NSTX-U
Download
1 / 28

NSTX-U Status and Plan - PowerPoint PPT Presentation


  • 80 Views
  • Uploaded on

NSTX-U. Supported by. NSTX-U Status and Plan. 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 Tsukuba U U Tokyo JAEA Hebrew U Ioffe Inst RRC Kurchatov Inst TRINITI NFRI KAIST POSTECH Seoul National U

loader
I am the owner, or an agent authorized to act on behalf of the owner, of the copyrighted work described.
capcha
Download Presentation

PowerPoint Slideshow about ' NSTX-U Status and Plan' - nishi


An Image/Link below is provided (as is) to download presentation

Download Policy: Content on the Website is provided to you AS IS for your information and personal use and may not be sold / licensed / shared on other websites without getting consent from its author.While downloading, if for some reason you are not able to download a presentation, the publisher may have deleted the file from their server.


- - - - - - - - - - - - - - - - - - - - - - - - - - E N D - - - - - - - - - - - - - - - - - - - - - - - - - -
Presentation Transcript

NSTX-U

Supported by

NSTX-U Status and Plan

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

Tsukuba U

U Tokyo

JAEA

Hebrew U

Ioffe Inst

RRC Kurchatov Inst

TRINITI

NFRI

KAIST

POSTECH

Seoul National U

ASIPP

ENEA, Frascati

CEA, Cadarache

IPP, Jülich

IPP, Garching

ASCR, Czech Rep

Columbia U

CompX

General Atomics

FIU

INL

Johns Hopkins U

LANL

LLNL

Lodestar

MIT

Nova Photonics

New York U

ORNL

PPPL

Princeton U

Purdue U

SNL

Think Tank, Inc.

UC Davis

UC Irvine

UCLA

UCSD

U Colorado

U Illinois

U Maryland

U Rochester

U Tennessee

U Washington

U Wisconsin

Masayuki Ono

NSTX-U Project Director

PPPL, Princeton University

In collaboration with the NSTX-U Team

The First A3 Foresight Workshop on Spherical Torus (ST) 

Jan. 14-16, 2013

SNU, Seoul, Korea


Talk Outline

  • NSTX-U Mission

  • NSTX Experimental Overview

  • NSTX-U Construction Status

  • NSTX-U Experimental Plan

  • Summary


NSTX-U Mission Elements

Fusion applications of low-A spherical tokamak (ST)

  • Develop plasma-material-interface (PMI) solutions for next-steps

    • Exploit high divertor heat flux from lower-A/smaller major radius

  • Fusion Nuclear Science/Component Test Facility (FNSF/CTF)

    • Exploit high neutron wall loading for material and component development

    • Utilize modular configuration of ST for improved accessibility, maintenance

  • Extend toroidal confinement physics predictive capability

    • Access strong shaping, high b, vfast / vAlfvén, and rotation, to test physics models for ITER and next-steps (see NSTX, MAST, other ST presentations)

  • Long-term: reduced-mass/waste low-A superconducting Demo


NSTX Upgrade will access next factor of two increase in performance to bridge gaps to next-step STs

Low-A

Power Plants

ARIES-ST (A=1.6)

JUST (A=1.8)

VECTOR (A=2.3)

  • * Includes 4MW of high-harmonic fast-wave (HHFW) heating power

Key issues to resolve for next-step STs

Confinement scaling (electron transport)

Non-inductive ramp-up and sustainment

Divertor solutions for mitigating high heat flux

Radiation-tolerant magnets (for Cu TF STs)


NSTX Upgrade will address critical plasma confinement and sustainment questions by exploiting 2 new capabilities

n

/

n

e

Greenwald

Previous

center-stack

New

center-stack

ST-FNSF

constant q, b, r*

2x higher BT and IP increases T, reduces n* toward ST-FNSF to better understand confinement

Provides 5x longer pulses for profile equilibration, NBI ramp-up

NSTX Upgrade

?

ITER-like scaling

Normalized e-collisionality ne*  ne /Te2

TF OD = 40cm

TF OD = 20cm

IP=0.95MA, H98y2=1.2, bN=5, bT = 10%

BT = 1T, PNBI = 10MW, PRF = 4MW

0.95

0.72

2x higher CD efficiency from larger tangency radius RTAN

100% non-inductive CD with q(r) profile controllable by:

tangency radius, density, position

RTAN [cm]

__________________

50, 60, 70, 130

60, 70,120,130

70,110,120,130

New 2nd NBI

Present NBI

J. Menard, et al., Nucl. Fusion 52 (2012) 083015


A schematic of the new center-stack and the TF joint area sustainment questions by exploiting 2 new capabilities

New TF-Flex-Bus Designed and Tested to Full Cycles

TF cooling lines

TF flex-bus

TF Coil

CHI bus

PF Coil 1c

PF Coil 1b

CS Casing

PF Coil 1a

OH Coil


The NSTX-U Inner TF Bundle Manufacturing Stages sustainment questions by exploiting 2 new capabilities

New Zn-Cl-Free Soldering Technique Developed


NSTX-U Support Structural Upgrades sustainment questions by exploiting 2 new capabilities

4x Electromagnetic Forces


Relocation of the 2 sustainment questions by exploiting 2 new capabilitiesnd NBI beam line box from the TFTR test cell into the NSTX-U Test Cell.


