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Towards the development of Thermal-Hydraulic Models aimed at the Auxiliary Systems Design in Nuclear Fusion Technology 1L. Batet, 2L. Sedano, 1C. Queraltó, 1E. Mas de les Valls, 1J.Fradera and 1F. Reventós 1Technical University of Catalonia 2Association Euratom-Ciemat for Fusion

Contents of the presentation • Introduction • Background and context • LM Breeding Blanket designs in EU • BB channels CFD modeling • Auxiliary Systems for BB • Auxiliary systems modeling • EU experimental facilities • Conclusions

Introduction • As for Gen IV Programmes, the developments in Nuclear Fusion Technology (NFT) are today demanding advanced computational capabilities for simulation and modeling of thermal-hydraulics of liquid metal systems. • Attending to its functional requirements (DEMO Fusion reactors are today functionally conceived), breeding blanket (and auxiliary systems) is probably the most complex component developed today by energy technology. It represents a key component/system in the way to fusion power production (DEMO). Breeding Blanket modules are just behind the first-wall, facing plasma. They perform three challenging key functions: • Coil shielding from plasma radiation • Heat recovery from neutron flux • Tritium breeding

Introduction • EU plan to test HCLL and auxiliary systems as Test Blanket Module (TBM) in ITER. The development of predictive tools aimed to the simulation of the behavior and to the analysis of those systems and components will be a key issue to licensing and to take advantage of the results from ITER experimental programmes. • As DEMO breeding blanket, EU is developing diverse liquid metal (Pb15.7Li)-based concepts: • Helium-Cooled Lithium-Lead (HCLL), • Dual-Coolant/Dual-Functional Helium/Lithium-Lead (DCLL), and • Self-Coolant Lithium-Lead Concepts (SCLL). Water-coolant (WCLL) concept is developed at the French National Programme. A Spanish National Programme (TECNO_FUS) is starting in Spain for Dual-Coolant concept development.

Introduction • The Thermal-Hydraulic Studies Group (THSG) of the Technical University of Catalonia (UPC), in collaboration with CIEMAT, is trying to reach computational simulation capabilities in relation with breeding blanket liquid metal auxiliary systems for EU ITER TBMs (ITER Test Blanket Modules) and beyond (design of auxiliaries for DEMO HCLL and advanced modular DCLL-like named DRM He/LiPb) . • A system TH code, like RELAP5-3D, will allow the formulation of the main mass and energy balances. • The detailed analysis capabilities of a CFD code (UPC is using OPENFOAM) will provide support to the modeling of tritium diffusion and permeation, form-losses evaluation, etc.

Background and context • Currently foreseen fusion reactors relay on the D+T reaction, which implies achieving tritium self-sufficiency. • While deuterium is abundant in nature (30g/m3 of water), tritium is not. • For a 1000 MWe reactor, the mass of tritium needed is some 360 g/day (compare with the production of one CANDU reactor, close to 2 kg/year). • The Breeding Blanket is the solution proposed to produce tritium inside the reactor facility.

Background and context • Tritium can be produced from Lithium: • Probability of first reaction is orders of magnitude larger than the second one. • Natural lithium has 92.44 % 6Li: isotopic enrichment is needed. • Neutron multiplier needed (Pb or Be) to achieve the necessary Tritium Breeding Ratio (TBR) of about 1.15

Background and context • Tritium cycle is complex and tritium control is difficult. • Diffusive properties of hydrogen isotopes are very large compared with other elements. Permeation is a key issue regarding tritium inventory. • Inventory control is of utmost importance: • Environmental impact: authorized release (it will depend on the chemical form of the emissions) 1g/year (27 Ci/day) in front a production of some hundreds of grams/day. • Self-supplying guaranty (very small margin between the maximum achievable TBR and the TBR needed)

Background and context • To demonstrate the tritium self-sufficiency capability, along with licensing requirements, a detailed study of tritium inventory in the channels of the breeding blanket is mandatory. • Tritium analysis require an exact knowledge of the hydrodynamic profiles in the surface layers, which can only be obtained through CFD simulation. • CFD codes should be validated under the BB working conditions. • First full experiment for TBM will be ITER.

