Tritium breeder blankets design and technologies in Europe: Development status of ITER Test Blanket Modules, test & qualification strategy and roadmap towards DEMO

Fusion Engineering and Design 85 (2010) 2340–2347

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Tritium breeder blankets design and technologies in Europe: Development status of ITER Test Blanket Modules, test & qualification strategy and roadmap towards DEMO Y. Poitevin a,∗ , L.V. Boccaccini b,1 , M. Zmitko a , I. Ricapito a , J.-F. Salavy c,1 , E. Diegele a , F. Gabriel c,1 , E. Magnani b,1 , H. Neuberger b,1 , R. Lässer a , L. Guerrini a a

Fusion for Energy (F4E), Barcelona, Spain Institut fuer Neutronenphysik und Reaktortechnik, FZK, Karlsruhe, Germany c CEA Saclay, DEN/DM2S, F-91191 Gif-sur-Yvette, France b

a r t i c l e

i n f o

Article history: Available online 30 October 2010 Keywords: TBM DEMO Breeder blanket Design Technology

a b s t r a c t Europe has developed two reference tritium breeder blankets concepts that will be tested in ITER under the form of Test Blanket Modules: (i) the Helium-Cooled Lithium–Lead which uses the liquid Pb–15.7Li as both breeder and neutron multiplier, (ii) the Helium-Cooled Pebble-Bed with lithiated ceramic pebbles as breeder and beryllium pebbles as neutron multiplier. An extensive development program has been carried-out over the last decade combining experimental and numerical simulations aimed at identifying and quantifying physical phenomena occurring in breeder blankets and at optimizing their design and technologies accordingly. On this basis, sound guidelines for the design and technological choices of Test Blanket Modules can be derived. In addition, regulatory and ITER project requirements, which prefigure the future DEMO blanket ones, are now integrated in the development and qualification program of the Test Blanket Modules. Their scope and implication are analyzed. © 2010 Elsevier B.V. All rights reserved.

1. Introduction Europe has developed two reference tritium breeder blankets (BB) concepts that will be tested in ITER under the form of Test Blanket Modules (TBMs) [1] (Figs. 1 and 2): (i) the Helium-Cooled Lithium–Lead (HCLL) which uses the liquid Pb–15.7Li as both breeder and neutron multiplier, (ii) the Helium-Cooled Pebble-Bed (HCPB) with lithiated ceramic pebbles as breeder and beryllium pebbles as neutron multiplier. Both concepts are using the EUROFER reduced activation ferritic-martensitic (RAFM) steel as structural material and pressurized helium technology for heat extraction (8 MPa, 300–500 ◦ C). A safe and reliable blanket design requires fulfilling regulatory requirements as well as additional standards that are specific to a tokamak environment. Section 2 is dedicated to the analysis of regulatory and ITER requirements which prefigure the ones of a future

∗ Corresponding author at: Fusion for Energy, c/ Josep Pla 2, Torres Diagonal Littoral, Ed. B3 Office 7-18, 09019 Barcelona, Spain. Tel.: +34 93 320 1812; fax: +34 93 320 1803. E-mail address: [email protected] (Y. Poitevin). 1 European Consortium of Associates for TBM. 0920-3796/$ – see front matter © 2010 Elsevier B.V. All rights reserved. doi:10.1016/j.fusengdes.2010.09.027

DEMO blankets. Section 3 addresses specifically the manufacturing of the TBM box and its qualification roadmap according to the relevant codes & standards. A sound breeder blanket design and the related technological choices require also developing a deep understanding and prediction capability of the physical phenomena involved. This is usually obtained through a R&D program implementing dedicated experiments/simulation, first focused on phenomenological aspects to acquire a good knowledge of physical and technical issues involved and their relevancy to the envisaged application. Then, experiments and numerical simulations are run at a more design specific level, aimed at qualifying the predictive tools and validating the design options. The European blanket R&D program has been addressing both levels, from phenomenological issues up to design validation experiments. The main outcome of this program is summarized in Sections 4–7, in the fields of thermal-hydraulics, pebble bed thermo-mechanics, magneto-hydrodynamics (MHD) and tritium control. Also, the key expected outcome from TBMs testing in ITER is highlighted, as well as the possible limitations for the validation of a DEMO blanket. One will refer also to [1] for a description of TBM systems design, to [2] for the development status of blankets functional materials and to [3] for the TBMs ancillary systems.

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Fig. 2. HCPB Test Blanket Module.

Fig. 1. HCLL Test Blanket Module ((A) first wall, (B) stiffening plate, (C) cooling plates, (D) rear collector system).

