is for the next 30 years. Without such fusion testing, material and blanket research loses “focus”, relevance: Why are we doing any research in these areas then?) • Develop the technology necessary to install breeding capabilities to supply ITER with tritium for its extended phase of operation • Resolve the critical “tritium supply” issue for fusion development - and at a fraction of the cost to buy tritium for large D-T burning plasma
Other Reasons why the US should Participate in ITER TBM • Test critical technologies for any further US development of fusion (CTF, DEMO, alternate confinement facilities, power plants) • Access R&D information from much larger ($10-20M per year) blanket programs (EU and Japan) and other international partners • To build US knowledge, experience, and competence in fusion nuclear and tritium technologies needed to develop practical and safe DT fusion devices (Building competence takes decades) • Keep the door open for the US to be a world leader in fusion again in the future For a few million dollars’ expenditure on test blanket modules, we will acquire vital data and develop critical technologies – an additional excellent return on the billions of dollars invested in ITER.
ITER Provides the First Integrated Experimental Conditions for Fusion Technology Testing • Simulation of all Environmental Conditions • Neutrons Plasma Particles • Electromagnetics Tritium • Vacuum Synergistic Effects • Correct Neutron Spectrum (heating profile) • Large Volume of Test Vehicle • Large Total Volume, Surface Area of Test Matrix
Tritium Consumption and Production • Tritium Consumption in Fusion is HUGE! • 55.8 kg per 1000MW fusion power per year • Production & Cost • CANDU Reactors:27 kg from over 40 years, $30M/kg (current) • Fission reactors:2-3 kg per year. It takes tens of fission reactors to supply one fusion reactor. • $84M-$130M per kg, per DOE Inspector General* • Conclusions • The cost of blanket development and ITER TBM is a fraction of the cost to “purchase” tritium for a burning plasma facility such as ITER. • “Availability” of external tritium supply for continued fusion development is an issue. • Large power DT facilities must breed their own tritium. (This is why ITER’s extended phase was planned to install a tritium breeding blanket.) *DOE Inspector General’s Audit Report, “Modernization of Tritium Requirements Systems”, Report DOE/IG-0632, December 2003, available at www.ig.doe.gov/pdf/ig-0632.pdf
World Tritium Supply Would be Exhausted by 2025 if ITER Were to Run at 1000MW at 10% Availability (OR at 500 MW at 20% availability) 30 25 CANDU Supply 20 w/o Fusion Projected Ontario (OPG) Tritium Inventory (kg) 15 1000 MW Fusion, 10% Avail, TBR 0.0 10 ITER-FEAT (2004 start) 5 0 1995 2000 2010 2015 2020 2025 2030 2035 2040 2045 2005 Year
III. History of TBM, Current TBWG Activities, and other parties’ TBMs
History of ITER Test Program: From prior to CDA to EDA to Now (and why things are the way they are!!) • Negotiations Prior to ITER CDA -- Providing Capabilities for testing was a central element in negotiating “Objectives” and “Design Envelope” -- Plasma physics parameters (steady state, pulse length, etc.), wall loads, etc. were most influenced by the need to do testing -- Formal agreements have always had “Blanket Testing” as a key ITER objective • During CDA (and early EDA) -- A powerful International “Testing Group” was formed by ITER Management. -- M. Abdou was asked to chair -- Played a key role in defining ITER parameters and features
Test Blanket Working Group (TBWG) – TBWG was proposed by SWG-1 and established by the ITER Council (IC-7 RECORDS, 14-15 December 1994) – A formal and detailed charter of TBWG was developed ■ Chair from the Parties and Vice Chair from Central Team ■ TBWG members: 3 from each Party and 6 from Central Team – TBWG Main Objectives: Develop a Coordinated Blanket Test Program and address the Interface between machine and the blanket modules – During EDA, TBWG addressed both: ■ the breeding blanket for the ITER extended performance phase ■ DEMO-relevant breeding blanket concepts of the Parties, modules of which were planned to be installed in the ITER machine for testing: developed GDRD and specified interfaces with ITER – During EDA, the US made major contributions to testing strategy, engineering scaling, test port, frame design, and machine interface