2 sustainment questions by exploiting 2 new capabilitiesnd NBI alignment performed and confirmed


Beam-line Component Refurbishment sustainment questions by exploiting 2 new capabilities

Ion Dump

Calorimeter upgrade

Bending Magnet

  • 11


JK cap tack welded to the vacuum vessel after completing alignments, and full welding is now underway (Jan. 3, 2013)


NBI Duct and Torus Vacuum Pumping System (TVPS) components being procured and fabricated

Rectangular bellows

Exit spool piece

40” Torus Isolation (Gate) Valve received

Spool section & supports

TVPS valves, hardware, TMPs, and shields

Circular bellows


NSTX In-Vessel View and CHI Gap Protection Enhancement being procured and fabricated

Expect x 10 Higher Heat Load Into the CHI Gap

CHI Gap

Center Stack

Secondary Passive Plates

PF 1C

PF 1C

NBI Armor

HHFW

Antenna

CHI Gap

Primary Passive Plates

CHI Gap

In-board

Divertor

Out-board

Divertor


Non-inductive ramp-up from ~0.4MA to ~1MA projected to be possible with new centerstack (CS) + more tangential 2nd NBI

  • New CS provides higher TF (improves stability), 3-5s needed for J(r) equilibration

  • More tangential injection provides 3-4x higher CD at low IP:

    • 2x higher absorption (4080%) at low IP = 0.4MA

    • 1.5-2x higher current drive efficiency

TSC simulation of non-inductive ramp-up from IP = 0.1MA, Te=0.5keV target at BT=1T

More tangential 2nd NBI

Present NBI


NSTX-U CHI Start-up Configurations possible with new centerstack (CS) + more tangential 2

X 2 Higher CHI Driven Currents Expected


NSTX-U ECH/EBW System for Non-Inductive Start-Up and Sustainment

28 GHz – 1MW Gyrotron by

U. of Tsukuba

A schematic of the NSTX-U ECH/EBW launcher


Stability control improvements significantly reduce unstable RWMs at low li and high bN; improved stability at high bN/li

Unstable RWM

Stable / controlled RWM

Resonant Field Amplification (RFA) vs. bN/li

  • Disruption probability reduced by a factor of 3 on controlled experiments

    • Reached 2 times computed n = 1 no-wall limit of bN/li = 6.7

  • Lower probability of unstable RWMs at high bN/li

unstable

mode

  • Mode stability directly measured in experiments using MHD spectroscopy

    • Stability decreases up to bN/li = 10

    • Stability increasesat higher bN/li

    • Presently analysis indicates consistency with kinetic resonance stabilization

S.A. Sabbagh

J. Berkery IAEA


Disruptivity studies and warning analysis of NSTX database are being conducted for disruption avoidance in NSTX-U

Disruptivity

Warning Algorithms

bN

q*

li

All discharges since 2006

  • Physics results

    • Low disruptivity at relatively high bN ~ 6; bN / bNno-wall(n=1) ~ 1.3-1.5

      • Consistent with specific disruption control experiments, RFA analysis

    • Strong disruptivity increase for q* < 2.5

    • Strong disruptivity increase for very low rotation

  • Results

    • ~ 98% disruptions flagged with at least 10ms warning, ~ 6% false positives

    • False positive count dominated by near-disruptive events

S. Gerhardt IAEA

  • Disruption warning algorithm shows high probability of success

    • Based on combinations of single threshold based tests


NSTX are being conducted for disruption avoidance in NSTX-U“Snowflake” Divertor Configuration resulted in significant divertor heat flux reduction and impurity screening

Higher flux expansion (increased div wetted area)

Higher divertor volume (increased div. losses)

  • Maintained stable “snowflake” configuration for 100-600 ms with three PF coils

  • Maintained H-mode confinement with core carbon reduction by 50 %

  • NSTX-U control coils will enable improved and up-down symmetric snowflake configurations

V. Soukhanovskii, NF 2009


Lithium Improved H-mode Performance in NSTX are being conducted for disruption avoidance in NSTX-U

Te Broadens, tE Increases, PH Reduces, ELMs Stabilize

Te broadening with lithium

No lithium (129239);260mg lithium (129245)

With Lithium

Without Lithium

H. W. Kugel, PoP 2008

tE improves with lower collisionality

tE improves with lithium

Pre-discharge lithium evaporation (mg)

S. Kaye, IAEA (2012)

R. Maingi, PRL (2011)


Li core concentration stays well below 0.1% for LLD temperature range of 90°C to 290°C

R=135-140 cm, t=500-600 ms

  • Li core concentration remained very low ≤ 0.05%. C remains dominant impurity even after massive (hundreds of milligrams) Li evaporation

  • No apparent increase in Li nor C core concentration even at higher LLD surface temperature.

