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

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1 3 rd IAEA-DEMO Program workshop May 2015 Design Strategies for High Availability: Accommodating In-Vessel Piping Services and Auxiliary Systems Tom Brown Princeton Plasma Physics Laboratory

2 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. The focuses of this talk looks at: design options to enhance in-vessel component maintenance, that accommodates diagnostics and heating systems and adds space for blanket piping services. As with the conventional tokamak device, the ST and Stellarator configuration also must address similar maintenance issues 2

3 Maintenance operations were central to the PPPL/TE ST- FNSF device defined with vertical maintenance Vertical maintenance 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 Cryostat dome insert 10 HTS TF magnet system Coolant supply at the bottom PPPL HTS-ST FNSF device configuration VV access cover Outboard blanket Upper PF coil support structure PF support off TF coil 3

4 Stellarator maintenance was improved with the effort to balance plasma surface reconstructions with targeted physics The PPPL COILOP physics code was upgraded to generate MC currents and shapes that are more amenable to stellarator maintenance. Code changes include: 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. Fixed Boundary Free Boundary A high AR device has been designed that meets the targeted ARIES-CS plasma boundary 4

5 Maintaining self-consistency between engineering and physics improved maintenance feasibility ARIES-CS 4.5 AR, 7.75-m R axis Greater access improves blanket maintenance (36) Moving from the 3-period quasi-axisymmetric ARIES- CS parameters (A=4.5, R 0 = 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 R 0 device with straightened back legs for Type A and B MC s that meets maintenance requirements and matches the targeted plasma boundary. 5

6 Multiple DEMO Tokamak options have been studied U.S. ARIES-AT Horizontal access JAEA, DEMO Limited horizontal access EU multi-module concept Vertical access 16-port vertical maintenance 16 radially extended TF coils Toroidal segmentation 16 modules 16-port horizontal removal 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 Adapted from paper by Hiroyasu Utoh et. al. Critical Design Factors for Sector Transport Maintenance in DEMO 6

7 As with ITER, vertical installation will be used to assemble DEMO setting the stage for vertical maintenance The building space above the device is set by machine assembly requirements Assembly within the tokamak pit ITER tokamak building 7

8 All DEMO designs strive for maintenance compatible arrangements that fosters high availability Blanket piping through upper port Maintenance cycle to achieve 75% availability for planned maintenance Many features will impact maintenance Two areas addressed: Auxiliary system interfaces Blanket piping services EU Demo Maintenance duration estimate for a DEMO fusion power plant, basedon the EFDA WP12 preconceptual studies, O. Crofts and J. Harman, Fusion Engineering and Design 89 (2014)

9 Alternate design strategies need to be developed and evaluated 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: 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? 9

10 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. 10

11 Port requirements for H&CD and diagnostics impacts the blanket design 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 CFETR EU MMC with large port opening E. Magnani - DEMO Technical Meeting - Garching 29,30 September

12 All interfacing auxiliary systems need to be disengaged for blanket removal A blanket module needing replacement requires retraction of all auxiliary systems interfacing with it. 12

13 The IV device design can limit the size of the equatorial diagnostic portplug ITER 3 segment diagnostic equatorial Portplug 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. 13

14 Can dismantling all blanket interfacing auxiliary systems occur within the one month planned cool down time? Section cut of a typical ITER equatorial diagnostic port Can auxiliary systems retract without dismantling the integrated system? 14

15 This illustration defines a concept design to disengage auxiliary systems without the need to dismantle them Insert Portplug in blanket 22 tonne Retract installation support guide Operating position Collar outside Portplug for attachment and neutron streaming Retract Portplug for blanket removal 15

16 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 16

17 Still work in progress version of K-DEMO 17

18 The K-DEMO in-vessel configuration design is progressing 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} 18

19 DN vs SN present implications for physics performance and machine design 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? 19

20 The double-null plasma and selected blanket breeding approach impacts divertor maintenance 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 20

21 Are there further divertor options to be considered that will enhance the device maintenance? 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? 21

22 Concluding remarks All fusion devices built to date are physics centered experiments. 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. 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 22

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