Dynamic retention and deposition in NSTX measured with quartz microbalances

Similar documents
Recent results in dynamic retention and dust

D. Smith, R. Fonck, G. McKee, D. Thompson, and I. Uzun-Kaymak

NSTX. Electron gyro-scale fluctuations in NSTX plasmas. David Smith, PPPL. Supported by

Supported by. Role of plasma edge in global stability and control*

David R. Smith UW-Madison

Supported by. Relation of pedestal stability regime to the behavior of ELM heat flux footprints in NSTX-U and DIII-D. J-W. Ahn 1

Supported by. Suppression of TAE and GAE with HHFW heating

Supported by. Validation of a new fast ion transport model for TRANSP. M. Podestà - PPPL

Supported by. The dependence of H-mode energy confinement and transport on collisionality in NSTX

NSTX Results and Plans toward 10-MA CTF

Alfvén Cascade modes at high β in NSTX*

Supported by. The drift kinetic and rotational effects on determining and predicting the macroscopic MHD instability

M. Podestà, M. Gorelenkova, D. S. Darrow, E. D. Fredrickson, S. P. Gerhardt, W. W. Heidbrink, R. B. White and the NSTX-U Research Team

Investigations of pedestal turbulence and ELM bursts in NSTX H-mode plasmas

UCLA POSTECH UCSD ASIPP U

EXD/P3-13. Dependences of the divertor and midplane heat flux widths in NSTX

Carbon Deposition and Deuterium Inventory in ASDEX Upgrade

THE ADVANCED TOKAMAK DIVERTOR

L-to-H power threshold comparisons between NBI and RF heated plasmas in NSTX

Recent Development of LHD Experiment. O.Motojima for the LHD team National Institute for Fusion Science

Divertor Heat Flux Reduction and Detachment in NSTX

EU Plasma-Wall Interactions Task Force

STEADY-STATE EXHAUST OF HELIUM ASH IN THE W-SHAPED DIVERTOR OF JT-60U

Tokamak Divertor System Concept and the Design for ITER. Chris Stoafer April 14, 2011

NSTX Plasma-Material Interface (PMI) Probe and supporting experiments

In-vessel Tritium Inventory in ITER Evaluated by Deuterium Retention of Carbon Dust

ITER A/M/PMI Data Requirements and Management Strategy

1999 RESEARCH SUMMARY

Effect of the Surface Temperature on Net Carbon Deposition and Deuterium Co-deposition in the DIII-D Divertor

Dust Studies in Fusion Devices

Deuterium and Hydrogen Retention Properties of the JT-60 and JT-60U Divertor Tiles

Power Deposition Measurements in Deuterium and Helium Discharges in JET MKIIGB Divertor by IR-Thermography

A Faster Way to Fusion

Fusion Nuclear Science Facility (FNSF) Divertor Plans and Research Options

ASSESSMENT AND MODELING OF INDUCTIVE AND NON-INDUCTIVE SCENARIOS FOR ITER

Thomas Schwarz-Selinger. Max-Planck-Institut for Plasmaphysics, Garching Material Science Division Reactive Plasma Processes

Effect of non-axisymmetric magnetic perturbations on divertor heat and particle flux profiles

Improved Plasma Confinement by Ion Bernstein Waves (IBWs) Interacting with Ions in JET (Joint European Torus)

Radiative type-iii ELMy H-mode in all-tungsten ASDEX Upgrade

Thin Faraday foil collectors as a lost ion diagnostic

ICRF Minority-Heated Fast-Ion Distributions on the Alcator C-Mod: Experiment and Simulation

Diagnostics for Burning Plasma Physics Studies: A Status Report.

1 EX/3-5. Material Erosion and Redeposition during the JET MkIIGB-SRP Divertor Campaign

Relevant spatial and time scale in tokamaks. F. Bombarda ENEA-Frascati, FSN-FUSPHY-SAD

Materials for Future Fusion Reactors under Severe Stationary and Transient Thermal Loads

STUDY OF ADVANCED TOKAMAK PERFORMANCE USING THE INTERNATIONAL TOKAMAK PHYSICS ACTIVITY DATABASE

In-Situ Tritium Measurements Of The Tokamak Fusion Test Reactor Bumper Limiter Tiles Post D-T Operations

Thermographic measurements of power loads to plasma facing components at Wendelstein 7-X

ANALYSIS OF PLASMA FACING MATERIALS IN CONTROLLED FUSION DEVICES. Marek Rubel

Plasma-Wall Interaction Studies in the Full-W ASDEX Upgrade during Helium Plasma Discharges

Critical Gaps between Tokamak Physics and Nuclear Science. Clement P.C. Wong General Atomics