2. Codes & standards TBMs operated in ITER facility have to conform two types of requirements: - Regulatory requirements, namely the European Pressurized Equipments Directive (PED) and the French ESPN Order for nuclear pressurized equipments; - ITER project requirements, the ones having the major impact on the TBM design and qualification being the General Safety Specification Requirements (GSSR), the Structural Design Criteria for In-Vessel Components (SDC-IC), the ITER Vacuum Handbook and the Reliability–Availability–Maintainability–Inspectability (RAMI) objectives. Other ITER project requirements are presently

under development in particular in view of equipments standardization. They are not discussed here because they are today of secondary impact on the TBMs design and qualification. A preliminary analysis of PED & ESPN regulatory requirements, to be confirmed by on-going detailed analyses, shows that the HCLL and HCPB TBMs shall be classified in category IV (pressure risk) and level N2 (nuclear risk). More precisely, the TBM helium pressurized compartments of the TBMs are classified in PED category IV due to the fact that the (service pressure × volume) exceeds 1000 bar l. The helium pressurized compartments do however not fall by themselves under the ESPN regulation because their normalized activity stays below the limit of 370 MBq (in ESPN Order, the tritium activity is to be divided by a factor 1000). However the HCLL PbLi and the HCPB breeder/multiplier compartments are classified by ESPN in level N2 because their activity overreaches the limit of 370 GBq. As a result, and because the TBM is being considered as a multiplecompartments component, its overall classification shall be the highest of each compartment, i.e. category IV, level N2.

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As a basis of the European strategy to conform to PED and ESPN requirements, it is recalled that: - The PED establishes Essential Safety Requirements (ESRs) for pressure equipments and, according to the standard FD E86-000, EN European Standards represent for the designer/manufacturer a privileged tool of presumption of conformity with the ESRs. - The ESPN Order requires the use of a professional guide to reach conformity. In Europe, the RCC-MR is the certified code the closest to the scope of TBM design and fabrication. However, it does not cover today the EUROFER material or some damage risks that are specific to tokamak in-vessel components (e.g. disruption, 14 MeV neutrons irradiation, etc.) as well as most of the welding processes envisaged for the TBM fabrication (e.g. diffusion and laser welding). In order to define a proper strategy able to meet the regulatory requirements, an exhaustive review of the position of TBMs design and fabrication with respect to RCC-MR and EN Standards has been performed [4]. Two key outcomes are: - Position of EUROFER steel with regard to EN standards and RCC-MR: According to standard EN 10020 (“Definition and classification of grade of steels”), the EUROFER steel belongs to the special alloy steels group which are characterized by an accurate adjustment of their chemical composition and their elaboration, giving them improved properties which are explicitly provided in combination and limited range. Also, EUROFER has also been developed on the basis of 9Cr martensitic steels resisting to creep like the Mod. 9Cr–1Mo steel, which is actually included in the EN 10088-1 (stainless steels) due to its belonging to the class of steels resisting to creep and despite the fact that its chromium content is lower to standard stainless steel. Consequently, standards dedicated to stainless steels resisting to creep are not strictly applicable to EUROFER but may be seen as containing some applicable rules that can be used for the EUROFER steel. The same analogy is made with regard to RCC-MR and rules related to low-alloy steels are considered. - Code strategy: A single code strategy (e.g. RCC-MR) is today insufficient for ITER TBMs. It has to be completed by other codes/standards, e.g. RCC-MX, EN, etc. In particular the RCC-MX includes design rules taking into account fluence and fabrication qualification rules for diffusion welding and beam welding processes. Also, although SDC-IC is not a certified professional guide – in particular, it does not allow today establishing any compliance with regard to PED & EPSN – it contains a set of design rules issued to take into account specificities of in-vessel components. In view of TBM licensing and, on a longer term, of DEMO design, it shall also be noted that the objective of the RCC-MRx Edition 2011 is to merge the RCC-MR and the RCC-MX into a single code. It is also recalled that SDC-IC is today widely based on RCC-MR/X rules. A long term strategy has now to be built in Europe for the development of a blanket relevant code which could, for instance, consist in completing already certified codes (e.g. RCC-MRx). This would require the integration of specific blanket structural material(s) appendix(es), design rules and welding processes. 3. Module box fabrication 3.1. Fabrication technologies The fabrication of TBM subcomponents, namely the first wall, box covers, stiffening plates and cooling plates, has to address specific difficulties like rectangular box shape, internal grid struc-

