1

ITER Test Blanket Module (TBM)

and

ITER Nuclear Science and Engineering

Mohamed Abdou

VLT/USIPO Meeting, October 20-21, 2004

2

Introduction

Recent Progress in ITER TBM Program

Issues and Recommended Approach

FY05 and FY06 Plans

Recommended Approach for Increased Effectiveness

Outline

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Introduction

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Integrated testing of breeding blanket and first wall components and materials in a Fusion Environment

Breeding Blankets/FWs 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.

What is the ITER TBM Program?

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Blanket Testing in ITER is one of ITER’s Key Objectives

Strong international collaboration among the ITER Parties is underway to provide the science basis and engineering capabilities for ITER TBMs

Bio-Shield Plug

Cryostat Plug

Transporter

Cryostat Extension

TBM Frame & Shield Plug

Breeder Concentric

Pipe

Drain Pipe

FW

US Solid breeder submodule

EU HCLL Test Module

Conceptual Liquid Breeder Port Layout and Ancillary equipment

6

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)

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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” window 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. (If the D-T cycle is not feasible the plasma physics and technology research would be very different.)

- Examples of questions to be answered:

– Will tritium self-sufficiency allow low plasma-edge recycling? – Are advanced physics modes acceptable? – Is the “temperature window” for tritium release from solid breeders sufficient for

adequate TBR? – Can circulated liquid metal breeders and coolants be used? – Is there a blanket/material system that can exist in this phase-space?– Are nuclear predictive capabilities (codes, data) adequate to predict accurate

TBR in realistic system

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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

ITER Blanket Testing is Essential to (cont’d):

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Nuclear Science and Engineeringis Critical to ITER and to the US Fusion Program

Nuclear Science and Engineeringis Critical to ITER and to the US Fusion Program

ITER will produce neutrons at a rate higher than any nuclear facility ever constructed (including operating fission reactors). It will have energetic neutrons, gamma rays, activated components, tritium, etc.

Challenges in predicting the radiation field:• Complex geometry, large system

• “Deep radiation penetration” problem in predicting dose to external components and radiation exposure to personnel

ITER base design changes, TBMs, port-based diagnostic packages, licensing, etc. will need more accurate nuclear simulation to ensure robust design, safe operation, and correct interpretation of experimental results.

US has been a leader in Nuclear Science and Engineering for fusion. Work on ITER base device and TBMs helps build US knowledge, experience and competence needed to develop practical and safe D-T fusion devices. (Building competence takes decades.)

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Recent Progress in ITER TBM Program

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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 IEA and 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

US Plasma Chamber Systems/Blanket Effort has been redirected to support ITER

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What should the US Blanket Options be for ITER TBM?

This has been a central question for the US community during FY 2004. 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, but potential for higher performance. More assessment needed.

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.

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US Selected Options for ITER TBM

Select a helium-cooled solid breeder concept with ferritic steel structure and neutron multiplier, but without an independent TBM. 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 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: 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 MS are being addressed under JUPITER-II and no additional work is planned under ITER-TBM.)

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Helium-Cooled Pebble Breeder Concept for EU

FW channel

Breeder unit

Helium-cooled stiffening grid

- The US can provide small breeder units “inside” the EU SB structure.

- US Issues: Tritium Release and Thermomechanical Interactions

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ITER Test Blanket Module Activities

Active participation in ITER test blanket working group (TBWG) for test and infrastructure planning WSG participation Ancillary equipment and machine interface definition

Evaluate blanket options for DEMO and evaluate R&D results for key issues to select primary US blanket concepts

Perform concurrently R&D on the most critical issues required MHD flow with insulators and inserts Tritium recovery and control SiC inserts compatibility and failure modes Solid breeder / multiplier / structure / coolant interactions

Develop engineering scaling and design, in collaboration with ITER partners, for TBMs.

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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

During EDA, the US made major contributions to testing strategy, engineering scaling, test port, frame design, and machine interface

TBWG Main Objectives: Develop a Coordinated Blanket Test Program and address the Interface between machine and the blanket modules

The TBWG was re-established in new conditions (6 partners) in late 2003 and has had 3 meetings in FY04

A. Provide the design documentation for the assessment of TBMs prepared by the Parties, integration of TBMs into the ITER machine

B. Promote cooperation among the ITER Parties’ TBMs

C. Verify integration of the TBMs in the safety and environmental evaluations of each ITER candidate site

D. Further develop the coordinated blanket test program

TBWG Scope of Activities:

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Two types of Solid Breeder TBMs have been studied in FY04

