Comprehensive nuclear materials 4 17 tungsten as a plasma facing material

Comprehensive nuclear materials 4 17 tungsten as a plasma facing material Comprehensive nuclear materials 4 17 tungsten as a plasma facing material Comprehensive nuclear materials 4 17 tungsten as a plasma facing material Comprehensive nuclear materials 4 17 tungsten as a plasma facing material Comprehensive nuclear materials 4 17 tungsten as a plasma facing material Comprehensive nuclear materials 4 17 tungsten as a plasma facing material Comprehensive nuclear materials 4 17 tungsten as a plasma facing material

4.17 Tungsten as a Plasma-Facing Material G Pintsuk Forschungszentrum Juălich, Juălich, Germany ò 2012 Elsevier Ltd All rights reserved 4.17.1 4.17.2 4.17.3 4.17.3.1 4.17.3.2 4.17.3.2.1 4.17.3.2.2 4.17.3.2.3 4.17.3.2.4 4.17.3.2.5 4.17.3.2.6 4.17.3.2.7 4.17.3.3 4.17.4 4.17.4.1 4.17.4.1.1 4.17.4.1.2 4.17.4.1.3 4.17.4.1.4 4.17.4.1.5 4.17.4.1.6 4.17.4.2 4.17.4.2.1 4.17.4.2.2 4.17.4.3 4.17.4.3.1 4.17.4.3.2 4.17.4.3.3 4.17.4.3.4 4.17.4.4 4.17.4.4.1 4.17.4.4.2 4.17.4.4.3 4.17.5 References Introduction Functional Requirements Material Selection Fabrication and Microstructure Advantages and Limitations for Fusion Application High atomic number: material erosion/melting Recrystallization Machinability, mechanical properties, and DBTT Component fabrication: CTE mismatch with heat sink Neutron embrittlement Neutron activation and radiological hazards Material availability Tungsten Grades Influence of In-Service Conditions Thermal Shock Resistance Microstructure, composition, and mechanical properties Power density and pulse duration Base temperature Repetition rate Thermal shock during off-normal events: disruptions Thermal shock during normal operation: ELMs Thermal Fatigue Resistance ITER Prototype and commercial reactors Neutron Irradiation Thermophysical properties and swelling Mechanical properties Thermal shock on irradiated W Thermal fatigue on irradiated W components Ion Irradiation and Retention He-irradiation Hydrogen-irradiation and retention Combined loading conditions Conclusion Abbreviations APS AUG CFC CTE CVD DBTT DEMO Atmospheric plasma spraying ASDEX-upgrade Carbon fiber composite Coefficient of thermal expansion Chemical vapor deposition Ductile to brittle transition temperature Demonstration fusion reactor ECAE ECAP ELMs fpy FTU ICRH IFE 552 553 554 554 555 556 556 556 557 557 558 558 559 561 561 561 562 562 563 563 564 566 566 567 568 568 569 569 570 570 570 571 575 576 576 Equal-channel angular extrusion Equal-channel angular pressure Edge localized modes Full power years Frascati tokamak upgrade (Frascati, Italy) Ion cyclotron resonance heating Inertial Fusion Experiment 551 552 Tungsten as a Plasma-Facing Material IFMIF International Fusion Materials Irradiation Facility ITER Tokamak, Latin for ‘the way’ JET Joint European Torus (Culham, UK) LPPS Low-pressure plasma spraying MIM Metal injection molding NIF National Ignition Facility (Livermore, CA, USA) PFC Plasma-facing component PFM Plasma-facing material PS Plasma spraying PVD Physical vapor deposition SC Single crystal SPS Spark plasma sintering TEXTOR Tokamak EXperiment for Technology Oriented Research (Juălich, Germany) TZM TiZrMo VPS Vacuum plasma spraying Symbols cp Tm l r The specific heat Melting temperature Thermal conductivity Density 4.17.1 Introduction Until the mid-1990s, only few fusion devices used high-Z elements in plasma-facing materials (PFMs).1 These devices either operated at high plasma currents and high plasma densities such as Alcator C-Mod2 and Frascati tokamak upgrade (FTU)3,4 or used high-Z materials only as test limiters such as Tokamak EXperiment for Technology Oriented Research (TEXTOR).5–9 Since then, high Z refractory metals have been attracting growing interest as candidates for PFMs because of their resistance against erosion and the need for low erosion and stability against neutron irradiation.10 Considerable effort has been made to study the behavior of high Z impurities in the core and edge plasmas, erosion/redeposition processes at the limiter/divertor surfaces, hydrogen isotope retention, and on material development and testing In particular, the modification of ASDEX-upgrade (AUG) into a fully tungsten machine,11–17 which was achieved in 2007, provided positive answers to critical questions on the reliability of tokamak operation with high-Z plasma-facing components (PFCs) and the compatibility with standard and advanced H-mode scenarios and with the available heating methods.10 Among the challenges, for tokamak devices, that still remain are the strong increase of the W source and W concentration resulting from ion cyclotron resonance heating (ICRH) and the need for rigorous modeling to support the extrapolation of current results to ITER conditions Clearly, not all questions posed by ITER can be answered by AUG only For example, the effects