  • Liquid

    Solid

    M. Podesta, IAEA (2012)

    Reason for low lithium core dilution?:

    • Li is readily ionized ~ 6 eV

    • Li is low recycling – sticks to wall

    • Li has high neoclassical diffusivity

    F. Scotti, APS (2012)


    Clear reduction in NSTX divertor surface temperature and heat flux with increased lithium evaporation

    • a)

    • b)

    • Lithiated graphite

    • c)

    • d)

    T. Gray. IAEA 2012

    • 2 identical shots (No ELMs)

      • Ip = 0.8 MA, Pnbi ~ 4 MW

      • high δ, fexp ~ 20

    • 2, pre-discharge lithium depositions

      • 150 mg: 141255

      • 300 mg: 138240

    • Tsurf at the outer strike point stays below 400° C for 300 mg of Li

      • Peaks around 800° C for 150 mg

    • Results in a heat flux that never peaks above 3 MW/m2 with heavy lithium evaporation


    Radiative Liquid Lithium Divertor Proposed heat flux with increased lithium evaporation

    Based largely on the NSTX Liquid Lithium Divertor Research

    Divertor Heat and Particles Flux

    Edge Plasma

    B0

    000000000000

    Liquid Lithium (LL)

    ~ 1 l/sec

    RLLD

    Core Reacting

    Plasma

    First Wall / Blanket

    At 500°C – 700°C

    Particle pumping by Li coated wall

    Flowing LL Particle Pumping Surfaces

    Li Radiative Mantle

    Li wall coating /

    condensation

    Scrape Off Layer

    Li+++

    Li path

    Li++

    Closed RLLD

    Li Evap. /

    Ionization

    Reduced Divertor Heat and Particle Flux

    Flowing LLD Tray

    200 – 450 °C

    Li+

    Heat Exchanger

    LL In

    LL In

    LL Out

    Divertor Strike Point

    Li0

    LL Purification System to remove tritium, impurities, and dust

    M. Ono. IAEA 2012


    Design studies focusing on thin, capillary-restrained liquid metal layers

    Combined flow-reservoir system in “soaker hose” concept

    Building from high-heat flux cooling schemes developed for solid PFCs

    Optimizing for size and coolant type (Helium vs. supercritical-CO2)

    Laboratory work establishing basic technical needs for PFC R&D

    Construction ongoing of LL loop at PPPL

    Tests of LI flow in PFC concepts in the next year

    Coolant loop for integrated testing proposed

    PPPL Liquid Metal R&D for Future PFCs

    For NSTX-U and Future Fusion Facilities

    Divertor Heat and Particle Flux

    Lithium Radiative Mantle

    Liquid Lithium Divertor Tray

    (LLDT)

    200°C – 400°C

    Valves

    EM Pumps

    Impurities

    M. Jaworski et al., PPPL


    Draft NSTX-U Research Plan metal layers

    Being Formulated


    Draft NSTX-U Research Facility Plan metal layers

    Being Formulated

    Upgrade Outage

    1.5  2 MA, 1s  5s

    Advanced PFCs, 5s  10-20s

    0.3-0.5 MA CHI

    0.5-1 MA CHI

    Start-up and ramp-up

    New

    center-stack

    Extend NBI duration or implement 2-4 MW off-axis EBW H&CD

    0.2-0.4 MA plasma gun

    up to 1 MA plasma gun

    ECH/EBW

    1MW

    2 MW

    Boundary physics

    Diagnostics for high-Z wall studies

    Divertor cryo-pump

    Divertor Thomson

    U.S. FNSF conceptual design including aspect ratio and divertor optimization

    Materials and PFCs

    All High-Z PFCs

    Hot High-Z FW PFCs

    U or L

    Mo divertor

    U + L

    Mo divertor

    Li granule injector

    Flowing Li divertor or limiter module

    Full toroidal flowing Li divertor

    Upward

    LiTER

    Lithium

    MGI disruption mitigation tests

    Enhanced RFA/RWM sensors

    NCC coils

    NCC SPA upgrade

    MHD

    Transport & turbulence

    DBS, PCI or other intermediate-k

    High kq

    dB

    polarimetry

    Waves and Energetic Particles

    HHFW straps for EHO, *AE

    Dedicated EHO or *AE antenna

    HHFW feedthru & limiter upgrade

    2nd NBI

    Scenarios and control

    Snowflake

    control

    Rotation control

    qmin control

    Control integration


    Summary metal layers

    • NSTX-U Aims to Develop Physics Understanding Needed for Designing Fusion Energy Development Facilities (ST-FNSF, ITER, DEMO, etc.)

    • Develop key toroidal plasma physics understanding to be tested in unexplored, hotter ST plasmas

    • Upgrade Project has made good progress in overcoming key design challenges

      • Project on schedule and budget, ~45-50% complete

      • Aiming for project completion in summer 2014

    • Detailed NSTX-U Research Plan is being developed


    ad