ture, meandering channels embedded within plates for pressurized helium coolant circulation. Several diffusion-welding (DW) based processes have been developed implementing Hot Isostatic Pressing (HIP) and/or arc/beam welding processes to comply with design specifications. An exhaustive review of the achieved processes development is given in [5]. The main outcome is briefly summarized hereafter. For the fabrication of TBM subcomponents (first wall, cooling plate, stiffening plate, covers): - YAG laser + HIP-DW process: This process consists basically in reconstructing the cooling channels by YAG laser welding of strips over a grooved plate, surfacing, positioning of a cover plate and applying a final HIP cycle with channels open to HIP pressure. The feasibility of this process was demonstrated on a cooling plate test mock-up with straight channels resisting up to 3000 thermal cycles. Two mock-ups with U-turn internal channels have also been successfully fabricated. This process allows fabricating thin plate subcomponents like the TBM cooling plates (thickness 6.5 mm) but it is also applicable to any other subcomponents (first wall, covers, stiffening plates) although its performance for curved shapes and larger surfaces was not yet demonstrated. - 2-Steps (closed/open channels) HIP-DW process: It is based on the diffusion welding of grooved plates. The bond line is located at the mid-height of the channels. During this first HIP cycle the channels are closed (i.e. not open to the HIP pressure) but the HIP external pressure is low enough to avoid deformation of the channels. After the first HIP cycle, the joints between channels are leak tight and the channels can be opened to apply a second high pressure HIP cycle. This fabrication process is restricted to subcomponents having a channels/ribs structure sufficiently resistant to the first HIP cycle conditions. Its feasibility was demonstrated on a first wall small-scale mock-ups resulting in good joint properties (e.g. impact toughness at 80% to the base material). - U-DW and HIP-DW (closed channels) processes: Uniaxial (U)-DW appeared difficult to transfer to industry and larger components due to the need of specific equipments and the difficulty to control the uniaxiality of the load and the uniformity of temperature. The HIP-DW gave promising results, even up to large mock-ups like a ¼ FW, but further developments are needed for reaching proper joint properties and overall deformation. - HIP-DW of rectangular tubes process: The HIP-DW of rectangular tubes consists in stacking rectangular EUROFER tubes together between two cover plates and then joining the whole assembly together by HIP-DW. Bending of rectangular tubes resulted in significant and uncontrollable geometrical deviations requiring further developments if this process shall be envisaged. Alternatively, rectangular tubes were re-constructed by longitudinal and butt-weldings of straight and curved machined pieces. But, some butt-welded areas were too brittle to withstand a stiff clamping and HIP cycle. This process is limited to the first wall and not applicable to subcomponents featuring internal channels with sharp turns. - Tubes forming + HIP-DW process: The tubes forming + HIP-DW process consists in cold forming soft-annealed round tubes to a rectangular shape and then inserting them into grooves machined in two symmetrical plates for final assembly by HIP-DW. It appears today complex to implement in particular due to the difficulty to predict conditions for tubes rupture during cold forming. It is also limited to the first wall and cannot be applied for subcomponents having internal channels with sharp turns. Concerning the box assembly by welding of subcomponents, it was found that:

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- Welding of 11 mm thick stiffening plates by conventional arc/beam process requires a multi-pass procedure, typically 2 passes with a 4 kW YAG laser and 6 passes with a 130 A TIG process. The induced deformations and stresses in the vicinity of the cooling channels that are located close to the welding joint led to investigate advanced concepts allowing single-pass and higher welding speed. Promising results have been obtained with a 10 kW YAG laser process up to the fabrication of three one-cell mock-ups where deformation remained limited (±60 ␮m). Multipass TIG welding might however still be needed in some steps of the stiffening grid assembly because gaps between plates will cumulate along the fabrication sequence. Therefore the TIG welding process shall be further optimized in particular with regard to distortion and effect on the neighboring channels. - The assembly of thicker box subcomponents like the first wall and the covers (30–35 mm thick) will necessitate a multi-passes welding process like the Narrow Gap TIG welding or the advanced Hybrid MIG/laser process. Both processes shall be discriminated in terms of residual deformation of the assembled TBM box. 3.2. Qualification The position of TBMs fabrication processes with regard to European EN standards and French RCC-MR was investigated [4]. The key standard considered for the specification and qualification of welding procedure is the EN 15607. It defines, first, the appropriate standard(s) specifying the format of the Welding Procedure Specification (WPS), depending on the welding process. In the case of TBM application, WPS format to be used for arc welding and laser beam welding is mentioned but not for other processes like diffusion welding. In this latter case, it will be necessary to propose to the regulator a WPS based on the closest available standards (e.g. RCC-MX). The standard EN 15607 defines also five methods, and their conditions of application, for the qualification of the WPS. In the case of TBM application, the only method that appears a priori applicable is the “qualification based on a pre-production welding test” (EN ISO 15613). This method is the only reliable method of qualification for welding procedures, in which the resulting properties of the weld strongly depend on certain conditions that cannot be reproduced by standardized test pieces. In the case of the TBM, one example of inadequacy between standardized/actual geometry is the butt welding for the grid assembly where the presence of a perpendicular wall in the vicinity of the weld perturbs the flow of the gas shroud compared to the geometry of a standardized flat sample and could result in an over-heating on one side of the weld. The standard EN 15613 (“qualification based on a pre-production welding test”) applies to fusion welding processes (arc or laser welding) but may also apply to other processes like diffusion welding. The section IV of the RCC-MR Edition 2002 has been also reviewed in particular with respect to rules proposed by the code for achieving qualification of welding processes. In general the content of this section is close to the relevant EN standards, but the RCC-MR often amplifies or completes the requirements of EN standards. A major difference lays in the fact that the unique method envisaged for arc WPS is the welding procedure test as specified in paragraph RS 3200. The method based on pre-production welding test is not envisaged. However the ESPN Order highlights the need to perform an adequate qualification of the entire fabrication sequence, in particular for the components featuring a risk of heterogeneity of their characteristics due to the complexity of the fabrication operations. Applied on the TBM, this requirement is actually close to the case covered by the EN standard for qualification base on pre-production welding test. For qualification of DEMO blankets, attention shall be paid to later developments in fabrication processes which might require a re-qualification. A particular attention shall be paid to diffusion-

