Design Strategies for High Availability: Accommodating In-Vessel Piping Services

and Auxiliary Systems

Tom Brown Princeton Plasma Physics Laboratory

3rd IAEA-DEMO Program workshop 11-14 May 2015

All fusion options need a machine arrangement that fosters high availability

The spherical Tokamak, conventional Tokamak and

stellarator physics concepts all contain details which

challenges maintenance operations.

As with the conventional tokamak device, the ST and Stellarator configuration also must address similar maintenance issues

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design options to enhance in-vessel component maintenance, that

accommodates diagnostics and heating systems and adds space for blanket piping services.

The focuses of this talk looks at:

Maintenance operations were central to the PPPL/TE ST- FNSF device defined with vertical maintenance

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10 HTS TF

magnet system

Vertical maintenance

Coolant supply at the

bottom

Cryostat dome insert

Upper PF coil support

structure

PF support off

TF coil Outboard blanket

VV access cover

Some prime ST characteristics include: • Low no. of TF coils and IV parts– 10 TF and 20 outboard

blanket modules, • With only inboard shield modules needed to protect TF coils, • Low heat flux double-null Super-X divertors

Leading design issues include: • Space needed for blanket piping and diagnostics may compete

with placement of outboard PF coils • Achieving TBR>1 - Issues that will be addressed within the

PPPL ST-FNSF study this FY

PPPL HTS-ST FNSF device

configuration

Stellarator maintenance was improved with the effort to balance plasma surface reconstructions with targeted physics

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The PPPL COILOP physics code was upgraded to generate MC currents and shapes that are more amenable to stellarator maintenance.

Code changes include:

A high AR device has been

designed that meets the

targeted ARIES-CS plasma

boundary

Fixed Boundary

Free Boundary

straightened MC back legs and engineering supplied surface geometry to locate the MC winding centers,

Fourier coefficients are still used to define the plasma surface but the MC winding geometry now is developed with spline representations,

coding effort was also made to smooth the shaping of the modular coil winding, add torsion constraints, target coil parameters and freeze coil geometries, and much more.

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Greater access improves blanket maintenance

Moving from the 3-period quasi-axisymmetric ARIES-CS parameters (A=4.5, R0 = 7.75m, B = 5.7T) to an AR 6.0 configuration while retaining fusion power, beta, plasma volume, and toroidal magnetic field leads to a 9.4m R0 device with straightened back legs for Type A and B MC’s that meets maintenance requirements and matches the targeted plasma boundary.

ARIES-CS 4.5 AR, 7.75-m Raxis

Maintaining self-consistency between engineering and physics improved maintenance feasibility

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Multiple DEMO Tokamak options have been studied

Adapted from paper by Hiroyasu Utoh et. al. “Critical Design Factors for Sector Transport Maintenance in DEMO”

U.S. ARIES-AT Horizontal access

16 radially extended TF coils

Toroidal segmentation

16 modules

16-port horizontal removal

16-port vertical

maintenance

EU multi-module concept

Vertical access JAEA, DEMO

Limited horizontal access

12 radially extended TF coils

Toroidal segmentation

12 modules

4-port horizontal removal

16 close-fitting TF coils

Toroidal & radial segmentation

64 modules

16-port vertical removal

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As with ITER, vertical installation will be used to assemble DEMO – setting the stage for vertical maintenance

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The building space above the

device is set by machine

assembly requirements

Assembly within

the tokamak pit

ITER tokamak building

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Maintenance cycle to achieve 75%

availability for planned maintenance

Maintenance duration estimate for a DEMO fusion power plant, basedon the EFDA WP12 pre-conceptual studies, O. Crofts and J. Harman, Fusion Engineering and Design 89 (2014) 2383–2387

Many features will impact maintenance Two areas addressed:

Auxiliary system interfaces Blanket piping services

All DEMO designs strive for maintenance compatible arrangements that fosters high availability

Blanket piping through upper port

EU Demo

Alternate design strategies need to be developed and evaluated

Can maintenance improve with the added space of

an enlarged VV and TF coil?

Within a vertical maintenance scheme, can

maintenance be enhanced with blanket piping

services coming from below?

Can auxiliary systems retract without dismantling

the integrated system?

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Defining a feasible fusion design that operates with high

availability is challenging…there is a need to develop many

competing concepts in order to develop one feasible solution.

