ITER Testing and Required RDs for Solid Breeder

ITER Testing and Required R&Ds for Solid Breeder TBMs Alice Ying/Mohamed Abdou Presented at APEX Meeting Nov. 5, 2003 UCLA

Current US Strategy for Solid Breeder Blanket TBMs Background: He-Cooled Solid Breeder Blanket with FS has been selected by all 6 parties and has the largest world R&D The US Strategy calls to Select He/SB/FS as an option but does not have an independent TBM. Rather, it plans on unit cell and submodule test articles that focus on particular technical issues of interest to all parties Objective of this presentation is to present a first cut scheme to implement this strategy

Underlying Rules in a Supporting Role For the US to serve as a supporting member for the solid breeder blanket TBMs, it implies that: • The US will adopt the DEMO SB blanket design of other countries (most probably EU) • The US will only test issue-specific submodules, which are incorporated into a port module. • The US will not play a major role in fabrication technique development. This includes breeder material fabrication, beryllium (or berylite) pebble fabrication, and structural fabrication for blanket module box. The US will procure pebble materials from JA and EU for use in all planned laboratory experiments. We have to pay JA or EU for materials. Is this an acceptable ground rule for the US? • The US will not develop auxiliary systems including the high pressure helium-coolant loop and tritium processing systems.

Summary of Critical R&D Issues for Fusion Nuclear Technology 1. 2. 3. 4. 5. 6. 7. 8. 9. D-T fuel cycle self-sufficiency Issues/Phenomena that 2. Tritium inventory and recovery in the may be explored during solid/liquid breeders under actual operating conditions ITER BPP for Solid Breeder TBM 3. Thermomechanical loadings and response of • Thermomechanics blanket and PFC components under normal and off • Tritium recovery, inventory, and normal operation permeation 4. Materials compatibility • Temperature window 5. Identification and characterization of failure for ceramic breeders modes, effects, and rates in blankets and PFC’s • Nuclear performance • Data on failure modes 6. Effect of imperfections in electric (MHD) insulators in liquid metal cooled blanket and PFC under thermal/mechanical/electrical/nuclear loading 7. Tritium permeation and inventory in blanket and PFC 8. Radiation Shielding: accuracy of prediction and quantification of radiation protection requirements 9. Lifetime of blanket, PFC, and other FNT

We should strive hard to test parameters and conditions critical to tritium fuel self-sufficiency ü Critical for testing the viability of the DT Fuel Cycle. Also, critical for tritium supply issue in devices beyond ITER. ü The one issue that physics–dominant fusion community tend to understand appreciate most ü A sophisticated neutronics analysis using ITER geometry to define one quarter port size blanket configuration with TBR requirement and tritium transport characteristics that would satisfy fuel self-sufficiency requirement for a set of Key Fusion Reactor parameters (doubling time, fractional burnup, days of tritium reserve, extraction efficiency, residence time in plasma exhaust system, blanket tritium inventory) ü Features of Neutronics and Tritium Transport TBM v First wall –Look-Alike (no first wall thickness increase considered) v Blanket zone- Act-Alike in order to obtain DEMO-type temperature magnitudes for tritium transport

Proposed US Issue Specific Submodule Tests 1. Thermomechanic Tests [Appropriate port location: EU-coordinated Hecooled SB Test Port] q 2 Act-alike submodules q 1 Look-alike submodule for data interpretation q Each submodule is about one tenth of the port size 2. Neutronics and Tritium Transport Tests [Appropriate port location: JA-coordinated Hecooled SB Test Port] 1. One quarter of port size module.