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welding processes where the surface state and the size/volume of the HIPed components are key elements of qualification. For other welding processes (beam/arc welding), the welding position is also an element which would require a re-qualification of the welding procedure in case different welding positions would be needed due to the size increase and/or geometry change of the blanket modules. 4. Thermal-hydraulic performances The prediction and validation of the helium thermal-hydraulic flow balance in the TBM is of major importance because it is directly linked to the prediction of the integrity of steel structures, the performance of functional materials in particular tritium release from breeder/multiplier materials and the parasitic tritium permeation through structures towards the helium coolant circuit. The capacity of numerical simulation is today too limited to represent the complex multi-stage collector system of the TBM and its whole set of cooled structures. The GRICAMAN experimental program has been launched to investigate the hydraulic flow balance and pressure drop in particular in the TBM grid/collector system [6]. It is based on dynamic similitude between the 8 MPa, 370 ◦ C helium flow model in the TBM collectors system and a 0.3 MPa, 25 ◦ C air experiment at identical scale. A limit in the similitude approach obviously exists for the flow in cooling structures (grid plates, caps) where non-adiabatic conditions are not represented in the air experiment. However the experiment is focused on the validation of the flow in collector systems where the helium flow can be considered, in a first approximation, as adiabatic. The preliminary air experiment in grid/cap plates is first intended to define a hydraulic (adiabatic) equivalent, e.g. a slide valve, for each single grid/cap plate. Assuming that all grid/cap plates will extract the same heat flux in the helium model, the deviation of the air model is assumed to be uniform in each equivalent grid/cap plate and thus, it shall not impact the relative effect of flow balance in the manifold systems. In the final experiment, two manifold plenums will be connected together by a set of parallel valves representing the grid/cap plates. The program is still on-going and preliminary results have shown that: - The flow balance between channels of a grid plate is very sensitive to the design of the local inlet collector. Relative flow discrepancies up to 24% were observe in STAR-CD numerical simulation of the plate which can be reduced down to ±4% by re-design of the local collector. - The discrepancy between computational and experimental mass flows in plate channels is ranging between 4 and 8%. The expected outcome of the GRICAMAN program is a validation of the collectors design (local channels collectors in plates and global plates collectors at the rear of the TBM) as well as a recommendation for a security coefficient to be applied in design studies for a potential discrepancy between mass flows in various cooling plates and/or channels of the TBM structure. Preliminary trends, summing-up all sources of uncertainties and remaining discrepancies after design optimization, would indicate that up to ∼17% flow scattering could remain between channels of a grid plate. The helium flow balance in the first wall and the prediction of convective heat transfer coefficients in channels facing the plasma side remain one of the major issues because the first wall is the component exposed to the highest thermal loading. The helium flow balance in first wall shall be addressed by a specific experimental program in continuation to GRICAMAN. For the predictability of the convective heat transfer coefficient, the European approach consists in the followings:

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- Presently, thermal-hydraulic calculations are performed with the Gnielinski correlation for the determination of the convective transfer coefficient. The application of this correlation is however limited to a uniform heat flux loading around cooling channels, which is actually deviating from the heterogeneous repartition expected in first wall channels. - A specific experimental program, HETRA, is on-going in Europe for establishing correction factors corresponding to the actual TBM thermal flux configuration, as well as recommended safety margins. 5. Pebble bed thermo-mechanical performances In HCPB blankets, tritium breeder (lithium orthosilicate Li4 SiO4 or lithium metatitanate Li2 TiO3 ) and neutron multiplier (beryllium) are implemented under the form of pebble bed layers. These materials are subjected to volumetric power deposition and damages under neutron irradiation. The heat deposited is extracted by cooling plates forming also the mechanical boundaries of the pebble beds. The prediction and validation of the thermal–mechanical behavior of the pebble bed is of major importance for the design of the BB modules and TBM, in particular for the verification of the following engineering parameters: - maximum temperature in lithium ceramic (<920 ◦ C); - maximum temperature in beryllium (<650 ◦ C) and maximum beryllium/EUROFER contact temperature (<550 ◦ C); - temperature profile of ceramic/beryllium for prediction and control of tritium release; - stress on pebbles for avoiding pebble fracture, dust formation and change in thermal conductivity of the bed; - effort on walls (cooling/stiffening plates) and verification of wall/bed thermal contact. The prediction of thermal–mechanical behavior of pebble beds is complex due to their discrete nature. In particular the thermal conductivity of the pebble beds depends, besides the nature and pressure of the purge gas, on the compressive strain of the bed induced by thermal expansion of pebbles during operation. When the thermal conductivity of the packed bed varies, it affects in turn the strain/stress distribution by changing the temperature distribution. Also, swelling of pebbles under irradiation can affect the thermal conductivity and the stress distribution in the bed. The mechanical behavior of a pebble bed is also in itself complex because it mixes a non-linear stress dependent elastic behavior, in particular during thermal unloading phases, with plastic and creep behaviors. Modeling approaches developed in Europe are based on a continuum approach where the discrete materials are modeled as continuous, homogeneous and isotropic materials. This continuum approach is here justified by the fact that the ratio between the pebbles diameter (∅ 0.25–0.63 mm for ceramic; ∅ 1 mm for beryllium) and the bed dimensions (min. thickness of ceramic beds in TBM/BB: 29/11 mm; 30/27 mm for beryllium beds) is high enough, ranging from 17 to 116. It shall be noted that this continuum approach is suitable for assessing most of the engineering parameters mentioned above, except the local stress on pebbles for which a local discrete element model could be used or an experimental verification reproducing the maximum hydrostatic pressure computed in a continuum model. - FZK has developed [7] a 2-dimensional thermo-mechanical model in ABAQUS including a non-linear elasticity law expressing the dependency, at power 2/3, of the Young’s modulus to the bed stress state represented by the von Mises stress and the