of material mixing with Be, the melt behavior under transients, or the change of the hydrogen retention due to damage by high-energy neutron irradiation18 cannot be addressed in AUG Answers to some of these issues may be provided by the ITERlike wall project in Joint European Torus (JET), which is installing a bulk tungsten component for the strike point and physical vapor deposition (PVD)-W-coated carbon fiber composite (CFC) tiles for the remaining parts of the divertor.19–21 The remaining questions have to be answered by dedicated experiments in other plasma devices or can only be assessed by modeling However, the results obtained so far not exclude the use of W in ITER as a standard PFM.10 Further investigations related to future fusion power plants such as demonstration fusion reactor (DEMO) have to focus on the minimization of plasma heat loads to the PFCs to increase their lifetime In particular, transient heat loads caused by instabilities significantly decrease the operation domain of PFCs, due to thermal stresses and consequent enhanced erosion.22 Therefore, it is also important to mitigate all instabilities, such as edge localized modes (ELMs), that cause significant plasma transient heat losses.23 Plasma scenarios need to be developed, such that the conditions for achieving the required fusion yield are maintained in steady state, while at the same time sustaining tolerable heat loads on the PFCs The above-mentioned upgrades to the JET24 and AUG15 will allow further optimization of the plasma scenarios under these conditions, in particular with DEMO relevant tungsten PFCs.25 These investigations will show how the identified deficiencies of W can be overcome or how they have to be dealt with In addition to the application of tungsten in ITER and in potential future tokamak devices such as DEMO,26–29 tungsten also became an interesting alternative for the divertor of stellarators, for example, Advanced Reactor Innovation Evaluation Studies – Compact Stellerator (ARIES-CS),30 and as a first Tungsten as a Plasma-Facing Material wall material for inertial fusion devices.31 Due to similar demands on the PFMs during the operation of all these devices, similar problems have to be solved for each application 4.17.2 Functional Requirements In the current design of the ITER divertor32–34 for the start-up phase, tungsten has been selected as armor for the divertor dome and the upper part of the divertor vertical targets In addition, due to excessive co-deposition of tritium in CFC raising regulatory concerns related to tritium inventory limits, a full tungsten divertor will be installed before the D–T phase of operation.32 The PFC design for ITER consists of bulk W bonded to an actively pressurized water-cooled Cu alloy heat sink Here W has no primary structural function However, due to the operating conditions listed in Table 1, the PFMs face large mechanical loads particularly at the interface to the heat sink material during cyclic steady state heat loads (see Section 4.17.4.2) and at the plasma-loaded surface during transient thermal events (see Section 4.17.4.1) Furthermore, the material response to these loads is influenced by the material damage or degradation due to neutron irradiation (see Table 1, Sections 4.17.4.3.3 and 4.17.4.3.4) Table 553 Along with thermally induced loads, the interaction of the PFM with the plasma, that is, the hydrogen isotopes D and T as well as the fusion product He, is of importance (see Section 4.17.4.4) because they have an influence on material erosion and nearsurface material degradation The further development of the ITER design led to four conceptual designs for the DEMO divertor.25,35 These designs include either water (inlet 140  C/outlet 170  C) or, due to the higher achievable efficiency, more probably He-cooling (inlet 540  C/outlet 700  C) In all cases bulk W is foreseen as the armor material that will have to face peak steady state heat loads of 15 MW mÀ2 in case of the water-cooled design and 10 MW mÀ2 for the He-cooled designs In contrast to ITER, off-normal events such as disruptions have to be avoided completely and transient thermal events during normal operation, for example, ELMs, have to be mitigated below the damage threshold of the material (see Section 4.17.4.1) This may be particularly important considering the expected neutron damage that will amount up to 40–60 dpa during the planned operation of the fusion reactor35 leading to a significant amount of transmutation products.36 However, the main limiting factor is expected to be the material’s erosion leading to a maximum lifetime of years for the divertor armor.35 Operating conditions assumed for the design of the ITER PFCs during D–T operation Material Number of replacements Baking temperature ( C) Normal operation Lifetime (number of cycles) Peak surface heat flux (MW mÀ2) Peak particle flux (1023 mÀ2 sÀ1) ELM