Question:

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Construction issues with large TF coils can be alleviated by a two winding scheme

Incorporating a pair of windings reduces pressure drop for long

CICC winding lengths of enlarged coils and reduces overall costs.

Initial calculations by Keeman Kim found a reduction for K-DEMO two-winding design to be approximately $1.7 B, compared with a single winding scheme.

Regardless of coil size the graded TF approach results in a more compact winding X-section, allowing more space for structure and higher field operation.

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Port requirements for H&CD and diagnostics impacts the blanket design

E. Magnani - DEMO Technical Meeting - Garching – 29,30 September 2009

Large mid-plane openings for auxiliary systems can cut the continuity of vertical blanket sector complicating piping, support and service details

EU MMS with small port opening

EU MMC with large port opening

CFETR

All interfacing auxiliary systems need to be disengaged for blanket removal

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A blanket module needing replacement requires retraction of all auxiliary systems interfacing with it.

The IV device design can limit the size of the equatorial diagnostic portplug

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Given a K-DEMO 16 TF coil arrangement, the desire to use equal toroidal blanket segments and the requirement not to split a blanket sectors vertically, resulted in toroidal space allowing a two segment equatorial portplug to be used instead of the ITER defined 3 segment diagnostic portplug.

ITER 3 segment diagnostic equatorial Portplug

Can dismantling all blanket interfacing auxiliary systems occur within the one month planned cool down time?

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Section cut of a typical ITER equatorial diagnostic port

Can auxiliary systems

retract without

dismantling the integrated system?

This illustration defines a concept design to disengage auxiliary systems without the need to dismantle them

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Insert Portplug in

blanket

Retract installation

support guide

Operating position

Retract Portplug for blanket

removal

Collar outside Portplug for attachment and neutron streaming

22 tonne

K-DEMO has evolved as a vertical maintenance design with expanded space for maintenance

Design features have been added to enhance maintenance of in-vessel systems:

A structural system that supports disruption loads and provide alignment for blanket instillation has been defined,

blanket segmentation has been sized for mid-plane auxiliary port openings without splitting sectors,

blanket labyrinth interfaces has been incorporated to minimize gap streaming of neutrons, and the general arrangement

promotes a facility design that allows port auxiliary systems to retract without dismantling the auxiliary equipment

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Still work in progress - 2014 version of K-DEMO

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The K-DEMO in-vessel configuration design is progressing

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• The concept is designed to support disruption loads,

• Provide alignment for blanket instillation, and

• Incorporates labyrinth interfaces to minimize gap streaming of neutrons.

16 inboard blanket modules

32 0utboard blanket modules

(1 under TF, 1 between TF}

DN vs SN present implications for physics performance and machine design

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DN operation requires space for a second divertor, reducing tritium breeding area…and impacts divertor maintenance,

A second divertor requires added supports, maintenance and services details,

A DN arrangement requires diagnostic viewing and pumping at both locations which adds additional equipment, but

A DN divertor has symmetrical PF arrangements which allow large

vertical ports top and bottom and provides the potential to incorporate advanced divertor designs,

A DN divertor has lower heat load on the inner divertor with reduced divertor length required… increasing space for TF shielding.

A DN inboard divertor and baffle has low heat loads with an expected lifetime longer than the outer divertor segment. Can we take advantage of this?

The double-null plasma and selected blanket breeding approach impacts divertor maintenance

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To reach TBR > 1 with K-DEMO

baseline solid breeding blankets

in a DN arrangement, divertor

modules can be replaced only by

removing some of the outboard

blanket modules.

Lower access port used to

disconnect water or LM pipes

Are there further divertor options to be considered that will enhance the device maintenance?

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Is it feasible to split the divertor assembly in a DN arrangement to foster maintenance advantages?

Can the SN divertor be located on the top to enhance maintenance for a vertical maintained device ?

As with the ST-FNSF design, can a conventional DN tokamak be configured with a Super-X divertor?

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Concluding remarks

For fusion to succeed not only must the plasma physics and the accompanying machine technology operate at creditable levels but the topology in which they exist must be amenable to inspection and maintenance to allow the device to operate with high availability.

All fusion devices built to date are physics centered experiments.

Serious effort is needed to look at options to define a magnetic fusion design that provides self-consistency between the plasma performance

and engineering requirements for component maintenance.

Thank you for your attention…