hydrostatic pressure. The plastic behavior of the bed is described through the modified Drucker-Prager-Cap model predicting the yielding and hardening behaviors. The time-dependent behavior or thermal creep is modeled by a so-called consolidation (cap) creep mechanism. The strain dependent thermal conductivity and thermal expansion are obtained in prior experiments and then implemented in modeling through empirical equations. The so-developed global model makes a full and non-linear thermomechanical 2-way coupling excluding the classical staggered approach solving first the thermal boundary values problem and then computing mechanical equilibrium of the obtained thermal strains. - Alternative models were also developed in DIN and NRG. For instance, DIN has implemented [8] a similar but 3-dimensional approach using also the ABAQUS code but with a different nonlinear elasticity law expressing the dependency of the Young’s modulus of the pressure state through a logarithmic law. The creep model is not implemented yet. Several engineering experiments [9] have been carried-out to simulate, and then compare with computational results, the thermo-mechanical behavior of horizontal ceramic/beryllium pebble beds embedded within cooling plates: HELICA simulates a 13.8 mm thick Li4 SiO4 pebbles bed (∅ pebbles: 0.2–0.4 mm; helium cooling 200–250 ◦ C) and HEXCALIBER (Fig. 3) simulates a stack assembly of 16 mm thick Li4 SiO4 pebbles bed (∅ pebbles: 0.2–0.4 mm) and 56 mm thick beryllium pebbles bed (∅ pebbles: 1 mm). In these experiments the volumetric power deposition expected in BB modules and TBM is simulated by plate electrical heaters (2 plates per pebble bed layer) reaching a maximum relevant power within typically 6 h by six successive one-hour steps then followed by an unloading phase. The cycle is then repeated several times. Several publications are devoted to the detailed analysis of these experiments and comparison with computational analyses [7,8]. The main conclusions that can be derived out of the global European effort are the following: - In general, a good accuracy is obtained between measured and calculated temperatures with a maximum discrepancy of about ±10%. - The order of magnitude of the peak hydrostatic pressure and von Mises stress in ceramic/beryllium beds is calculated to be of few MPa (typically ∼1.5–5 MPa). However it was demonstrated that the presence of plate heaters tends to increase these values due to difference in thermal expansion between heaters and cooling plates and friction forces associated to heater/bed mechanical contact. The presence of heaters tends actually to shear the bed

Fig. 3. HCPB HEXCALIBER mock-up for engineering/code validation of pebble beds performances.