1. Unit cells (3)192.5 mm x 211 mm x 650 mm

2. Quarter-port Submodules730 mm x 910 mm x 600 mm

• The proposal calls to share the port space to test contemporaneously independent unit cells/submodules

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Various analyses have been performed for TBM designs to preserve key prototype parameters

Stress analysis shows max = 268.4 MPa located at the

corner of the front inner wall

Maximum

temperature at the

first wall = 528o C

5 channels

per flow path

He In

He out

The displacement profile shows a non-uniform distribution with a maximum displacement of 3.51 mm

SB

Be

He in He out

Unit cell temperature profile (edge-on design)

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Sophisticated 2-D neutronics analysis shows testing objective can be achieved for a proposed NT TBM

0

1 10-5

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Distance from Frame, cm

Left Configuration Right Configuration

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Left TBM WallBe Layer-Left Config.Left VCP-Left Config.Br1Right TBM WallBe Layer- MiddleBe-Rt. SubmduleBe Layer-Rt. Config.Rt. VCP- Left Config.Left VCP-Rt. Config.Rt. VCP-Rt. Config.

Toroidal Distance from Frame, cm

Depth = 42 mm behind FW

Breeder (Lft. Config.)

Be (Rt. Config.)

Proposed NT TBM

JA TBMTritium production profiles are nearly flat over a reasonable distance in the toroidal direction allowing accurate measurements be performed

Finding:Flat nuclear heating and tritium production profiles allow two designs to be evaluated in a ¼ port submodule

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The reason fusion pursued high temperature structural materials is for high coolant temperature.

MHD effects in high-velocity channel flows leads to very high primary stresses.

IDEA of the Dual Coolant Concept*

FS/He/PbLi Dual Coolant Blanket Concept: Higher Performance Potential

with current generation ferritic structure

– Cool structure with He so that FS can be used. “Decouple” surface from bulk heating.

– Flow PbLi for self-cooling at low velocity

– Use a SiC insert to electrically and thermally insulate the LM from the wall, so LM bulk temperature can be higher than the wall temperature (use the poor thermal and electrical conductivity of SiC as an advantage).

– Result: potential for high bulk temperature with lower MHD pressure drop using Ferritic Steel.

Self-Cooled

Breeder Zone

ARIES-ST DCLL Breeder Unit

*Dual Coolant concepts proposed by ARIES and EU

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Structure, Insert, and Breeder Temperatures

Temperature drop across the FCI is 175 C

FS grid

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Complex geometry MHD codes already being applied to DCLL blanket with SiC

Flow Channel Inserts 2D and 3D codes

(developed for Liquid walls) have been modified for DCLL

Initial results show strong sidelayer jets at SiC = 500 S/m with current DCLL design

2D and 3D codes give conflicting results concerning flow in the “stagnant” gap region.

Code improvements and debugging, and continued simulations planned for FY05.

Slice from 3D Simulation

Velocity profile from2D Simulation

Strong negative flow jet near pressure equalization slot not seen in 3D simulation

Gap corner jets not seen in 2D simulation

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Study of DCLL system and SiC FCI feasibility has shown:

SiC flow channel inserts in DCLL systems have an interesting combination of effects that reduce pressure drop as compared to self-cooled LM systems Sensitivity to flaws and flow in complex geometry flow elements will dominate

pressure drop and must be carefully examined Flow balancing technique must be analyzed and tested in more detail. SiC FCI properties and fabrication must be established more definitively

Compatibility of FCI in a PbLi/RAFM steel system is a critical development issue Compatibility and corrosion are linked issues with DCLL design and MHD

behavior More definitive first and second level compatibility experiments and modeling

are necessary – definition of needed experiments could be an IEA Liquid Breeder Topic

Tritium control is a key issue which will also be affected by the baseline MHD and subsequent FCI compatibility and corrosion

Effective DCLL testing in ITER is possible and desirable MHD testing building towards integrated FCI performance Synergism with EU PbLi and RAFM program, Scaling and test program planning in the US underway

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Technical Details from the US ITER-TBM efforts this year were presented at ANS TOFE Meeting

in Madison

e.g. Abdou, "US Plans and Strategy for ITER Blanket Testing" Morley, “Thermofluid Magnetohydrodynamic Issues for Liquid

Breeders” Ying, “Engineering Scaling Requirements for Solid Breeder Blanket

Testing” Smolentsev, “MHD Effects on Heat Transfer in a Molten Salt Blanket Youssef, “Activation Analysis for Two Molten Salt Dual-Coolant

Blanket Concepts for the US Demo Reactor” Wong, “Assessment of Liquid Breeder First Wall and Blanket Options

for the Demo Design” Sawan, “Neutronics Assessment of Molten Salt Breeding Blanket

Design Options” Youssef, "On the Strategy and Requirements for Neutronics Testing

in ITER“ …

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J2 thermofluid experiments establish reference database for molten salt

turbulence and heat transfer

Careful turbulence measurements using PIV technique show excellent agreement with Direct Numerical Simulation in near wall and in channel core (see figs right )

Heat transfer reference cases also showing good agreement with standard correlations.

