DCLL TBM:

Comments on Design Strategy, and

Experimental plans, schedule and required diagnostics

Compiled by Neil Morley for the TBM Meeting

May 10, 2006

My suggestions for near term Design focus (slight modification of clement’s)

Basic TBM fabrication and mechanical design– FS Fabrication, requirements and limitations– Application of FW heat transfer enhancement techniques– Routing and uncertainties in He and PbLi flow distribution (supply,

manifolds, routing)– Basic diagnostics placement and feedthroughs

Demonstration and optimization of DCLL DEMO characteristics– Desired DEMO operating point(s), loads, parameter ranges– Desired DEMO material properties targets– Scaling and geometric similarity of TBM experiments from DEMO

conditions to ITER D-T– Initial testing in ITER H-H

ITER Safety and Reliability Requirements– Licensing envelope and accident analysis– Allowable volume of PbLi– Design rules– Disruption response analysis

Proceed with most critical R&D

Thermofluid / MHD modeling tools progress and preparation of manifold MHD experiments (Smolentsev)

1st Generation FCI development and characterization (Katoh)

Material System Compatibility literature and existing data review (no presentation)

FS fabrication development using HIP, welding development, investment casting assessment study, test methods and ITER database (Rowcliffe/Kurtz)

Virtual TBM – foundations and technical planning (no presentation)

Diagnostics planning and survey (I will begin to discuss)

Current ITER Operational Schedule

TBM Name Sequential TBM Experimental GoalsITER

Phase & Duration

1st TBM

EM / Structural

Establish testing capability and system performance baseline and operation experience prior to D-T (nuclear) operation, including heat transfer and thermal time constants

Validate general DCLL TBM Structure and FCI response to EM/Plasma environment

Perform initial studies of MHD effects in ITER fields (pressure drop and flow balancing)

HHfor

2.5 years

2nd TBM

Nuclear Field/Tritium Production

Neutron field measurements database for various types of ITER discharges and conditions

Measure tritium production rate (TPR), and nuclear heating rates FW He cooling at full load and tritium implantation effects

DD +Early DT

for2 years

3rd TBM

Thermofluid / MHD

Thermal and electrical insulation properties of the FCI and FCI failure modes and effects

Tritium permeation through FCIs Velocity measurements with FCI gap flow and natural convection Initial data on activation products and chemistry control

Low duty DT for2 years

4th TBM

Integrated

Investigate high temperature TBM operation including flow channel inserts behavior, effect on tritium permeation and corrosion and activation products

Investigate online tritium recovery from PbLi and He streams Investigate online PbLi purification systems Explore longer term Integrated operation of the system including small

accumulation of radiation damage in FCIs and RAFS joints

High duty DTfor

3 years

ITER’s Description of Early ITER Phases

Integrated Commissioning - This phase completes the construction of ITER by ensuring all systems operate together and includes the preparation of the machine to attain the first hydrogen plasma.

Hydrogen Phase - This phase allows full commissioning of the tokamak system in a non-nuclear environment without depending on fully-remote handling.

Major aspects of the full DT discharge scenario can be checked, including: – plasma current initiation, current ramp-up & ramp-down, formation of a divertor configuration, – control of and loads due to disruptions or vertical displacement events, – access to H-mode and adequacy of heating installed, – operational limits on density, beta, safety factor, – requirements/capability of steady state operation.

The peak heat flux onto the divertor target will be of the same order of magnitude as for the full DT phase.

Some important issues cannot be fully tested in this phase. These include: – evaporation of the divertor target surface expected during a disruption, – effects of neutron irradiation of the in-vessel materials, – alpha-particle heating of the plasma.

Although there are no neutrons in this phase, test blanket module electromagnetic and hydraulic tests can take place and give very useful information further iterated system designs to be installed for DT operation.

The actual length of this phase depends on the merit of the ongoing experimentation with regard to the later DT operation, in particular the ability to achieve good H-mode confinement with a suitably high plasma density.

Loading conditions in H-H and D-T Phases

Loading ParametersH-H phase

Design (Typical) ValuesD-T phase

Design (Typical) Values

Operation peak heat flux (MW/m2)0.11 for 600 cycles/y and

1000 cycles for 2.5 y0.27-038

for 3000 cycles/y

Max. FW surface heat flux (MW/m2) 0.3 localized for MARFE0.5 localized for 100

cycles/y

Neutron wall load (MW/m2) - 0.78 (0.78)

Pulse length (sec) Up to 400 400 up to 3000

Duty cycle 0.22 > 0.22

Av. FW neutron fluence (MWa/m2) - 0.1 (first 10 y) up to 0.3

Av. Plasma Availability 0.5% 3.8%

Answers needed on Plasma Operation during H-H phase to plan testing

Will plasma pulses be grouped into operating campaigns, with ITER not operating for extended periods?