EU ITER-Test Blanket Module Test Blanket Module (TBM) : mock-up of a DEMO blanket in ITER test port TBM System : TBM + various associated systems in Tokamak & other buildings HCS = (He) Coolant System, located in TWCS vault, connected with ITER heat rejection system Joaquín Sánchez TES = Tritium Extraction System, located in Port Cell, connected Tritium building CPS = Coolant Purification System, TWCS vault

LM Breeding Blanket designs in EU ► Structures: RAF/M Steel (HCLL,WCLL) and (DCLL) or SiCf/SiC(SCLL) ► Multiplier/Breeder: Eutectic Pb-Li ► Coolant: He at 8 MPa, 300/500°C (only) or with LiPb at ~460-480°C/650-700°C Helium-Cooled LLE blanket (HCLL) Water-Cooled LLE blanket (WCLL) Dual-Coolant LLE blanket (DCLL) Self-Cooled LL blanket (SCLL) ► EUROFER He-cooled steel box, highly modular, radial He-cooling plates ► T steel: 350°C/550°C ► T interface (steel) < 520°C ► LiPb velocity < 1 mm/s ► Need of T-perm. Barr., ►6Li 90% ► Water-cooled FM steel box, large banana shape modules, poloidal tubes ► T steel: 350°C/550°C ► T interface (steel) < 520°C ► LiPb velocity ~ mm/s ► Need of T-perm. Barr. ►6Li 90% ► EUROFER/ODS+ SiC/SiC flow channel inserts as el./th.Insulator ►LLE breeder & coolant ►T steel: 350°C/550°C ►T interf. SiC/SiC< 700°C ► LLE velocity ~0.1 m/s ► 6Li 90% ► SiC/SiC large banana-like structures, ► LLE breeder & total coolant ► T interface (SiC/SiC) < 700°C ► LiPb velocity ~1 m/s ► 6Li 90%

LM Breeding Blanket designs in EU • CIEMAT and UPC are dealing with HCLL (ITER TBM) and DRM (for DEMO) Helium- Cooled Lithium-Lead ½ V port size . • In cooperation with ENEA and CEA, a PbLi database compiled. J.Nuc.Mat: E.Mas de les Valls et al., Lead-lithium eutectic material database for nuclear fusion technology. 2008

BB channels CFD modeling • THSG (UPC) is developing a CFD tool (based on the open source OpenFOAM code) to simulate the complex phenomena occurring in the liquid BB modules channels, including HCLL and the Dual Coolant Concept. • Steps to follow: • MHD implementation • Validation using cases with analytical solution • Preparation of the post-processing for tritium • Validation and analysis of numerical diffusion (DT~10-9 m2/s) • Temperature coupling • Validation (?) • Application to the BB modules

BB channels CFD modeling • To adapt and validate OpenFOAM it has been necessary to implement a suitable numerical algorithm to simulate the TBM working conditions, i.e. MHD advection coupled to the tritium diffusion/permeation under an intense magnetic field (high Hartmann, high Interaction Parameter) and large thermal loads (neutron radiation). • The implemented MHD code has been satisfactorily validated. • Future improvements include: • Implementation of thin wall boundary condition for finite conductivity walls [Müller,2001] • Implementation of wall functions [Muck,2000][Pothérat,2002] for Hartmann surface layers (work in cooperation with Alban Pótherat and Vincent Dousset of Coventry University)

BB channels CFD modeling • Tritium post-processing as a passive scalar has been implemented and is in a validation process. One difficulty is the low diffusion coefficient (DT~10-9 m2/s) which makes numerical diffusion to be dominant. • Appearance of lateral surface layer MHD instabilities must be analyzed in detail. • In general the results obtained with OpenFOAM as a MHD tool are satisfactory. • Work performed to date is key to future development that should allow the full CFD simulation of liquid metal TBM.

BB channels CFD modeling

BB channels CFD modeling • 3D application (4 nodii in Hartmann surface layer, 10 in lateral surface layer), total 777600 control volumes • Re=480 Ha=1740 Gr=2.3·107 and Sc=134 • Purely hydrodinamic case Some instabilities appear that are damped when a magnetic field is introduced • Temperature-MHD case (not shown): Due to buoyancy, vortices appear, further analysisi needed ST = 6.64·10-12 r -1.2473 mol/m3 s U (m/s) U (m/s) C (mol/m3) C (mol/m3)

BB channels CFD modeling • Currently there do not exist experiments considering all the phenomena (MHD, heat transfer, tritium and helium) • Experimental data for uncoupled phenomena (MEKKA, LIBRETTO irradiation tests) • This fact makes code validation difficult . • An international effort exists (F4E) aiming to the development of computational tools able to simulate the TBM channels.

Auxiliary Systems for BB Pb-Li Ancillary System for HCLL TBM Main goals: • To develop PbLi Ancillary System for the HCLL TBM • To develop PbLi Purification System to be incorporated in the PbLi ancillary system (e.g. cold trap, magnetic trap,…) • Study of underlying processes (e.g. PbLi purification, corrosion products transport – dissolution & deposition, impurities behaviour, transmutation products formation and behaviour,…) • Development and testing of suitable components (e.g. PbLi pump, connecting flanges, valves, diagnostics,…)