Special 2T gap magnet with unobstructed viewing access for laser diagnostics designed and nearing completion. To be installed at UCLA in Dec. 2004.

100 101 1020

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20

y+

u+

Umean(DNS) Umean(PIV)

Re=5300

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Re = 5286 Re = 11,300

Re stress Re stress

Ave Velocity Ave Velocity

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Two types of Solid Breeder TBMs have been studied in FY04

1. Unit cells (3)192.5 mm x 211 mm x 650 mm

2. Quarter-port Submodules730 mm x 910 mm x 600 mm

• The proposal calls to share the port space to test contemporaneously independent unit cells/submodules

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Simulation and experimental measurement of

CERAMIC BREEDER PEBBLE BEDs

Due to the high stress at the local contact area, especially at the high temperature, creep deformation occurs.

Discrete Element Method (DEM) is able to capture thermal deformation characteristics and can provide the detail stress distribution inside the pebble beds at both steady and transient states.

Force distribution in the pebble bed simulated by DEM – line widthgives a measure of contact force

Li4SiO4/Li2O pebble beds

plate

Capacitance displacement sensor

Time (hour)

Cre

ep

Co

mp

act

ion

stra

in(%

)

2 4 6 80

0.05

0.1

0.15

0.2

0.25

0.3

Experimental Data (700oC)

Experimental Data (750oC)

Experimental Data (800oC)

FEM (700oC)

FEM (750oC)

FEM (800oC)

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Lithium Ceramic (Solid Breeders) Pebble Bed thermal conductivity data collected from all world

fusion program experiments

Significant work this year at UCLA to design an experimental apparatus to measure effective conductivity of

Li-ceramic pebble bed for temp. (400 to 850C)

interface conductance for temp. up to 550C.

Published studies show a wide spread in data, little error analysis, and conflicting temperature dependence

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Issues and Recommended Approaches

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Tritium Self-SufficiencyTritium self-sufficiency condition: Λa > Λr

Λr = Required tritium breeding ratioΛr 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 ratioΛa 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.) Uncertainties in nuclear data required for accurate determination of TBR

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.

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Tritium Consumption and ProductionTritium 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

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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)

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w/o Fusion

ITER-FEAT(2004 start)

1000 MW Fusion,

10% Avail, TBR 0.0

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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

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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

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The US plans on unit cells and submodule tests to address specific technical issues such as temperature window and temperature control (not a fully independent TBM, uses other Parties ancillary equipment)

The world program (particularly EU and Japan) in this area is strong

There are very good reasons why the US program should contribute to this area

1. the only universally accepted concept,

2. exciting issues with a lot of science,

3. Nearest term breeding capability for extended phase in ITER

Strategy for Testing Solid Breeder (He/SB/Be/FS) Concepts

36

Present U.S. Solid Breeder R&D EffortCarried out mostly in collaboration with other countries (IEA, JUPITER-II)

Solid Breeder Blanket Specifics: Focus on niche areas of solid breeder blanket material system

thermomechanics interactions (Primary organizations: UCLA, Support: ORNL, PNL)

• design database on effective thermo-physical and mechanical properties for breeder and beryllium pebble beds

• experiments and modeling development on evaluation of thermomechanical states of blanket element pebble beds under different loading conditions

Material/Blanket Experiments Interface: Development of Web based INTEGRATED FUSION MATERIALS

DATABASE (UCLA, UCBS, ANL, ORNL) Construction of VISTA (VIRTUAL INTERNATIONAL STRUCTURAL

TESTING ASSEMBLY) modelling tool, to evaluate a range of potential interactions and failure paths (perform “Virtual Experiments”). (UCLA, UCBS, FZK, ANL, ORNL)

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DCLL Test Phases in ITER

H and D Specific electromagnetic structure and MHD TBM:

Structural TBMs reaction to field environment and various transient plasma events – for instance “water hammer” effect during rapid plasma current quench

Scaled pressure drop tests, flow balance test and critical velocity profiles affecting heat transfer in increasing field strength

SiC FCI integrity under MHD loading

Low Duty DT

Thermofluid TBMs:

Scaled stress response of thermally loaded SiC inserts and the effects of failures on pressure drop and thermal field

Tritium production and permeation

High Duty DT

Partially-integrated Thermafluid TBMs:

Corrosion and compatibility in nuclear field Radiation damage ~1 dpa in the inserts Tritium control

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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.