The US was “out” during EDA Extension and ITER Transitional Agreement (but TBWG continued) • The TBWG was re-established in new conditions (6 partners) in 2003 • TBWG Scope of Activities: • Provide the design documentation for the assessment of TBMs prepared by the Parties, integration of TBMs into the ITER machine (including physical interfaces, auxiliaries, support facilities, machine operation, safety, waste management, reliability and maintenance) • Promote cooperation among the ITER Parties’ TBMs • Verify integration of the TBMs in the safety and environmental evaluations of each ITER candidate site • Further develop the coordinated blanket test program • Reporting (elaborate and specific) -- For Activities A, C and D: “report” to the Interim Project Leader and the Preparatory Committee. “Inform” the PTs leaders -- For Activity B: “report” to the PTs leaders
Current TBWG Effort, Port Allocation, and Preferences of the Parties • Since TBWG was reactivated in October 2003, several meetings have been held to: 1) discuss and agree on port allocation and concepts to be tested, 2) plan for international collaboration, and 3) provide ITER with details impacting the ITER device and plant (e.g. plasma exposure, off-normal events, ancillary equipment). • There are 3 ports for testing - There is a Port Master for each port • Five Working Subgroups were formed to examine technical issues and coordinate among the parties
Blanket Concepts for ITER-TBM Selected by the Various Parties • Solid Breeders • He/SB/Be/FS: All parties are strongly interested • H2O/SB/Be/FS: Only Japan (some interest from China) • Liquid Breeders • He/LiPb/FS (Separately cooled): EU lead (one of two main concepts for EU, interest from other parties) • Dual Coolant (He/LiPb/FS with SiC): US lead, strong interest from EU and other parties • Li/V (Self-cooled): Russia is main advocate (but no significant resources on R&D!) • Molten Salts: US and Japanese Universities want the option to decide later whether to test • He/Li/FS: Korea’s proposal
Port Allocations for ITER TBM Port Allocations for ITER TBM *Members nominated by each interested party (not necessarily members of TBWG). *Members nominated by each interested party (not necessarily members of TBWG).
IV. US Strategy for ITER TBM and Brief Description of Effort and Selected Concepts
Redirecting US Plasma Chamber Systems Effort to support ITER • With the US rejoining ITER, the Blanket/Chamber community concluded that it is very important for the US to participate in the ITER Test Blanket Module (TBM) Program (March 2003). • Extensive deliberations have occurred in the US since March 2003 among the community, DOE and VLT. • Reached consensus on a general framework for the direction of activities in the US Chamber/Blanket Program: • Provide fusion nuclear technology (FNT) support for the basic ITER device as needed • Participate in ITER TBM program and redirect good part of resources toward R&D for TBM • Encourage partners in international collaborations, such as JUPITER-II to focus more on ITER TBM • Important work has been carried out to implement the strategy. • A study of ITER TBM issues and US options was initiated • Some R&D was initiated • Rejoined TBWG, strong participation • The US interacted with all the other 5 parties to identify areas of collaboration
Highlights of US Strategy for ITER TBM(Evolved over the past year by the community, DOE and VLT) • The US will seek to maximize international collaboration. There is a need for all parties to collaborate, and to possibly consider a more integrated plan among the ITER parties for carrying out the R&D and construction of the test modules. • ITER TBM should be viewed as a collaborative activity among the VLT program elements. While the Blanket/Chamber Program provides the lead role for ITER TBM, major contributions from other programs (e.g., Materials, Safety, PFC) are essential. • The US community has now reached consensus on preferred options for ITER TBM (see separate slide), following assessment of new technical results obtained over the past few years.
What should the TWO US Blanket Options be for ITER TBM? • This has been a central question for the US community since March 2003. A study was initiated to select the two blanket options for the US ITER TBM in light of new R&D results from the US and world programs over the past decade. • Key conclusions reached early in the study: • - Selection between solid and liquid breeders can not be made prior to fusion testing in ITER. • - All Liquid Breeder Options have serious feasibility (“Go/No-Go”) issues. Need assessment. • - Solid breeders are accepted by all parties. • For the past year, the study has focused mostly on assessment of the critical feasibility issues for liquid breeder concepts. Examples of issues are MHD insulators, MHD effects on heat transfer, tritium permeation, corrosion, SiC insert viability, and compatibility. • The study has been led by the Plasma Chamber community with strong participation of the Materials, Safety and PFC Programs. Many international “Experts” in key areas participated in several meetings and provided important input.