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(increasing the von Mises stress) and push them towards the edge of the bed (leading to a peak in hydrostatic pressure). Actual values in TBMs and BB modules shall be lower. - The above remark leads also to conclude that the insertion of heaters in TBM during a testing phase without neutrons/volumetric power deposition (H-H plasma) shall not be of additional interest compared to a dedicated experiment out-ofITER like HEXCALIBER. On the other hand, it also demonstrates the necessity for achieving a sound DEMO BB design to perform validation tests in ITER nuclear phases. - Creep phenomena are predicted to occur in the beds, being induced by thermal expansion of pebbles and reaching a peak at maximum temperature. This prediction was also indirectly confirmed in experiments through the strain dependent nature of the thermal conductivity model and the final good agreement between computed and measured temperature. The peak volumetric creep strain is predicted to be ∼3.8% in ceramic beds. Computed analyses have also shown how the hydrostatic pressure is simultaneously affected (i.e. decreased) by the pebbles re-arrangement during creep. The outcome is that the implementation of a creep model for the design of TBM and BB modules is essential to represent accurately the stress/strain state and, consequently the thermal control. 6. PbLi MHD flow performances In the HCLL blanket, the PbLi circulates for tritium extraction outside the blanket. Inside the blanket, the PbLi flows in straight radial ducts of rectangular section delimited by stiffening/cooling plates and three-dimensional expansions/contractions in inlet/outlet manifolds or in areas connecting ducts with opposite flow direction. The particularity of the MHD flow in straight rectangular ducts is that, although they are hydraulically separated from each other, their dividing walls (e.g. stiffening/cooling plates) are electrically conducting and current across these walls can couple neighboring channels through exchange of electrical power. The needed PbLi recirculation rate is rather small and leads to average velocities in rectangular duct of typically 1–1.5 mm/s for 10 recirculations per day. At such low velocities and for typical HCLL blanket conditions (TPbLi = 550◦ , B = 5 T, half toroidal width of a duct = 0.09 m), it is found that the Hartmann number (Ha) and the interaction parameter (N) reach very high values, Ha = 1.1 × 104 , N = 1.8 × 105 , so that viscous and inertia forces are of minor importance compared to electromagnetic forces in most of the flow domain. Not only the prediction of the pressure drop in the blanket is of importance – it is an engineering issue linked to the design of the circulation pump and the optimization of global DEMO reactor efficiency – but also the precise prediction of the PbLi MHD flows, in the core of the flow as well as locally along duct boundaries, is essential. It is indeed linked to the following design issues: - in-blanket PbLi flow dead zones or recirculation loops have to be avoided for minimizing the inventory of tritium and activated corrosion products that are to be extracted in the external PbLi extraction/purification system; - tritium permeation from PbLi to cooling structures (e.g. cooling/stiffening plates) is influenced by the PbLi velocity and the coupled tritium concentration profiles along these structures (see Section 8); - corrosion of EUROFER steel structures by PbLi depends on the local PbLi velocity and temperature at interface. FZK has developed a 3D modeling of MHD flows in coupled domain with the CFX software extended through user defined

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Fig. 4. HCLL MHD mock-up for testing in MEKKA facility (FZK).

subroutines to take into account the momentum equation for the Lorentz force, the Ohm’s law and the conservation of charge combined to a Poisson equation for electrical potential [10]. The latter is a diffusion type equation which is solved using an analogy between temperature and electrical potential and the CFX heat equation solver in the non-convective regime. For fully established 3D flow conditions, good results can be obtained up to ∼1000 Ha. For other 3D flows, the limit is closer to 500, limited by computational resources. Besides numerical simulation, a theoretical approach, the asymptotic theory, is also possible for predicting flows but is limited to single duct configuration and regimes where N  1. In total, the fluid domains in the complicated three-dimensional structure of the module with a large number of cooling plates, expansions and contractions are far from being resolved with present numerical/theoretical capabilities. Some improvement areas exist in particular with the implementation of a boundary layer model and multi-processors parallel calculation. At present, in order to confirm and complement the first theoretical/numerical predictions, an experimental campaign has been carried-out in which the MHD flow in a prototypical mock-up (Fig. 4) of the HCLL blanket is investigated in the liquid metal MHD laboratory MEKKA of FZK [11]. It allowed validating preliminary numerical simulations available in some parts of the flow path and, more generally, defining scaling laws needed for the development of a sound TBM design for ITER. The main conclusions can be summarized today as follows: - The total expected pressure drop in the TBM has been extrapolated from experimental results and is ∼0.7 MPa for a magnetic field of 6 T and an average velocity in breeder units of 1 mm/s. This pressure drop occurs at ∼98% in draining/feeding pipes of the breeder units and poloidal manifolds. In a DEMO BB module, the pressure drop will further increase mainly due to longer poloidal pipes and higher velocity in these pipes. - A comparison between measured and calculated electric potentials has shown a good agreement for, typically, Ha = 3000 and Re varying in the range 600–1200. - For a fully toroidal magnetic field, the current induced in the core of the channels formed by cooling plates crosses these plates perpendicularly and induces a strong electrical coupling between co-current flow channels. As a result, the flow is characterized by a uniform core velocity and a slight increase at the wall of the cooling plates. This slight increase is even a little bit more pronounced at the junction between the cooling plate and the side walls, the latter being perpendicular to the magnetic field (“Hartmann wall”) and inducing local current loops closing in Hartmann PbLi layers and walls, affecting locally the flow velocity. In the stiffening plates separating counter-current flows of two breeder units, the current streamlines are running tangentially through the plates and a weak electrical coupling between