1 EX/P4-8. Hydrogen Concentration of Co-deposited Carbon Films Produced in the Vicinity of Local Island Divertor in Large Helical Device

Temporal Evolution and Spatial Distribution of Dust Creation Events in Tore Supra and in ASDEX Upgrade Studied by CCD Image Analysis

Heating and Confinement Study of Globus-M Low Aspect Ratio Plasma

Scaling of divertor heat flux profile widths in DIII-D

Charge-exchange neutral hydrogen measurements in TFTR using Pd-MOS microsensors

The Spherical Tokamak as a Compact Fusion Reactor Concept

and expectations for the future

Characterization of heated dust particles using infrared and dynamic images in LHD

Divertor Power Handling Assessment for Baseline Scenario Operation in JET in Preparation for the ILW

Direct measurements of the plasma potential in ELMy H mode. plasma with ball pen probes on ASDEX Upgrade tokamak

The Path to Fusion Energy creating a star on earth. S. Prager Princeton Plasma Physics Laboratory

Tungsten: An option for divertor and main chamber PFCs in future fusion devices

Spherical Torus Fusion Contributions and Game-Changing Issues

INTRODUCTION TO MAGNETIC NUCLEAR FUSION

Plasma-wall interaction studies in the full-w ASDEX Upgrade during helium plasma discharges

Physical Design of MW-class Steady-state Spherical Tokamak, QUEST

Current Drive Experiments in the HIT-II Spherical Tokamak

Overview of Tokamak Rotation and Momentum Transport Phenomenology and Motivations

Scaling of divertor heat flux profile widths in DIII-D

Power Exhaust on JET: An Overview of Dedicated Experiments

Atomic physics in fusion development

Neutral beam plasma heating

Bolometry. H. Kroegler Assciazione Euratom-ENEA sulla Fusione, Frascati (Italy)

Introduction to Fusion Physics

Exhaust scenarios. Alberto Loarte. Plasma Operation Directorate ITER Organization. Route de Vinon sur Verdon, St Paul lez Durance, France

Introduction to Nuclear Fusion. Prof. Dr. Yong-Su Na

ADVANCES IN PREDICTIVE THERMO-MECHANICAL MODELLING FOR THE JET DIVERTOR EXPERIMENTAL INTERPRETATION, IMPROVED PROTECTION, AND RELIABLE OPERATION

Prospects of Nuclear Fusion Energy Research in Lebanon and the Middle-East

Latest Results from the Globus-M Spherical Tokamak

Local organizer. National Centralized Tokamak

Fast ion physics in the C-2U advanced, beam-driven FRC

Electrode and Limiter Biasing Experiments on the Tokamak ISTTOK

Helium effects on Tungsten surface morphology and Deuterium retention

Physics of fusion power. Lecture 14: Anomalous transport / ITER

TOKAMAK EXPERIMENTS - Summary -

Overview of Pilot Plant Studies

(Inductive tokamak plasma initial start-up)

II: The role of hydrogen chemistry in present experiments and in ITER edge plasmas. D. Reiter

Plasma Wall Interactions in Tokamak

Recent results from lower hybrid current drive experiments on Alcator C-Mod

Ion beam analysis methods in the studies of plasma facing materials in controlled fusion devices

Global migration of impurities in tokamaks: what have we learnt?

PHYSICAL VAPOR DEPOSITION OF THIN FILMS

Study and Optimization of Boronization in Alcator C- Mod using the Surface Science Station (S 3 )

Non-Solenoidal Plasma Startup in

Improved Plasma Confinement by Ion Bernstein Waves (IBWs) Interacting with Ions in JET

Deuterium Inventory in Tore Supra : reconciling particle balance and post mortem analysis

Impact of Neon Injection on Electron Density Peaking in JET Hybrid Plasmas

Long Term Reduction of Divertor Carbon Sources in DIII-D

Transcription:

Supported by Office of Science Dynamic retention and deposition in NSTX measured with quartz microbalances College W&M Colorado Sch Mines Columbia U Comp-X General Atomics INEL Johns Hopkins U LANL LLNL Lodestar MIT Nova Photonics New York U Old Dominion U ORNL PPPL PSI Princeton U SNL Think Tank, Inc. UC Davis UC Irvine UCLA UCSD U Colorado U Maryland U Rochester U Washington U Wisconsin Charles Skinner, Henry Kugel, Lane Roquemore, PPPL Rajesh Maingi, ORNL, William Wampler, SNL Volker Rohde, Asdex-U 7th Div/SOL ITPA Meeting Nov 6-9th, 2006 Toronto, Canada. Culham Sci Ctr U St. Andrews York U Chubu U Fukui U Hiroshima U Hyogo U Kyoto U Kyushu U Kyushu Tokai U NIFS Niigata U U Tokyo JAERI Hebrew U Ioffe Inst RRC Kurchatov Inst TRINITI KBSI KAIST ENEA, Frascati CEA, Cadarache IPP, Jülich IPP, Garching ASCR, Czech Rep U Quebec 1

