FNS related activities in the ARIES program
description
Transcript of FNS related activities in the ARIES program
FNS related activitiesin the ARIES program
M. S. Tillack
Fusion Nuclear Science and Technology Annual Meeting
3 August 2010
UCLA
Topics
1. Design and analysis of plasma-facing components: pushing the limits of performance
2. A new systems analysis technique – VASST
3. Evaluation of R&D pathways using “Technology Readiness”
Several W-He divertor designs have been studied and evolved in ARIES
Tapered T-tube
Fingers without W/FS joints
Pin-fin cooling
Finger/platecombinations
The T-tube divertor is limited by W temperature more than stress, so tapering is helpful
Tapering reduces temperature and increases stress
Pin-fin experiments demonstrated very high heat removal
• Increases qmax to 18 MW/m2 at expected Re, and to 19 MW/m2 at higher Re
• Allows operation at lower Re for a given qmax lower pressure drop
For plate divertor, pin-fin array
5
qm
ax [
MW
/m2
]
Re (/104)
2 mm Bare 2 mm Pins 0.5 mm Bare 0.5 mm Pins
Fabrication to failure (birth to death) analysis
Fabrication Cycle
Operation Cycle,warm shutdown
Time-dependent analysis including fabrication, shakedown, cycling (warm or cold shutdown), off-normal events.
Development of reference scenarios for power plants: ELM’s, disruptions, VDE’s
Pushing the limits: beyond 3Sm
past present
future
ASME code (3Sm)
½ ue
(necking)creep, crack growth,
irradiation effects
Self-relieving thermal stress is a large portion of the total stress in HHF components.
Accounting for inelastic behavior (yielding) can expand the design window considerably.
Our goal is steady-state divertor heat fluxes >10 MW/m2 (>1 MW/m2 for FW’s) and accommodation of transients.
Elastic-plastic analysis allows for higher performance in the first wall
stress-free-temperature is 1050 ºC
Heat Flux
Elastic Analysis Plastic Analysis
Stress inODS FS
MPa
Stress in F82HMPa
Stress in ODS FS
MPa
Stress inF82HMPa
0.0 MW/m2 (390˚C warm
shutdown)1530 1520 483 370
1.0 MW/m2 1140 1500 231 202
2.0 MW/m2 1520 1510 315 271
A modified first wall concept has been examined
W pins are brazed into ODS steel plate, which is attached to ferritic steel cooling channels
Pins help resist thermal transients and erosion
Minor impact on neutronics
1 MW/m2 normal, 2 MW/m2 transient applied
1 mm FW leads to ratchetting, 2 mm is stableF82H
ODS steel
W
F82H
ODS steel
W
External transition joints help alleviate one of the most challenging aspects of HHFC’s
mat’l ε2d εallowable
ODS 0.77% ~1%Ta 0.54% 5-15%W ~0 % ~1%
2D plane strain analysis•elastic/plastic, bilinear isotropic hardening•1050 ºC stress-free (brazing) temperature•100 cycles 20–700˚C•3D analysis will be performed next
Future plans on HHFC (tentative)
Full elastic/plastic analysis of plate and finger concepts
3D analysis of joint
Design integration of pin fins in the plate concept
Come see us at TOFE: “Optimization of ARIES T-tube divertor concept,” J. A. Burke, X. R.
Wang, M. S. Tillack
“Elastic-plastic analysis of the transition joint for high performance divertor target plate,” D. Navaei, X. R. Wang, M. S. Tillack, S. Malang
“High performance divertor target concept for a power plant: a combination of plate and finger concepts,” X.R. Wang, S. Malang, M. S. Tillack
“Innovative first wall concept providing additional armor at high heat flux regions,” X.R. Wang, S. Malang, M. S. Tillack
“Ratchetting models for fusion component design,” J. P. Blanchard, C. J. Martin, M. S. Tillack, and X. R. Wang
“Developing a new visualization tool for the ARIES systems code,” L. C. Carlson, F. Najmabadi, M. S. Tillack
Systems analysis in ARIES has evolved during the past 2 years
ASC VASST
Determination of an optimum design point.
Single-parameter scans around the design point.
Difficult to use, maintain and modify.
Non-interactive tool for self-consistency and costing.