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counter-current flows is actually seen. As a result the PbLi velocity increases, by less than a factor 2, compared to the electrical coupled channels. - Considering that channels formed by cooling plates are strongly electrically coupled, the number or even the presence of cooling plates has only a limited impact on the flow distribution in the core of the flow. About the same velocities would be obtained without cooling plates. - 2D numerical simulations have also shown that when a poloidal component is added to the toroidal magnetic field (e.g. 10% of the toroidal field intensity like expected typically in ITER for Q = 10 after full establishment of the plasma current) then the current path and PbLi velocities are slightly affected. Typically, although the overall structure of velocity field described above remains, it becomes slightly unbalanced on the two sides of the stiffening plates. On one side velocities are enhanced, on the other side they are reduced or even locally reversed near Hartmann walls (side walls). As a result there are local reverse flows in some corners near the stiffening plate. Although the MEKKA experimental campaigns allow defining scaling laws for TBM engineering parameters like pressure drop and conditions for appearance of undesired flow regimes like internal recirculation loops, it remains that some key test conditions cannot be obtained outside a tokamak environment like for instance the radial dependency of the magnetic field intensity or Ha >∼6000. The detailed analysis of TBM MHD test data will require further development in numerical simulation which shall also go in the direction of coupling with tritium permeation. 7. Tritium permeation and control The knowledge and control of the tritium permeation through EUROFER cooling structures into the helium coolant is of importance for keeping the size of the helium coolant purification system (CPS) within reasonable technological limits in a DEMO reactor. Also safety limits on tritium release from the secondary coolant circuit require the monitoring and control of the tritium permeation from the helium coolant through the structures of the heat exchanger (e.g. Incoloy, Inconel). In the HCLL breeding blanket, tritium is produced in the moving liquid PbLi and transported by diffusion and advection. The PbLi flow develops boundary layers (thermal, fluid and concentration) at cooling/stiffening plates which act as a mass transfer resistance and have an impact on the tritium permeation towards the He circuit, increasing the effective permeation reduction factor. The thickness of the concentration boundary layer is generally dependant on the velocity profile, which is the result of the Lorentz force (MHD interaction), the driven pressure pumping (forced convection) and possible buoyancy convection effects due to high thermal gradient. The effect of the velocity profile on tritium permeation through cooling plates has therefore been assessed using a local tritium transfer model based on the resolution of a MHD problem, a heat transfer and a mass transfer balances on a simple but realistic representation of the geometry of blanket [12]. Global engineering parameters characterizing the tritium transfer within the breeding blanket have been defined, namely the permeation through the helium circuit, the mean outlet tritium concentration in Pb–15.7Li and the ratio between the permeation through the helium circuit and the production rate. Using a factorial design analysis, it is shown that, for the mid equatorial blanket, those outputs are insensitive to the intensity of the magnetic field and the buoyancy effects foreseeing a similar behavior of inboard and outboard modules in terms of tritium transfer towards coolant.

Local sensitivity analyses have also been carried on the blanket system using an isothermal model to assess the sensitivity of the operating parameters and physical characteristics on the engineering outputs previously defined. Normalized local sensitivity coefficients have been computed to assess the percentage of change of an output variable caused by 1% change of an input parameter. It appears that operating parameters are the most sensitive parameters and particularly the temperature. Regarding the physical parameters, it has been shown that the sensitivity coefficient ranking is not dependent on the value of the Sievert constant. Results also demonstrate that global outputs are almost insensitive to the kinematics viscosity and the electrical conductivity of PbLi. The Eurofer diffusivity, the Eurofer tritium solubility and the PbLi tritium solubility have almost the same normalized sensitivity coefficients. It is also worth noting that the tritium diffusivity in PbLi has a normalized sensitivity coefficient of the same magnitude impact as other mass transfer thermophysical characteristics although this parameter is not taken into account in most of the system tritium transfer analyses. These analyses have shown that the concentration boundary layer is to be regarded as an equivalent permeation reduction factor (PRF) of 30. Considering the difficulty encountered in the past years to obtain stable and reproducible high PRF with Al-based coating on Eurofer, such a result is of importance for the control of the tritium inventory in the HCLL blanket and should relax the specification on the helium coolant purification system. Although the use of tritium permeation barriers (TPB) might not be required in TBM systems, development in this area is justified by the need to get the full picture of the tritium cycle and control in a DEMO reactor. Also, the TPB could be tested in TBMs when appropriate conditions are met in particular during long/back-toback pulses. In the past years, technologies for the realization of TBP were developed, mainly by deposition of ceramic materials on cooling structures. However, these technologies were not found mature yet [13,14]. In additional, complexity of the shapes to be coated, interaction of the TPB with the breeder materials and compatibility of the coating method with the fabrication sequence of the module box have justified the investigation of alternative methods. An easier solution was investigated consisting in nucleation and growth of natural oxides on the surface of the cooling structures that is in contact with helium [15]. Experiments were conducted on EUROFER, Incoloy and Inconel disk samples exposed on one of their side to a controlled H2 /H2 O addition to the gas stream. In order to avoid parasitic effects like outgassing from the apparatus walls, the continuous flow method was preferred. Results on EUROFER samples showed that a permeation reduction factor (PRF) of ∼30 is observed at the centre of an H2 /H2 O molar ratio of 60/3-85/3. Further and finer examination might even reveal a higher peak value in this interval. It was observed that the PRF effect is obtained in the first minutes of the oxidation and that constant permeation rates are measured for long oxidation time. Micrographs revealed that, at the maximum obtained PRF, a ∼5 ␮m thick and compact oxide layer covers a large fraction of the surface. However this coverage is not fully uniform and is concentrated on crests of the surface (this could be due to a higher energy potential remaining in these areas caused by residual stress from sheet production and creating favorable conditions for oxide nucleation). It can be assumed that the limiting factor of the PRF is the coverage of the specimen surface by the oxide layer. Incoloy and Inconel specimens submitted to similar oxidizing atmosphere gave no significant variation of the permeation rate when compared to a non oxidizing test atmosphere. This is probably due to the natural protective oxide formed on nickel alloys before testing. The PRF was then estimated by comparing to the literature data on unoxidized Incoloy 800 and was found to be in