Will disruptions and VDEs be intentionally triggered at known times with known, increasing intensity?

Will “actual length” likely be shorter? Or longer?

DCLL H-H Phase Testing Strategy & Overall Goals

DCLL TBM “Prototype” testing Strategy : – Gather a database of operational information via repeated parameter (flowrates,

temperatures, plasma conditions) scans– Operate TBM empty of PbLi until early disruption tests demonstrate TBM and system

integrity.

1. Establish testing capability, performance baseline and operational experience of the TBM and ancillary systems– Integration of control systems and diagnostics with ITER systems– Demonstration of required port integration and remote handling procedures– Measurement of thermal time constants and heat loss, control time constants– Testing heating/filling/draining/remelting and accident response procedures

2. Validate general TBM structure and design – Mechanical response of the TBM structure to transient EM loads– Determine ferromagnetic and MHD flow perturbation of ITER fields– Measure thermal and particle load effects on plasma facing surface (Be) and FW

structure/heat sink3. Perform initial studies of MHD effects and Flow Channel Insert performance

– MHD flow distribution and pressure drop in toroidal field and toroidal + poloidal field– FCI performance changes as a function LM exposure time and loading from EM

events– Map ITER field in TBM area

Strawman Testing Schedule in Early H-H

Integrated Commissioning Phase

(12 mo)

Loops static leak tests and flow tests with blank operation, He and Pb

TBM Installation and connection tests with RH equipment

System static leak tests: PbLi and He systems

He system operational scan (thru He flowrates and inlet T)Control and diagnostic operationThermal time constants (no PbLi), He temp changes, Heat lossTBM temp, stress, vibration with He flow (no PbLi)

1st H-H Plasma Campaigns

(3 mo, 400 pulses – interlaced with downtime)

Repeat He system operational scan

Magnet field measurements in and around TBM

First wall thermal imaging during plasma operation

Response of structure to early disruption, VDEImpulse structural loads, currents (?!)Deformation (strain, visual imaging)He leakage tests (into VV and into empty PbLi channels)

1st H-H Extended Downtimes

(9 mo – interlaced with uptime)

PbLi heating/filling/draining/refilling using Helium preheating and trace heating

He/PbLi system operational scan (thru He and PbLi flowrates and inlet Ts)Control and diagnostic operationThermal time constants (w/ PbLi), He temp changes, Heat lossTBM temp, stress, vibration with He flow, (w/ PbLi)PbLi pressure drop, flow distribution

Strawman Testing Schedule in Late and Post H-H

Later H-H Plasma Campaigns

(5 mo, 600 pulses – interlaced with downtime)

Repeat He/PbLi system operational scans - higher surface heat flux

First wall thermal imaging during plasma operation

Response of structure and FCIs to higher power disruption, VDE with PbLiStructural loads, currents (?!)Deformation (strain, visual imaging)He leakage tests (into VV and into PbLi channels)Changes in PbLi pressure drop or flow distribution

Later H-H Extended Downtimes

(13 mo – interlaced with uptime)

LOCA/LOFA simulations

Repeat He/PbLi system operational scans

FCI stress simulations - high PbLi T and Flowrate, low He temp and dTchanges in pressure drop, flow distributionchanges in temperatures

Post H-H Phase

TBM Removal and disconnection tests with RH equipment

Precise TBM dimensional changes and diagnostic sensor condition

Physical examination of FCIsDimensional changes, crackingLM infiltration

Corrosion/damage of structures and couponsVisible damage, Mass changesMechanical changes in material samples

Proposed transducers needed to meet H-H phase TBM testing goals

Measured Field Proposed Sensor TypeQuantity

(InVes/ExVes)

DC and AC fieldsHall Probes

Flux Loops

6/2

6/0

Forces Load cells at attachment points (possible?) 4/0

StrainOptical strain gauges

Conventional strain gauges

12/0

0/8

Temperature Thermocouples – various locations 20/20

AC Current Embedded Coils 6/0

PbLi flowPotential measurements at various locations

Vortex flow meter

12/2

0/3

PbLi/He pressure Piezo-sensors 24/12

He FlowHot wires

Standard coriolos (or other) industrial flow meters

12/0

0/4

Leaks RGA 2/2

total Too many??? 92/53

Check existing plasma exp and ITER basic machine

Impact on Proposed Testing on Design of H-H Phase DCLL TBM and systems

H-H TBM – should include FCIs and be – based on the fabrication technology and materials proposed for

the D-T phase – both structure and FCIs Number and position of diagnostics must be determined

concomitant with the analysis and design– Feed-through for diagnostic lines in the TBM– Attachments and lead routing inside the TBM

