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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

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slide1
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

slide2
Talk Outline
  • NSTX-U Mission
  • NSTX Experimental Overview
  • NSTX-U Construction Status
  • NSTX-U Experimental Plan
  • Summary
slide3
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
slide4
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)

slide5
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

slide6
A schematic of the new center-stack and the TF joint area

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

slide7
The NSTX-U Inner TF Bundle Manufacturing Stages

New Zn-Cl-Free Soldering Technique Developed

slide8
NSTX-U Support Structural Upgrades

4x Electromagnetic Forces

slide11
Beam-line Component Refurbishment

Ion Dump

Calorimeter upgrade

Bending Magnet

  • 11
slide12
JK cap tack welded to the vacuum vessel after completing alignments, and full welding is now underway (Jan. 3, 2013)
slide13
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

slide14
NSTX In-Vessel View and CHI Gap Protection Enhancement

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

slide15
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

slide16
NSTX-U CHI Start-up Configurations

X 2 Higher CHI Driven Currents Expected

slide17
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

slide18
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

slide19
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
slide20
NSTX “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

slide21
Lithium Improved H-mode Performance in NSTX

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)

slide22
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)

slide23
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
slide24
Radiative Liquid Lithium Divertor Proposed

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

slide25
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

slide27
Draft NSTX-U Research Facility Plan

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

slide28
Summary
  • 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
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