NSTX-U Status and Plan

A3 Foresight Workshop on ST Jan. 14- 15, 2013NSTX-UNSTX-U

NSTX-U Status and PlanMasayuki Ono

NSTX-U Project DirectorPPPL, Princeton University

In collaboration with the NSTX-U Team

Culham Sci CtrU St. Andrews

York UChubu UFukui U

Hiroshima UHyogo UKyoto U

Kyushu UKyushu Tokai U

NIFSNiigata U

Tsukuba UU Tokyo

JAEAHebrew UIoffe Inst

RRC Kurchatov InstTRINITI

NFRIKAIST

POSTECHSeoul National U

ASIPPENEA, Frascati

CEA, CadaracheIPP, Jülich

IPP, GarchingASCR, Czech Rep

Columbia UCompXGeneral AtomicsFIUINLJohns Hopkins ULANLLLNLLodestarMITNova PhotonicsNew York UORNLPPPLPrinceton UPurdue USNLThink Tank, Inc.UC DavisUC IrvineUCLAUCSDU ColoradoU IllinoisU MarylandU RochesterU Tennessee U WashingtonU Wisconsin

NSTX-UNSTX-U Supported by

The First A3 Foresight Workshop on Spherical Torus (ST)

Jan. 14-16, 2013

SNU, Seoul, Korea

A3 Foresight Workshop on ST Jan. 14- 15, 2013NSTX-UNSTX-U 2

Talk Outline• NSTX-U Mission

• NSTX Experimental Overview

• NSTX-U Construction Status

• NSTX-U Experimental Plan

• Summary


• 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 , 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-U Mission ElementsFusion applications of low-A spherical tokamak (ST)

3


• Confinement scaling (electron transport)• Non-inductive ramp-up and sustainment• Divertor solutions for mitigating high heat flux• Radiation-tolerant magnets (for Cu TF STs)

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

VECTOR (A=2.3)

JUST (A=1.8)

ARIES-ST (A=1.6)

Low-A Power Plants

Key issues to resolve for next-step STs

Parameter NSTX NSTX Upgrade

Fusion Nuclear Science Facility Pilot Plant

Major Radius R0 [m] 0.86 0.94 1.3 1.6 – 2.2

Aspect Ratio R0 / a 1.3 1.5 1.5 1.7

Plasma Current [MA] 1 2 4 – 10 11 – 18

Toroidal Field [T] 0.5 1 2 – 3 2.4 – 3

Auxiliary Power [MW] ≤ 8 ≤ 19* 22 – 45 50 – 85

P/R [MW/m] 10 20 30 – 60 70 – 90

P/S [MW/m2] 0.2 0.4 0.6 – 1.2 0.7 – 0.9

Fusion Gain Q 1 – 2 2 – 10

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

4


TF OD = 40cm

Previous center-stack

TF OD = 20cm

RTAN [cm]__________________

50, 60, 70, 130 60, 70,120,13070,110,120,130

IP=0.95MA, H98y2=1.2, N=5, T = 10%

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

2x higher CD efficiency from larger tangency radius RTAN

100% non-inductive CD with q(r) profile controllable by:tangency radius, density, position

New 2nd NBIPresent NBI

Normalized e-collisionality e* ne / Te2

ITER-like scaling

ST-FNSF

?

constant q, ,

NSTX Upgrade

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

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

Newcenter-stack

5

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

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


TF flex-bus

CS Casing

TF cooling lines

TF Coil

OH Coil

PF Coil 1a

PF Coil 1b

PF Coil 1c

CHI bus

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

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


The NSTX-U Inner TF Bundle Manufacturing Stages

New Zn-Cl-Free Soldering Technique Developed


NSTX-U Support Structural Upgrades

4x Electromagnetic Forces


•9

Relocation of the 2nd NBI beam line box from the TFTR test cell into the NSTX-U Test Cell.


•10

2nd NBI alignment performed and confirmed

A3 Foresight Workshop on ST Jan. 14- 15, 2013NSTX-UNSTX-U •11

Beam-line Component Refurbishment

Ion Dump Calorimeter upgrade Bending Magnet


JK cap tack welded to the vacuum vessel after completing alignments, and full welding is now

underway (Jan. 3, 2013)


40” Torus Isolation (Gate) Valve received

TVPS valves, hardware, TMPs,

and shields

TVPS valves, hardware, TMPs,

and shields

Spool section & supportsCircular bellows

Exit spool piece

Rectangular bellows

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

and fabricated


PF 1C

PF 1CCHI Gap

CHI Gap

CHI Gap

In-boardDivertor

Out-boardDivertor

HHFWAntenna

Center Stack

Primary Passive Plates

Secondary Passive Plates

NBI Armor

NSTX In-Vessel View and CHI Gap Protection Enhancement

Expect x 10 Higher Heat Load Into the CHI Gap


• 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 (4080%) at low IP = 0.4MA– 1.5-2x higher current drive efficiency