Helium coolant systems and PbLi systems– Variable flowrate (bypasses of VFD) and inlet temperature

control (Heaters and/or HX)– Leak checking system (tracer gas?)– Include diagnostics and control elements

ITER’s Description of ITER D-D phase plasma operation In this phase, neutrons will be produced, and tritium will be produced

from DD reactions. – Part of this tritium will then be burnt in DT reactions. – the activation level inside the vacuum vessel will not allow human access after a

few deuterium discharges with powerful heating. – the capacity of the heat transfer system (except for the divertor and heating

devices) could initially be minimal, and demand for the tritium processing system would be very small.

– later in the phase integrated DT commissioning can take place, with short pulses at high fusion power.

The major achievements would be as follows: – replacement of H by clean D plasma; – confirmation of L-H threshold power and confinement scalings; – establishment of a reference plasma (current, heating power, density,

detached/semi-detached divertor, ELMy H-mode, etc.); – particle control (fuel/ash/impurity/fuelling/pumping); – steady-state operation with full heating power; – finalisation of nuclear commissioning with a limited amount of tritium; – demonstration of high fusion power comparable to the nominal value for the full

DT burn, for a short time. Some information can be provided in this phase on test blanket

neutronics behaviour, allowing optimisation of the designs for later DT operation.

Nuclear Field / Tritium Production (N/T) TBM

Purpose:– database of neutron field measurements for various types of ITER

discharges and conditions– characterize tritium production rate (TPR), and nuclear heating rates. – FW He cooling and tritium implantation– Testing of tritium processing and control equipment – 1st permeation

data

Design: – Similar design and structure as the Integrated-TBM– Rabbit-style tube system for deploying/retrieving activation foils into

several location in the TBM– Deploy tritium processing system– Tritium diagnostics (Li glass, other)– Nuclear heat (micro-calorimeter, other) diagnostics

Testing during DD and early DT phase:– ~2 years in-ITER– Required operational conditions still to be determined

ITER’s Description of ITER Low Duty D-T phase plasma operation

Fusion power and burn pulse length will be gradually increased until the inductive operational goal is reached.

Non-inductive, steady-state operation will also be developed.

Test blanket modules will begin to accumulate results in a situation resembling their operating environment, allowing fine tuning of the designs, and a reference mode of operation for that testing will be established.

Thermofluid / MHD (T/M) TBM Purpose:

– thermal and electrical insulation properties of the FCI with nuclear heating– FCI failure rates and effects with nuclear heating– tritium permeation through FCIs and into helium coolants– Pbli natural convection effects on thermal performance– Initial data on activation products and chemistry control

Design: – Aspects of TBM itself still TBD based on ongoing R&D and scaling

• 1 module or multiple submodules on a strongback?• Multiple channel sizes and FCIs

– Temperature, electric potential diagnostics– Calorimetry– Tritium counting diagnostics

Testing in low duty DT phase – ~2 years in-ITER– moderate temperature (<400-500C) operation of the TBM with PbLi, – possibly with short excursions with PbLi above 500C

Description of ITER High Duty D-T phase plasma operation

Improve overall performance, emphasise – testing of components and materials with higher neutron fluences, – aim for high availability– further improved modes of plasma operation.

The implementation and length of this phase will be depend on the results from the preceding three phases and assessment of the merits and priorities of programmatic proposals.

Whether and when to incorporate tritium breeding during this phase will be decided on the basis of the availability of tritium from external sources, the results of breeder blanket testing, and experience with plasma and machine performance. – such a decision would lead to a non-operating period of about 2 years

while the blanket system is installed in the outboard plasma region, as provided for in the design and initial installation

– the opportunity would undoubtedly also be taken to upgrade ancillary equipment at that time.

Integrated Operation (I) TBM

Purpose: – Explore longer term Integrated operation of the system while thermally

simulating as near as possible DEMO operation conditions including small accumulation of radiation damage in FCIs and RAFS joints

• test improved FCI materials

• corrosion in thermal/chemical environment

• tritium permeation

– Investigate/improve online tritium recovery from PbLi and He streams

– Investigate/improve online PbLi purification systems

Design– Preliminary design outlined in DDD (shown by Wong and Dagher)

Testing in high duty DT phase– ~3 years in-ITER

– Continuous operation in long campaigns looking for changes in performance and failures