Motivation: ITER vacuum vessel Tritium retention and removal is a high risk and high consequence issue for ITER. predictive understanding lacking Conventional Wall Diagnostics: 1. Tile / coupon samples Well defined spatial location Time integrated archeology - no correlation with plasma events 2. Gas balance (fueling & exhaust) Well defined time resolution No spatial information These typically give widely different values for retention (TFTR a notable exception). ITER Quartz Microbalances (QMBs) offer both space and time information. limitations: cannot be used in high heat flux area. relatively sparse coverage due to cost 2

Quartz Microbalances in NSTX Quartz Crystal Thermocouple Deposition changes quartz crystal oscillation frequency. film mass crystal mass Frequency measured by pulse accumulator controlled by 20 MHz reference oscillator. Resolution < one monolayer = Infincon XTM/2 frequency change bare crystal frequency Mass consistent with independent NRA measurements 3

QMB temperature compensation: Frequency Change (Hz) -100-110 -120-130 -140 20040813TempCal Temperature Calibration y = 56.224-5.199x R= 0.99991-150 30 32 34 36 38 40 Degrees C "Angstrom" after temperature compensation Temperature Equilibration Time ~ 1-2 mins 160 120 80 40 0 V test -2-1 0 1 2 3 4 5 6 HEAT time (minutes) QMB frequency is temperature dependent thermcouple monitors qmb temperature. temperature calibration derived from weekend temperature drift: 1 degree C increase is equivalent to: 3.7 Å deposition. temperature induced frequency changes are subtracted from data Transient temperature differences between qmb crystal and thermocouple possible. Soldering iron applied to qmb housing for 20 s. Frequency changes with temperature. Temperature compensated thickness recovers after 1-2 mins. 4

Overview: a complex pattern of material gain and loss Deposition thickness at midnight during campaign Upper QMB Lower QMB Boron 23 April 2005 25 August 2005 Correlation of thickness with boronization apparent zoom in to discharges >> 5

Four similar discharges have very different mass change: QMB mass change Staircase of discharge-by-discharge erosion or deposition not observed. Transient rise seen followed by decay decay longer than thermal relaxation time QMB asymptotic level can be higher, lower or the same as before discharge. First shot of day always shows a step-up in asymptotic level. 6

Mass changes over 24 hours: General Features: Discharges show transient rise followed by decay. Large step-up in mass on 1st shot of day Effect independent of: prior He glow discharge, discharge type: Ohmic (115476) or NBI, He or D Some discharges show step-up or step-down in asymptotic level. Slow mass loss at end of day. Behavior does not fit simple deposition/erosion picture Y-axis zero point arbitrary µg/cm 2 (relative) µg/cm 2 (relative) µg/cm 2 (relative) 2.0 1.6 1.3 1.0 0 5 10 15 20 1.0 0.8 0.5 0.3 He fueled D fueled 0.0 0 1.4 5 10 15 20 1.0 0.7 0.4 first four shots similar No initial GDC upper lower GDC 20050510 20050823 0.0 0 5 10 15 20 20060309 Time of day 7

Behavior can be explained by dynamic retention of D Overnight outgassing desaturates wall. D implanted on 1st shot adds mass, saturates wall. Outgassing between shots too slow to desaturate wall. Mass gain on subsequent discharges is much less. Look at numbers: 5e21 D atoms fuel each discharge Interior surface area of NSTX = 40.66 m 2 = 4e21 Å 2 Fueling equivalent to about one D atom per Å 2 or 0.04!g/cm 2 if smeared over NSTX interior. Initial mass gain of QMBs is " 0.35!g/cm 2 i.e. 10 x higher. Outgassing rate (atoms/s) 1e20 1e18 1e16 1e14 100 1e4 1e6 1e8 Time (s) Model of outgassing of H in traps and solution in carbon [Andrew & Pick JNM 220-222 (1995) 601] Strong dependence of outgassing rate on H concentration CONCLUDE: Dynamic retention concentrated in plasma shadowed regions. 8