Multi-dimensional parameter space scans.
Large database of physics operating points stored.
Graphical user interface.
Interactive tool for concept exploration.
Fusion Eng. & Design, 85 (2), 243-265, 2010.
Systems analysis flow
Plasmas that satisfy
power and particle balance
Inboard radial build
and engineering
limits
Top and outboard build,
costing
physics engineering build out & costing
Scan several plasma parameters to generate large
database of physics operating points
Screen physics operating points thru
physics filters, engineering
feasibility, and engineering filters
Surviving feasible operating points are built out and costed, graphical display of
parameters (e.g. COE)
1. Toroidal magnetic fields
2. Heat flux to divertor3. Neutron wall load4. Net electric power
Filters include e.g.
VASST GUI v.2
Pull-down menus for common parameters
Blanket database used
Number of points in database
Constraint parameter can restrict database
Auto-labeling
Correlation coefficient
Save plot as TIFF, JPEG, BMP, PNG…
Color bar scale
Edit plotting properties
(Visual ARIES Systems Scanning Tool)
Turn on ARIES-AT point design for reference
The code has multiple applicationsscan parameter space, explore tradeoffs (e.g. conservative vs.
aggressive), describe a design point, and even evaluate research facilities
We are currently exploring the 4 corners of tokamak design space to better understand
tradeoffs
Example: optimum size and N in the aggressive/aggressive
corner
We adopted “technology readiness levels” as the basis for the evaluation
of progress
TRL Generic Description (defense acquisitions definitions) Facilities (mst)
1 Basic principles observed and formulated.
2 Technology concepts and/or applications formulated.
3 Analytical and exptl demo of critical function and/or proof of concept. Single effects
4 Component and/or bench-scale validation in a laboratory environment. Multiple effects
5 Component and/or breadboard validation in a relevant environment. Multiple effects
6 System/subsystem model or prototype demo in relevant environment. FNSF (PoP)
7 System prototype demonstration in an operational environment. Prototype (ITER?)
8 Actual system completed and qualified through test and demonstration. Demo reactor
9 Actual system proven through successful mission operations. Power plant
TRL’s express increasing levels of integration and environmental relevance, terms which must be
defined for each technology application
Fusion Science & Tech 56 (2) August 2009.
Utility Advisory Committee“Criteria for practical fusion power
systems”
Have an economically competitive life-cycle cost of electricity
Gain public acceptance by having excellent safety and environmental characteristics No disturbance of public’s day-to-day activities No local or global atmospheric impact No need for evacuation plan No high-level waste Ease of licensing
Operate as a reliable, available, and stable electrical power source Have operational reliability and high availability Closed, on-site fuel cycle High fuel availability Capable of partial load operation Available in a range of unit sizes
J. Fusion Energy 13 (2/3) 1994.
These criteria for practical fusion suggest three categories of technology readiness
A. Power management for economic fusion energy1. Plasma power distribution2. Heat and particle flux management3. High temperature operation and power conversion4. Power core fabrication5. Power core lifetime
B. Safety and environmental attractiveness6. Tritium control and confinement7. Activation product control and confinement8. Radioactive waste management
C. Reliable and stable plant operations9. Plasma control10.Plant integrated control11.Fuel cycle control12.Maintenance
• LWR spent fuel processing• Waste form development• Fast reactor spent fuel
processing • Fuel fabrication • Fuel performance
cf. GNEP issues:
Example TRL table: Heat & particle flux handling
Issue-Specific Description Program Elements
1System studies to define tradeoffs and requirements on heat flux level, particle flux level, effects on PFC's (temperature, mass transfer).
Design studies, basic research
2PFC concepts including armor and cooling configuration explored. Critical parameters characterized.
Code development, applied research
3Data from coupon-scale heat and particle flux experiments; modeling of governing heat and mass transfer processes as demonstration of function of PFC concept.
Small-scale facilities:e.g., e-beam and plasma simulators
4Bench-scale validation of PFC concept through submodule testing in lab environment simulating heat fluxes or particle fluxes at prototypical levels over long times.
Larger-scale facilities for submodule testing, High-temperature + all expected range of conditions
5Integrated module testing of the PFC concept in an environment simulating the integration of heat fluxes and particle fluxes at prototypical levels over long times.