Why Thermomechanic Tests? Thermomechanic performance of breeder and beryllium pebble beds (under temperature, stress, and irradiation loading conditions) is the key feasibility issue for solid breeder blanket concepts ITER is the only machine that offers environmental conditions for performing the tests: q a large test volume q a correct neutron spectrum and profile q a volumetric nuclear heating These conditions are impossible to obtain outside of a fusion testing environment

Neutronics and Tritium Transport Tests Objectives ² To check and validate the capability of the neutronic codes and data to predict the nuclear responses in the TBM. This includes the ability to accurately predict the tritium breeding in the TBM. ² To check and validate the capability of the tritium transport code. This can be achieved by comparing the measured and calculated tritium concentration in the helium purge gas line over a given irradiation time. Both are part of important processes involved in tritium fuel self-sufficiency assessment.

Types of ITER-FEAT Test Port Configuration with Common Frame TBM Attachment on Common Frame Shield plug is set behind the TBM. Shield plug has feed through for pipings and instrumentation. TBM is held on the shield plug by a similar mechanism as the ITER shield blankets (flexible support). Attach modules and feed pipes through shield plugs of rear side. 534 1288 1740 760 Plasma Port 18 (water cooled) Port 1 (helium cooled)

Presently Proposed JA He Cooled TBM Layout may be Modified to Accommodate the US Neutronics/ Tritium Transport Test Module Port 1 (helium cooled) 1288 Configuration can be modified to test US submodule. Need US specific purge gas line and helium coolant line for temperature and purge gas composition control First Wall 1280 1 st Breeder Layer 1 st Multiplier Layer 2 nd Multiplier Layer 3 rd Breeder Layer 2 nd Breeder Layer 600 Cooling Tube 760 3 rd Multiplier Layer 4 th Breeder Layer 4 th Multiplier Layer Coolant manifold

EU Demo Helium-Cooled Solid Breeder Blanket Concept has been modified Significantly to Account for the Structural Box Survivability under an unlikely accident scenario

Breeder Unit for EU HCPB Concept

Possible Configuration of the new EU-HCPB TBM ² The US could insert 3 thermomechanic submodules into this port configuration. ² The challenge is to have an engineering manifold design to accommodate additional three independent inlet/outlet coolant lines.

Alternative way to utilize the port to incorporate independent coolant lines to submoduels Such an alternative way is to allocate a fraction of the port specifically for the US TBMs: • May require US’ own auxiliary systems • Require sophisticated test port engineering designs to accommodate large numbers of service lines • Larger R&D and testing required 1/2 half-port space to allocate neutronics and tritium transport test Port 1 (helium cooled) 1288 760 1/3 to 1/4 half-port space to allocate 3 thermomechanic submouldes

Critical Factors that Impact US Solid Breeder TBM Program

Exposure to Plasma ðExposure to plasma is important for testing in ITER in order to evaluate the performance of the first wall design. A first wall includes a front wall, a coolant channel, and a back wall. ðIn a typical solid first wall design, the first wall is connected to the side wall and the coolant manifold located at the back Thus, testing of a stand alone first wall may not give much value JA HCB First Wall Mock-up EU-First wall design

Time Schedule of Thermomechanics Submodules Overlaps EU ITER Time Schedule US TBM Submodule Group A Submodule Group B Submodule Group C

Time Schedule of Neutronics and Tritium Transport Submodules (3 Submodules in Sequence) Overlaps JA ITER Time Schedule Year Mile Stone 1 2 3 4 Full Field, Current & H/CD Power First Plasma Baking & Conditioning 5 Short Burn 200 MW 6 7 7 9 10 Fu Non-inductive Q=10 Cu s 500 MW 400 Current Drive D-Plasma (limited T) H-Plasma - Machine com. Low Duty DT - Machine commissioning • Commissioning with plasma High Duty DT - Achieve good with neutron - Development of full DT high Q Heating & CD vacuum & wall - Reference with D - Development of non-inductive - Improvement of inductive and nonexperiments condition - Short DT burn inductive operation aimed at Q=5 - Reference scenarios Demonstration of a high duty operation with H Equivalent Number of Burn 1 750 1000 1500 2500 3000 pulses (500 MW x 400 s) Operation Cumulative Neutron Fluence at Test Port (MWa/m 2) Installation & Commissioning Blanket Test Phase Basic Installation & Commissioning 9. 9 x 10 -6 0. 0074 For Activation Phase 0. 032 0. 057 For High Duty Operation 0. 087 0. 12 Upgrade System Checkout & Characterization TBM W-1, TBM H-1 TBM W-2, TBM H-2 - Overall functions, - Remote - Neutronics environment Submodule 1 instrumentation, flow control, Submodule 1 handling measurement US Neutronics and heat loss, etc. test - Indication of - Integrity to EM & thermal loads during Tritium Transport Test tritium production - FW heat removal shut & extraction - Ferritic steel effect on plasma down Indication of heat control generation Test Key Words 0. 017 System Checkout & Test Environment Measurement Neutronics TBR Measurement Test Performance Test TBM W-3, TBM H-3 Submodule 2 - High grade heat generation and extraction - Continuous tritium recovery - Electricity generation with TBM w-3 Thermo-mechanical Test T and Heat Production (long term)