End of day behavior supports dynamic retention: 0.5 0.4 upper Mass loss continues for ~ > 1 h after last discharge Outgassing of deuterium evident in ion gauge and rga data. µg/cm 2 0.3 0.2 0.1 mid lower Compare material loss to exhaust pumpout: Average mass loss from 17h - 18 h is 0.11!g/cm 2 RGA shows exhaust mostly D (>70%) Mass of deuterium pumped by turbopumps from 17h - 18h is ~ 984!g 1 h exhaust is equivalent to mass loss of 0.1!g/cm 2 over 1 m 2 Suggests dynamic retention occurs over only a small fraction of the NSTX vessel area (consistent with gas uptake on first shot of day) torr 20050614 0.0 17 17.5 18 18.5 19 Time of Day (h) 1.5 10-6 Vessel pressure 1 10-6 5 10-7 0 17 17.5 18 18.5 19 Time of Day RGA 4 (D2) 28 (CO) 18 (H2O) 44 (CO2) 12, 14, 40 9

Erosion / deposition apparent over long term Erosion/deposition is small compared to dynamic retention Ion beam analysis of quartz crystals after 2005 run: (Wampler, SNL) H bottom H top change April 23 - Sept 13 2005-0.3 2.6 µg/cm2 change due to boronization 4.2 2.1 µg/cm2 plasma only -4.5 0.5 µg/cm2 average rate -0.0028 0.0003 µg/cm2/shot average rate -0.0051 0.0006 µg/cm2/sec average rate -0.32 0.03 Å/s Oxygen and protium not measured. For 2004 crystals surface oxygen measured by XPS was 13% of carbon density. TMB = B(CD 3 ) 3 used for boronization, has different stochiometry to deposits: 9% B, 27% C, 81% D. (D removed by He GDC) 10 17 B/cm 2 10 17 D/cm 2 10 17 C/cm 2 QMB Bay H top 0.35 1.13 5.16 QMB Bay I mid 6.81 11.67 22.40 QMB Bay H bot 0.16 0.37 1.13 Nuclear reactions used: 12 C( 3 He,p) 14 N, 11 B( 3 He,p) 13 C, 2 H( 3 He,p) 4 He 10

Volker Rohde et al. QMB 11

12

Conclusions: QMB s offer time and space resolved data on wall coatings. Dynamic retention observed in mass gain after 1st shot-of-day and transient material loss after discharges. also seen on Asdex-U Step up or step down in asymptotic level observed, depending on plasma shape including shape of prior discharges (memory effect), plasma energy and duration and surface temperature. Quantative comparison indicates mass gain / loss at measured rate over ~10% of the vessel area can account for D fueled and pumped. Long term erosion / deposition small compared to dynamic retention at these locations. 13

Extras: 14

Mass gain/loss depends on plasma shape: Transient at discharge time due to dynamic retention (after 1-2 mins). Layer mass asymptotes to level that can be the same or higher/lower than before discharge. Interaction at upper divertor increases in double null discharges. change in SOL flows? - (Fundamenski) Or surface temperature effect? µg/cm 2 2.5 2 1.5 1 0.5 0 lower single null 8 10 12 14 16 20050727 18 Time of Day (h) double null QMBs 15

Some correlation of asymptote with stored energy Contrasting plasmas: Morning: high performance NBI LSN high triangularity; Afternoon: LSN ohmic low triangularity Note change in asymptote with change in triangularity - strike points moved to fresh divertor surface Also: Tile surface temperature rises with discharge. IR camera data shows at Bay H bottom QMB: Rise = 5 C for short pulse ohmic Rise = 30 C for long pulse NBI Change in asymptote with duration and stored energy 16

Quartz Crystal Microbalance: Deposition changes crystal oscillating frequency. Widely used for process control during vacuum deposition. Used in TdeV, ASDEX, JET TEXTOR and NSTX tokamaks. Frequency measured by pulse accumulator controlled by 20 MHz reference oscillator. Commercially available system is relatively fast (1/4 s), precise, able to measure heavily loaded crystals and has immunity from mode hopping. Crystal!wobbles" in thickness shear mode Crystal frequency spectrum Inficon film mass crystal mass = frequency change bare crystal frequency more complex relation available that takes account of acoustic properties of film and is valid up for to 40% frequency shifts. Calculated sensitivity: -81 Hz / µg /cm 2 or -13 Hz /n.m (for film density of 1.6 g/cm 3 ) Caveat: frequency also sensitive to temperature, electronics for ITER environment may need development. Inficon 5.9 MHz fundamental resonance constantly identified by zero phase difference between applied signal voltage and current passing through crystal. 17

The National Spherical Torus Experiment (NSTX) research program is aimed at: Exploring the physics of high beta and high confinement in a low aspect ratio device Demonstrate non-inductive current generation and sustainment. plasma major radius: 0.85 m, minor radius: 0.68 m, toroidal field of up to 0.55 T, plasma current up to 1.5 MA pulse duration up to 1.5 s. 7 MW neutral beam injection 6 MW of high harmonic fast wave RF at 30 MHz. Plasma facing components that are in contact with the plasma are protected by a combination of graphite and CFC tiles. 18