DCLL Concepts Issues:SiC FCI Properties

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DCLL Concept Issues:MHD effects in self-cooled liquid metal

breeding regions Primary issue for blanket application and ITER testing is the

MHD pressure and flow distribution for complex geometry flow elements: SiC FCI overlap regions (stovepiping) Defects in FCIs Flow balancing sections Turns, Field entrance/exit regions, Manifolds, Expansions/Contractions Coaxial Pb-Li supply/return lines

MHD velocities profiles can exhibit strong jets next to regions of stagnation and even reversed flow

Large temperature gradients can drive natural convection flows that MHD effects do not damp – can swamp forced flow velocity in slow moving breeder zone regions

All of these MHD issues, through effects on temperature, velocity shear and convection, and determination of LM-facing material, will have effects on the corrosion and tritium permeation of the concept

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• Tritium Permeation is emerging as a persistently serious issue for most (some believe for ALL) concepts

• Developing acceptable solutions requires integrated efforts among many programs: Materials, Tritium, PFC, Safety, Plasma Chamber

Tritium Permeation

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FY05 and FY06 plans

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ITER TBWG Participation, Planning and

Documentation Requirements

ITER TBWG ~quarterly meetings and contribute to the final TBWG report on test strategy

For both Solid Breeder and Dual Coolant Pb-17Li Concepts, prepare and deliver to TBWG, a complete preliminary Design Description Document containing TBMs design Fabrication Qualification program plans Optimization of TBM system Provide initial TBM delivery date

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Solid Breeder and Multiplier tasks

Participate in the collaborative tasks that have been defined in the IEA solid breeder (and liquid breeder subgroups), in which R&D results improve/impact TBM designs• design database on effective thermo-physical and mechanical

properties for breeder and beryllium pebble beds• experiments and modeling development on evaluation of

thermomechanical states of blanket element pebble beds under different loading conditions

44

IEA collaboration on solid breeder pebble bed time dependent thermomechanics

interactions/deformationPrimary Variables• Materials• Packing• Loadings• Modes of operation

Irradiation Effect(NRG)

Primary & Secondary Reactants:• Temperature magnitude/ gradient• Differential thermal stress/contact pressure• Plastic/creep deformation• Particle breakage• gap formation

Goal:Performance/Integrity prediction & evaluation

Partially integrated out-of-pile and fission reactor tests (NRG,ENEA)

Finite Element Code (ABQUS, MARC)

(NRG, FZK, UCLA)

Discrete Element Model (UCLA)

Design Guideline and Evaluation (out-of-pile & in-pile tests, ITER TBMs)

Database Experimental Program

(FZK, JAERI, CEA,UCLA)

Thermo-physical and Mechanical PropertiesConsecutive equations

Single/multiple effect experiments(NRG, UCLA)

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Pb-17Li dual coolant blanket concept TBM

Main R&D Tasks, Multi-year:

Strong effort on development of simulation capability (see slice of 3D simulation, right) and experimental test plan, facility construction for LM-MHD effects in closed channels with complex geometry and heterogenous wall conductivity

Assessment, compatibility and property experiments on the feasibility of the SiC insert

DCLL Reference and scaled TBM design for dual-coolant PbLi concepts in collaboration among US Plasma chamber community

Example simulation: MHD boundary layer jets formed in a channel with SiC flow channel insert

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What experiments needed to determine Pb-Li SiC compatibility?

FY05 - More definitive capsule experiments required to show first level of compatibility More prototypic SiC composite should be used Long exposure time to overcome any incubation period for wetting Pressurization, impurities typical of RAFM steel Careful post-examination of both SiC crucible for evidence of attack and

PbLi melt for accurate concentration of dissolved SiC As an intermediate step suitable models should be developed and the

basic material properties (most important diffusivity and saturation concentration as functions of temperature) should be obtained from the literature when available, and from dedicated experiments as necessary.

FY06 and beyond – Quantification of the allowable interface temperature in a corrosion flow loop with relevant temperature gradients, typical materials, and realistic flow conditions (Provided SiC/PbLi passes first level of assessment )

47

All experiments are directed to solve key feasibility issues for the molten salt, Li/V and solid breeder ITER test modules.

1. Flibe REDOX control Completed flibe purification process. Completed flibe mobilization experiment. Started hydrogen isotopes (D) permeation, diffusion and solubility measurements. Preparation for the REDOX experiments. Presented two papers at ICFRM and Be workshop.