Auxiliary Systems for BB Specification for the PbLi ancillary system • PbLi mass flow in the range: 0.1-1 kg/s (0.3 kg/s is approx. 10 recirculation per day) • PbLi temperature range: 300-550°C • PbLi mass flow through the purification unit: TBD (estimated approx. 10% of total flow) • Inert atmosphere: Argon • Maximum pressure in the system: 8 MPa (He leak from HCS) • System structure material: ferritic steel (Cr 10-12%) • PbLi composition: • Pb-15.7at%Li eutectic alloy • Intermetallic compounds (e.g. PbLi) • Structure material corrosion products (e.g. Fe, Cr, Mn, Ni) • Impurities (e.g. Bi, Sn, Si, Al,…) • Impurities transmutation products (e.g. Po, Tl, Hg,…)

HCLL TBM Pb-Li auxiliary system (1) The PbLi auxiliary system should ensure feeding and circulation of PbLi liquid metal in the breeding blanket and removal of tritium produced by a nuclear reaction in TBM. Flow diagram of Pb-17Li ancillary circuit Pb-Li ancillary system flow diagram Connection flanges Heating Detritiation unit HCLL TBM PbLi filling CT draining Pb-Li system box Sampling Pb-Li draining Pb-Li tank Cold trap Mechanical circulation pump Conceived and designed NRI-IPP. Chek Rep.

HCLL TBM Pb-Li auxiliary system (2) Views of Pb-Li ancillary system box • The PbLi circuit is a closed loop with a slow forced circulation of PbLi. • Main components are: • a PbLi feeding and storage tank, • a liquid metal pump (variable flow velocity 0.1 to 1 kg/s), • a detritiation unit to remove Tritium from PbLi (R&D still ongoing), • a cold trap to remove corrosion products and impurities (R&D still ongoing). • The container with the PbLi auxiliary system (dimensions H x L x W: 2.315 m x 2.19 m x 1.6 m) will be placed in the port cell Conceived and designed NRI-IPP. Chek Rep.

DCLL power extraction system (interest of Sp. Nat. Program) • A system TH code, like RELAP5-3D, will allow the formulation of the main mass and energy balances. • Interest on advanced HEX 2 (LM 2ry SC-CO2). • Interest on compact permeators against vaccum • Interest on two phase models (helium bred in LiPb)

Auxiliary systems modeling • Challenges • Tritium and helium transport • Magnetic field • Complex geometries • A RELAP5-3D model has been developed at UPC of a hypothetic two-fluid experimental facility (partially inspired in Mekka) • Double loop: NaK/He

Auxiliary systems modeling Heat up during initial transient

Auxiliary systems modeling • Difficulties: • MHD head losses (it seems like RELAP5-3D should compute them, but it does not) have been not considered at the moment. They could be introduced as Re dependent form losses. • NaK used instead of PbLi, not only because it is the fluid in Mekka but because PbLi property tables are not available in our Athena version. • No axial fluid conduction.

Compatibility of EUROFER with Pb-Li EU experimental facilities • Compatibility tests at different Pb-Li temperatures 480-550°C and flow velocities 1-30 cm/s; HCLL TBM conditions • Various test facilities used (Picolo at FZK, LIFUS-2 at ENEA Brasimone, Pb-Li facility at NRI Czech Rep.) • Effect of magnetic field (1.7 T) on EUROFER corrosion in flowing Pb-Li (Latvia) LIFUS-2 loop PICOLO loop EMP = Electromagnetic Pump FM = Flowmeter TS = Test Section EH = Electrical Heater AC = Air cooler CFHE = Counter Flow Heat Exchanger MT = Magnetic Trap FZK ENEA-Brasimone

Form-closed barriers MEKKA laboratory EU experimental facilities Design of the experiment 4 breeder units, poloidal manifold, access tubes MHD experiments with a HCLL Mock-Up in the MEKKA laboratory at FZK

EU experimental facilities - Nominal mass flow rate: 0.2 kg/s - Maximum temperature: 500°C - Minimum temperature: 350°C - Stripping gas: Argon - LM inventory in the loop: about 25 l + 80 l in the circulation tank The TRIEX Facility: test of Tritium extractor, ENEA Brasimone

Conclusions • If nuclear fusion has to be a major source of energy in the future it is necessary to overcome some important challenges facing the international scientific community today. • Several areas of technological knowledge are involved in the design of presently conceived fusion reactors, and brought to their limits. • One of the most technology demanding components is the Breeding Blanket Module, together with the associated Tritium Extraction System.

Conclusions • UPC and CIEMAT, together with other Spanish organizations, are trying to acquire de simulation capabilities needed to design and to analyze such complex systems. • Simulation efforts focus in two lines: • MHD modeling of the BB channels (involving tritium transport and heat transfer) • System code modeling of the whole BB system (power extraction + ancillary systems)

Conclusions • CFD modeling is progressing at good pace, and results to date are encouraging. Though much effort still needs to be devoted to it. • System code simulation is just starting. The success will depend on our ability to take advantage of the existing code capabilities, and the possibility of improving the existing codes.

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