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the range of 1000–10,000. The same can reasonably be assumed for Inconel. 8. Conclusion The R&D effort carried out by Europe over the last decade has allowed accumulating sufficient modeling capability and phenomenological or engineering data to build a first image of what should be a sound breeder blanket for DEMO. It appears also now clearer that the TBM design and its technology objectives are attainable over the construction time of ITER, in particular in front of regulatory nuclear requirements. On the other side, the detail analysis of the TBM test data from ITER, and in particular the ones resulting from coupled phenomena in the blanket or more complex tokamak loading conditions, require further and significant development effort of modeling tools, in particular in the field of MHD, tritium permeation and cycle modeling, tritium release from solid breeder/multiplier, pebble bed thermo-mechanics. Likely the integration of instrumentation in ITER TBMs will also be limited and mostly engineering TBM data will be available for analysis and interpretation. It is then essential to keep validating modeling tools out-of-ITER in order to provide an accurate prediction of the physical phenomena in the TBM so that engineering data can be interpreted properly. Acknowledgement This work, supported by the Euratom Communities under the contracts of association between EURATOM-CEA, EURATOM-FZK and EURATOM-ENEA was carried out within the framework of the European Fusion Development Agreement and of Fusion for Energy. The views and opinions expressed herein do not necessarily reflect those of the European Commission.

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References [1] L.V. Boccaccini, J.-F. Salavy, H. Neuberger, I. Ricapito, P. Sardain, et al., Fusion Eng. Design 84 (2009) 333–337. [2] M. Zmitko, Y. Poitevin, L.V. Boccaccini, J.-F. Salavy, R. Knitter, A. Moeslang, et al., Development and qualification of functional materials for the EU Test Blanket Modules: strategy and R&D achievements, in: Proceedings of the 14th International Conference on Fusion Reactors (ICFRM-14), September 6–11, 2009, J. Nucl. Mater., in press. [3] I. Ricapito, O. Bede, L.V. Boccaccini, A. Ciampichetti, B. Ghidersa, L. Guerrini, et al., The ancillary systems of the European Test Blanket Modules: configuration and integration in ITER, in: Proceedings of the 9th International Symposium on Fusion Nuclear Technology (ISFNT-9), October 11–16, 2009, Fusion Eng. Design, in press. [4] R. Pascal, B. Calcagno, First review of the EU TBMs design and manufacturing with respect to ITER Codes & Standard – ESC06-002 Del. 2, ASSYTEM report ASV25540A, 2008. [5] Y. Poitevin, Ph. Aubert, E. Diegele, G. de Dinechin, J. Rey, M. Rieth, et al., Breeder blanket manufacturing technologies, in: Proceedings of the 14th International Conference on Fusion Reactors (ICFRM-14), September 6–11, 2009, J. Nucl. Mater., in press. [6] M. Ilic, B. Kiss, T. Ihli, Fusion Eng. Design 83 (2008) 1253–1257. [7] Y. Gan, M. Kamlah, J. Nucl. Mater. 386–388 (2009) 1060–1064. [8] P.A. Di Maio, G. Dell’Orco, R. Giammusso, A. Malavasi, I. Ricapito, A. Tincani, et al., Fusion Eng. Design 83 (2008) 1287–1293. [9] G. Dell’Orco, A. Ancona, P.A. Di Maio, M. Simoncini, G. Vella, Out-of-pile thermo-mechanical testing of breeder pebble beds for HCPB TBM for ITER, in: Proceedings of CBB-I 13, Santa Barbara, USA, December 2005. [10] L. Bühler, Fusion Eng. Design 75–79 (2005) 923–926. [11] L. Bühler, H.-J. Brinkmann, S. Horanyi, K. Starke, Magnetohydrodynamic flow in a mock-up of a HCLL blanket. Part II: experiments, FZK EFDA task report TW5-TTBC-006-D1, 2008. [12] F. Gabriel, Simulation of tritium permeation, CEA EFDA task report TW5-TTBC005-D5, 2008. [13] A. Aiello, et al., Fusion Sci. Technol. 41 (2002) 872–876. [14] I. Ricapito, Tritium management in the HCLL-DEMO blanket, ENEA report LBG-R-008, 2004. [15] A. Aiello, M. Utili, Experimental study of the efficiency of oxide layers in Helium Coolant System (HCS) for reduction of H permeation through EUROFER and heat exchanger materials (Incoloy, Inconel), ENEA EFDA task report TW5-TTBC-005D4, 2008.