Integrated large facility:Prototypical plasma particle flux+heat flux (e.g. an upgraded DIII-D/JET?)
6Integrated testing of the PFC concept subsystem in an environment simulating the integration of heat fluxes and particle fluxes at prototypical levels over long times.
Integrated large test facility with prototypical plasma particle and heat flux
7 Prototypic PFC system demonstration in a fusion machine.Fusion machineITER (w/ prototypic divertor), CTF
8Actual PFC system demonstration & qualification in a fusion reactor over long operating times.
Demo
9Actual PFC system operation to end-of-life in fusion reactor with prototypical conditions and all interfacing subsystems.
Commercial power plant
The level of readiness depends on the design concept
Issue-Specific Description Program Elements
1System studies to define tradeoffs and requirements on heat flux level, particle flux level, effects on PFC's (temperature, mass transfer).
Design studies, basic research
2PFC concepts including armor and cooling configuration explored. Critical parameters characterized.
Code development, applied research
3Data from coupon-scale heat and particle flux experiments; modeling of governing heat and mass transfer processes as demonstration of function of PFC concept.
Small-scale facilities:e.g., e-beam and plasma simulators
4Bench-scale validation of PFC concept through submodule testing in lab environment simulating heat fluxes or particle fluxes at prototypical levels over long times.
Larger-scale facilities for submodule testing, High-temperature + all expected range of conditions
5Integrated module testing of the PFC concept in an environment simulating the integration of heat fluxes and particle fluxes at prototypical levels over long times.
Integrated large facility:Prototypical plasma particle flux+heat flux (e.g. an upgraded DIII-D/JET?)
6Integrated testing of the PFC concept subsystem in an environment simulating the integration of heat fluxes and particle fluxes at prototypical levels over long times.
Integrated large test facility with prototypical plasma particle and heat flux
7 Prototypic PFC system demonstration in a fusion machine.Fusion machineITER (w/ prototypic divertor), CTF
8Actual PFC system demonstration & qualification in a fusion reactor over long operating times.
Demo
9Actual PFC system operation to end-of-life in fusion reactor with prototypical conditions and all interfacing subsystems.
Commercial power plant
Power plant relevant high-temperature gas-cooled PFC’s
Low-temperature water-cooled PFC’s
The current status was evaluated for a reference ARIES power plant
Level completed Level in progress
TRL
1 2 3 4 5 6 7 8 9
Power management
Plasma power distribution
Heat and particle flux handling
High temperature and power conversion
Power core fabrication
Power core lifetime
Safety and environment
Tritium control and confinement
Activation product control
Radioactive waste management
Reliable/stable plant operations
Plasma control
Plant integrated control
Fuel cycle control
Maintenance
For the sake of illustration, we considered a Demo based on the ARIES advanced tokamak DCLL power plant design concept
He-cooled W divertor, DCLL blanket @700˚C, Brayton cycle, plant availability=70%, 3-4 FPY in-vessel, waste recycling or clearance
The ITER program contributes in some areas, very little in others
TRL
1 2 3 4 5 6 7 8 9
Power management
Plasma power distribution
Heat and particle flux handling
High temperature and power conversion
Power core fabrication
Power core lifetime
Safety and environment
Tritium control and confinement
Activation product control
Radioactive waste management
Reliable/stable plant operations
Plasma control
Plant integrated control
Fuel cycle control
Maintenance
ITER promotes to level 6 issues related to plasma and safety
ITER helps incrementally with some issues, such as blankets, PMI, fuel cycle
The absence of reactor-relevant technologies severely limits its contribution in several areas
Major gaps remain for several of the key issues for practical fusion energy
TRL
1 2 3 4 5 6 7 8 9
Power management
Plasma power distribution
Heat and particle flux handling
High temperature and power conversion
Power core fabrication
Power core lifetime
Safety and environment
Tritium control and confinement
Activation product control
Radioactive waste management
Reliable/stable plant operations
Plasma control
Plant integrated control
Fuel cycle control
Maintenance
A range of nuclear and non-nuclear facilities are required to advance from the current status to TRL6
One or more test facilities such as CTF are required before Demo to verify performance in an operating environment