Present NBIMore tangential

2nd NBI

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

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

15


NSTX-U CHI Start-up ConfigurationsX 2 Higher CHI Driven Currents Expected


NSTXPlasma

Low-lossCorrugatedWaveguide

SteerableMirror

Launcher

NSTX Vacuum Vessel

28 GHz – 1MW Gyrotron by U. of Tsukuba

A schematic of the NSTX-U ECH/EBW launcher

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


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

– Reached 2 times computed n = 1 no-wall limit of N/li = 6.7

• Lower probability of unstable RWMs at high N/li

N

li

BetaN vs.li - Gridlines

0

1

2

3

4

5

6

7

8

0.0 0.2 0.4 0.6 0.8li

beta

N

N/li 13 12 11 10n = 1 no-wall beta limit line

0

1

2

3

4

5

6

7

8

0.0 0.2 0.4 0.6 0.8li

beta

N

14

N/li = 6.7

BetaN vs.li - XP948

0

1

2

3

4

5

6

7

8

0.0 0.2 0.4 0.6 0.8li

beta

N

n = 1 no-wall limit

ST-CTFST-Pilot

RWM State Space Control

n =

1 R

FA (

G/G

)

N/li5 10 15

1.5

1.0

0.5

0.0

Mode stability directly measured in experiments using MHD spectroscopy

Stability decreases up to N/li = 10

Stability increases at higher N/li Presently analysis indicates consistency

with kinetic resonance stabilization

Resonant Field Amplification (RFA) vs. N/li

unstablemode

J. Berkery IAEA

Unstable RWMStable / controlled RWM

S.A. Sabbagh

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

18


• Disruption warning algorithm shows high probability of success

– Based on combinations of single threshold based tests

Results ~ 98% disruptions flagged with at

least 10ms warning, ~ 6% false positives

False positive count dominated by near-disruptive events

Disruptivity

Physics results Low disruptivity at relatively high N ~ 6;

N /Nno-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

Warning Algorithms

S. Gerhardt IAEA

All discharges since 2006

N

liq*

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

19


• 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

• Higher flux expansion (increased div wetted area)• Higher divertor volume (increased div. losses)

V. Soukhanovskii, NF 2009

NSTX “Snowflake” Divertor Configuration resulted in significant divertor heat flux reduction and impurity screening

20

NSTX-U snowflake simula


With Lithium

Without Lithium

Lithium Improved H-mode Performance in NSTX

Te Broadens, E Increases, PH Reduces, ELMs Stabilize

Pre-discharge lithium evaporation (mg)

Te broadening with lithium

E improves with lithium

R. Maingi, PRL (2011)

21

E improves with lower collisionality

S. Kaye, IAEA (2012)

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

H. W. Kugel, PoP 2008


• 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.

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

M. Podesta, IAEA (2012)

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

LiquidSolid

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)

22


•a)

•b)

•c)

•d)

• 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

•Lithiated graphite

T. Gray. IAEA 2012

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

23


Radiative Liquid Lithium Divertor Proposed Based largely on the NSTX Liquid Lithium Divertor Research

Flowing LL Particle Pumping Surfaces

Li+

Li++

Li+++

Li0

Heat Exchanger

B0

Divertor Heat and Particles Flux

Liquid Lithium (LL)

~ 1 l/sec

LL Purification System to remove tritium, impurities, and dust

Li Evap. /

Ionization

Li Radiative Mantle

Li wall coating /condensationLi path

Reduced Divertor Heat and Particle

Flux

Particle pumping by Li

coated wall

Divertor Strike Point

M. Ono. IAEA 2012

First Wall / Blanket

At 500°C – 700°C

000000000000000000000000

Core Reacting

Plasma

Edge Plasma

Scrape Off Layer

Flowing LLD Tray

200 – 450 °C

Closed RLLD

LL Out LL InLL In

RLLD

24


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 PFCsFor NSTX-U and Future Fusion Facilities

25

M. Jaworski et al., PPPL

Valves

Impurities

EM Pumps

Liquid Lithium Divertor Tray

(LLDT)

200°C – 400°C

Divertor Heat and Particle Flux

Lithium Radiative

Mantle


Draft NSTX-U Research Plan Being Formulated

26


Draft NSTX-U Research Facility Plan Being Formulated

27

2014 2015 2016 2017 2018 2019 2020 2021 2022 2023

1.5 2 MA, 1s 5s

NCC coils

0.2-0.4 MA plasma gun

0.3-0.5 MA CHI

0.5-1 MA CHINew

center-stack

2nd NBI

Upgrade Outage Advanced PFCs, 5s 10-20s

up to 1 MA plasma gun

U or LMo divertor

U + LMo divertor

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

1MW 2 MWECH/EBW

Divertor cryo-pump

UpwardLiTER

Flowing Li divertor or

limiter module

Li granule injector

All High-Z PFCs

Full toroidal flowing Li divertor

HHFW straps for EHO, *AE

Hot High-Z FW PFCs

NCC SPA upgrade

Enhanced RFA/RWM sensors

DBS, PCI or other intermediate-kHigh k

Bpolarimetry

Divertor Thomson

Dedicated EHO or *AE antenna

HHFW feedthru & limiter upgrade

U.S. FNSF conceptual

design including

aspect ratio and divertor optimization

Rotation control

qmin control

Snowflakecontrol

Control integration

Diagnostics for high-Z wall studies

MGI disruption mitigation

tests

Start-up and

ramp-up

Boundary physics

Materials and PFCs

Lithium

MHD

Transport & turbulence

Waves and Energetic Particles

Scenarios and control

A3 Foresight Workshop on ST Jan. 14- 15, 2013NSTX-UNSTX-U 28

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

Documents

NSTX-U Status and Plan