US Selected Options for ITER TBM The initial conclusion of the US community, based on the results of the technical assessment to date, is to select two blanket concepts for the US ITER-TBM with the following emphases: • Select a helium-cooled solid breeder concept with ferritic steel structure and neutron multiplier, but without an independent TBM (i.e. support EU and Japan using their designs and their TBM structure and ancillary equipment). Contribute only unit cell and sub-module test articles that focus on particular technical issues of unique US expertise and of interest to all parties. (All ITER Parties have this concept as one of their favored options.) • Focus on testing Dual-Coolant liquid breeder blanket concepts with potential for self-cooling. Develop and design TBM with flexibility to test one or both of these two options (decision in 1-3 years): • a helium-cooled ferritic structure with self-cooled LiPb breeder zone that uses SiC insert as MHD and thermal insulator (insulator requirements in dual-coolant concepts are less demanding than those for self-cooled concepts); • a helium-cooled ferritic structure with low melting-point molten salt. Because of the low electrical and thermal conductivity of molten salts, no insulators are needed. (The key issues for molten salt are being addressed under JUPITER-II and no additional work is planned under ITER-TBM.)
Helium-cooled stiffening grid Breeder unit FW channel Helium-Cooled Pebble Breeder Concept for EU • The US can provide small breeder units “inside” the EU SB structure. • US Issues: Tritium Release and Thermomechanical Interactions
Module box(container & surface heat flux extraction) Breeder cooling unit (heat extraction from PbLi) [0.5-1.5] mm/s [18-54] mm/s Stiffening structure (resistance to accidental in-box pressurization i.e He leakage) He collector system (back) HCLL PbLi flow scheme EU – The Helium-Cooled Lead Lithium (HCLL) DEMO Blanket Concept
RAFS/He/Pb-17Li Dual Coolant Blanket Concept • The reason fusion pursued high temperature materials is for high coolant temperature for efficient power conversion. • MHD effects in high-velocity channel flows leads to very high primary stresses that materials must accommodate. • IDEA – the Dual Coolant Concept: use the poor thermal and electrical conductivity of SiC as an advantage. • Cool structure with He so LM velocity can be low and LM bulk temperature can be higher than the wall temperature. • Use a SiC insert to electrically and thermally insulate the LM from the wall. • Result: potential for high bulk temperature with lower MHD pressure drop using RAFS.
Close-up of a Pb-17Li Breeder Channel Self-Cooled Breeder Zone (from ARIES-ST) Dimensions in mm
Structure, Insert, and Breeder Temperatures FS grid Temperature drop across the FCI is 175 C
Dominant Issues for Dual Coolant Blankets: FCI Properties and Failures A) Electrical and thermal conductivity of the SiC/SiC perpendicular to the wall should be as low as possible to avoid velocity profiles with side-layer jets and excess heat transfer to the He-cooled structure. B) The inserts have to be compatible with Pb-17Li at temperatures up to 700-800°C. C) Liquid metal must not “soak” into pores of the composite in order to avoid increased electrical conductivity and high tritium retention. In general “sealing layers” are required on all surfaces of the inserts. - Even if the change in conductivity results in modest increase in pressure drop, it could seriously affect flow balance. D) There are minimum primary stresses in the inserts. However, secondary stresses caused by temperature gradients must not endanger the integrity under high neutron fluence. E) The insert shapes must be fabricable and affordable.
V. Programmatic Issues Requiring DOE and US ITER Management Attention
V. Programmatic Issues Requiring DOE and US ITER Management Attention • Include ITER-TBM in US ITER Organization • a “box/area” for TBM in the Organizational Chart • budget from DOE can remain separate from ITER construction funds • Recognize ITER-TBM as a US Program and organize it as a community effort • Led by Plasma Chamber but with major contributions from Materials Program and significant contributions from PFC, Safety, and Tritium Programs • Form Steering Committee with members from TBWG , Plasma Chamber, Materials, PFC, Safety, etc. • Resources Needed • Technical expertise is available • Required financial resources will be estimated over the next few months, but preliminary estimates show they are modest, not much beyond present budgets (mostly “refocus” and “rebalance” with modest increases, if possible) • Resources consist of: • annual R&D: starting now; already started under Plasma Chamber with some support from Materials, Safety , and PFC. Need to do more under other programs. • Need a lot more from the Material Program in particular (discussions to refocus parts of Materials to serve ITER-TBM needs is underway) • cost of constructing Test Articles (to be inserted in ITER): needed 7 years from now . It is modest and can be adjusted based on the role US wants to play.
V. Programmatic Issues Requiring DOE and US ITER Management Attention (Cont’d) • Formal International Agreement for ITER TBM • Letter from TBWG to ITER Interim Project Director and ITER PT Leaders (7/22/2004): “ It has become abundantly clear that planning for and conducting the ITER TBM program requires a structured framework among the Parties either as a new arrangement under the ITER Joint Implementation Agreement or even as part of it……..” • Recommend the US support this