2. Flibe heat transfer and flow mechanics Measured straight pipe velocity profile and turbulent statistics. Constructed 304SS heat transfer test section, with some initial data available. New acrylic PIV attachment section constructed and tested. Prepared for the MHD experiment with a US supplied magnet.

3. MHD coating development Coatings of AlN, Y2O3 and Er2O3 have been tested in Li up to 800C. Vacuum distillation system developed and tested to remove residue lithium from test coupons. Resistivity experiments conducted for Er2O3, Y2O3 and (Y,Sc)O3 to confirm sufficient resistivity for

MHD coating.

4. SiC/pebble bed thermomechanics experiments Two configurations were developed.

JUPITER-II collaboration for Plasma Chamber Systems

48

Nuclear analysis for ITER TBM

• This effort is ongoing and funded as part of the US ITER TBM activities

Nuclear support for basic ITER Device• Nuclear support is needed particularly for US preferred procurement packages (diagnostics,

plasma heating and CD, module 18 (baffle), CS coils)

• This support will ensure that the US delivers components with great confidence in its successful performance in ITER nuclear environment

• ~$200K (~1 FTE) a year needed. Small burden to the US ITER construction budget to support the other programs

• The US neutronics community (primarily Universities) has strong interest in providing this support. This allows training students and young scientists

Development of CAD/Neutronics coupling• Needed to facilitate the 3-D neutronics analysis of ITER components effectively• Allows using geometric models directly from the actual design system, resulting in

more accurate results and reduce time between CAD-based design changes and

analysis• ITER International Team requested that support

• Resources needed ~$100K (~0.5 FTE) per year for two years

ITER Nuclear Analysis Needed in Three AreasITER Nuclear Analysis Needed in Three Areas

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TBM designs will be developed and modeled for 3-D neutronics calculations with all design details

The TBM 3-D model will be integrated in the complete basic ITER machine 3-D model

Perform 3-D neutronics calculations using the integrated model

Neutronics calculations will provide important nuclear environment parameters (e.g., radiation damage, tritium production, transmutations, radioactivity, decay heat, and nuclear heating profiles in the TBM) that help in analyzing TBM testing results

Nuclear Analysis for TBMNuclear Analysis for TBM

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Detailed 3-D nuclear analyses needed in support of US procurement packages

Radiation leakage through holes and other penetrations must be fully assessed to establish activation levels for personnel access

Nuclear Analysis for ITER Basic MachineNuclear Analysis for ITER Basic Machine

As ITER proceeds to construction, individual component designs become more clearly defined and more accurate nuclear analyses are needed. Accurate evaluation of the nuclear radiation level at these components is essential to assess their survivability and performance capability

Special attention should be devoted to components that are sensitive to radiation and have penetrations with radiation streaming concern

Module 18

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Use Sandia’s CGM interface to evaluate CAD directly from MCNP» CGM provides common interface to multiple

CAD engines, including voxel-based models

Benefits:» Dramatically reduce turnaround time from

CAD-based design changes– Identified as key element of ITER

Neutronics analysis strategy» No translation to MCNP geometry commands

– Removes limitation on surface types– Robustness improved by using same engine for CAD and MCNP

» Can handle 3D models not supported in MCNP Status: prototype using direct CAD query from MCNP Issues/plans:

Benchmarking the current prototype version of MCNP/CGM for ITER analyses. Explore various approaches to transferring geometry from the ITER CAD design system (possibly CATIA) to the ACIS-based MCNP/CGM

The most immediate limitation is that it is slower than MCNP alone. The focus will be to speed up the ray-tracing portion of the code (lots of acceleration techniques possible)

Goal: speed comparable to MCNP, but using direct CAD evaluation

CGMACIS Pro/E Voxels

MCNPMCNPNative

Geometry

ARIES-CS PlasmaClothespin, w/helical

spring surface

CAD-Based MCNP DevelopmentCAD-Based MCNP Development

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Recommended approaches to increased effectiveness

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Programmatic Issues Requiring DOE, VLT, and US ITER Management Attention

Include ITER-TBM and Nuclear Science and Engineering in US ITER Organization a “box/area” for TBM in the Organizational Chart

– this is essential to keep ITER link for TBMs that will influence basic machine design and building space

– budget from DOE can remain separate from ITER construction funds

a “box/area” for Nuclear Science and Engineering– Small budget should come from ITER funds to support US

procurement packages in diagnostics, module 18, etc.

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.

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Programmatic Issues Requiring DOE, VLT, and US ITER Management Attention (Cont.)

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)

Resources consist of:a) 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)

b) 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.