R&D Issues for US solid breeder blanket TBMs Ø Material Fabrication/Procurement (solid breeder, beryllium, ferritic steel structure) Ø Testing Facility (for out-of-pile and in-pile nuclear submodule tests. For key issues such as ceramic temperature window, material thermomechnaics, interaction, effects of cycling, etc. ) Ø Fabrication Technology (FW, coolant structural panel, manifold) Ø Auxiliary Systems for TBM (helium cooling and tritium processing) Ø Instrumentation Development and Integration

Issues for US solid breeder blanket TBMs Breeder pebbles fabrication: In the past, we received pebbles from FZK and JAERI through the IEA collaboration free of charge for thermomechanic experiments. (In fact, JAERI had to purchase the pebbles from the Kawasaki Heavy Industry and ship them to us). Subsequently, the quality of the pebbles is beyond our control. We may receive pebbles which are produced for some research purpose, but are abandoned for use in reactor. For example, the Li 4 Si. O 4 we received included Ta dopping, which is not acceptable from the material compatibility point of view. The issue is how will we continue to obtain breeder pebbles for our experiments. Shall we specify what we like and purchase from the Kawasaki? or devote some resources to develop our own fabrication techniques? or what are the other options?

Beryllium Pebble Fabrication The “spherical" beryllium pebbles produced from Brush Wellman were part of the by-product during the production of metallic beryllium. The beryllium metal pebbles contain 97 percent beryllium along with entrapped reduction slag and unreacted magnesium. The NGK fabricate beryllium pebbles using a rotating electrode method from beryllium metal. NGK currently has research activity on producing beryllium intermetallic (Be 12 Ti) compounds. Issues: Shall we agree with the beryllium fabrication method from NGK and purchase pebbles as needed for near-term experiments? Shall we engage in developing beryllium intermetallic pebbles?

High Temperature Helium-coolant Loop Needed for Out-of-Pile Thermomechanics Tests Justification Characterization of the elastic and creep properties of the ceramic breeder and beryllium pebble beds under thermo-mechanical loads is necessary to ensure that bed integrity does not become the blanket lifetime limiting factor. Previous experimental results of ceramic breeder Li 4 Si. O 4 pebble bed thermomechanical tests at temperatures above ~650 o C (~ 0. 5 Tm) have shown a significant amount of thermal creep deformation, which is not revealed at lower operating temperatures. The existing Parathern NF coolant loop used to support thermomechanics experiments at UCLA provides a maximum coolant temperature of about 150 o. C (Not adequate at all!). This limits the present experiments to operating the bed at a maximum temperature of 400 o C, whereas a typical beryllium pebble bed operates at a maximum of 600 o C and a ceramic breeder pebble bed at 950 o C. Numerical simulation code development is underway, with the goals of predicting stress relaxation, thermal creep deformation, and bed integrity. However, data is needed to guide and verify model development.

High Temperature Helium-coolant Loop (Cont’d) Mission and Expected Results: To generate thermo-physical property data, and to understand obtain key parameters relevant to thermo-mechanical interaction phenomena at high operating temperatures for ceramic breeder and beryllium pebble bed material systems. The objective of such a high temperature coolant loop facility is to provide an adequately high temperature boundary condition needed for thermo-mechanical interaction experiments of solid breeder material system and thermophysical and mechanical property acquirements. High Temperature Helium Loop Functionality: Ø To support high temperature operations for thermo-physical and thermomechanics design database Ø To provide prototypical coolant medium for ITER TBM out-of-pile and mockup tests and for JUPITER II Task 2 Si. C/Si. C for evaluations of coolant tubes hermeticity and integrity of joining technique

Where will the nuclear submodule tests be performed? Ø JMTR will be shut down around 2006. Ø Test rig at HFR in Pattern tends to be small. EU HCPB Pebble Bed Assembly Irradiation at HFR Question Will the US propose to use HFIR or ATR for submodule nuclear tests? Note: no additional irradiation test planned in EU for TBM R&D 6. 75 cm diameter x 12. 5 cm height

Auxiliary Systems for He-Cooled Ceramic Breeder Blanket TBMs Helium Coolant Sys. -I Tritium Extraction Sys. - He Coolant Purification Sys. -I

A possible mean of having US independent coolant lines without major development on Helium Cooling System is to create a by-pass line (marked red) after the circulator EU HCPB TBM (2001) T P Heater US TBMS

2003 2005 2008 2010 2013 2015 2017 2018 LEGEND Breeder and Pebble Bed Characterization and Development Multiplier and Pebble bed Characterization and Development Blanket Thermal Behavior Advanced In-Situ Tritium Recovery (Fission Tests) Thermo-physical properties/design database E Terminate Task E OP Thermo-physical properties/design database E T Evaluation Point Operate Major Experiment Terminate Major Experiment Information Flow Unit Cell Thermomechanics E OP T Submodule Thermomechanics OP E T R&D Highlighted with Purple color to fit ITER Timeline Nuclear Submodule 1 E T OP E Nuclear Design and Analysis (Modeling Development) OP E Thermomechanics modeling Fusion Test Modules Design Fabrication and Testing Initiate Task ITER First Plasma T E Integrated modeling Prototype Mockup Testing Nuclear Submodule 2 E ITER Material and Structural Response Testing Component Design Issues Tritium Permeation and Processing Permeation Rate Measurement TBMs E T OP Fusion Test Module (e. g. CTF) OP Blanket structure fabrication Instrumentation Test Sequence for Major Solid Breeder Blanket Tasks E OP

Example of Major Activities Sequence for TBM Final input of the US TBMs to ITER JCT by Dec. 2004 Continue to dialogue with Parties on the proposed collaboration scheme Evaluate and Select a design concept Define test objectives and test parameters Develop conceptual TBMs to meet the proposed test objectives Perform nuclear, thermal and mechanics analysis 2004 -2008 Define and perform out-of-pile tests 2005 -2010 Define and perform In-pile tests Develop engineering design 2008 -2011 Define and perform mockup test Construct TBMS

US TBM ready for integration

Summary ¨ If agreed by major responsible parties, the US as a support role can reduce a significant amount of financial burden, yet obtain critical data for fusion blanket technology development ¨ Nevertheless, the US TBMs should be ready for integration into (e. g. EU) Port Module in 2012 to be inserted into ITER (per ITER current calendar) ¨ The 1 st WSG meeting for He-cooled ceramic breeder blanket TBMs will be held on Dec. 14, in Tokyo (right before CBBImeeting). Input on the US SB-TBM Program is needed for discussion on international collaboration. ¨ A complete design description document on the international Helium-cooled ceramic breeder blanket TBM program is due by December, 2004.

Summary (Cont’d) Would like to come up with a sum of 900 K effort on Solid Breeder Blanket TBM for 2004 25 K = 1 man-month effort