energy density (MJ mÀ2) controlled/uncontrolled ELM duration (ms) ELM frequency (Hz) controlled/uncontrolled Maximum radiation damage (dpa) Operation temperature design window during normal operation ( C) Off normal operation: disruptions Peak surface heat load (MJ mÀ2) Duration (ms) rise time/decay time Frequency (%) Divertor target Divertor baffle/dome CFC/W 240 W 240 3000–10 000 $10a $10 0.3–0.5/6–10 0.25–0.5 20–40/1–2 0.7b 200–1000 3000–10 000 1600  C, and good machinability.68,77,78,157 W–Si–Cr as a ternary or even by the addition of another element as a quarternary alloy is a Tungsten as a Plasma-Facing Material operation but also understanding the manufacturing accuracy and reproducibility because tens of thousands of armor/heat sink joints will be produced Studies on this issue have shown that the current W monoblock design with a defect extension up to 50 appears to be suitable for the upper part of the vertical target (P ¼ 10 MW mÀ2), but is not well adapted to a heat flux of 20 MW mÀ2, which is necessary for application at the strike point of the vertical target, as systematic defect propagation was observed A tungsten flat tile design with 6-mm long defects in the material interface was studied and proved to be compatible with fluxes of MW mÀ2 but was unable to sustain cyclic fluxes of 10 MW mÀ2.212 These results confirm that the monoblock geometry generally proves to have superior behavior under high heat flux testing when compared with flat tile geometry However, it is worthwhile to continue the investigation of the flat W tile design for low-flux regions despite the hazard of cascade failure of the flat tiles106 for two reasons: cost and weight Besides this characterization, a number of high heat flux tests have been carried out on mock-ups and prototypes without artificial defects representing the different design options to assess the ‘fitness for purpose’ of the developed technologies.33,90,161,213–218 The results obtained for small test mock-ups of the flat-tile and monoblock design can be transferred to large-scale prototypes for the divertor vertical target Independent of the type of pure Wor W–La2O3 armor material used in these prototypes, the W parts survived in the nonneutron-irradiated condition up to 1000 cycles at 20 MW mÀ2 in the monoblock design213,219 and up to 1000 cycles at 18 MW mÀ2 in the flat tile 567 design (see Figure 7).213 This is far beyond the design requirements for use in the upper part of the vertical target (P ¼ MW mÀ2) and, in case of the monoblock design, even meets the design requirements for the strike point area of the vertical target Alternative concepts such as explosive bonding of tungsten to a heat sink material,220 PS on a Cu-alloy216 or on EUROFER steel44 could probably be of use in the divertor but even more for first wall applications for fusion machines beyond ITER However, these concepts often suffer from high interfacial stresses as a result of the CTE difference between the W coating and the substrate 4.17.4.2.2 Prototype and commercial reactors There are many design proposals for a He-cooled first wall and divertor concept for DEMO and ARIES-CS.37,160,221 Among these, the He-cooled modular design with jet cooling (HEMJ)102 is the most developed and qualified in terms of microstructural response,103 having survived at reduced coolant temperatures of 450–550  C at least 100 cycles at 11 MW mÀ2 without failure In contrast to the results obtained for water-cooled components for ITER, no influence of grain orientation on the components performance was observed.102 This might be a result of the higher temperature, which was always above the DBTT Nevertheless, some difficulties in the design still have to be resolved First, there are problems related to temperature with a desired coolant temperature of !600  C; these include material recrystallization at the top surface and the necessary high temperature joining to the W-based heat sink material Second, W macrobrush dpa: 1000 cycles at 18 MW m−2 0.6 dpa: 1000 cycles at 10 MW m−2 (increasing of Tsurf) W monoblock dpa: 1000 cycles at 20 MW m−2 0.6 dpa: 1000 cycles at 18 MW m−2 (no degradation) Figure Thermal fatigue testing results of W macrobrush and W monoblock mock-ups before and after neutron irradiation Tungsten as a Plasma-Facing Material issues related to the material’s mechanical properties must be solved, in particular for the ductility of W-based structural material and its neutron irradiation resistance (see Section 4.17.4.3) Finally, the manufacturing reproducibility has to be at a high level because of the large number of small units (1 unit % Â 10À4 m2) necessary for cladding the DEMO divertor 4.17.4.3 Neutron Irradiation The irradiation of tungsten and tungsten alloys with energetic neutrons (14 MeV) resulting from the D–T reaction causes radiological hazards that were already discussed in Section 4.17.3.2.6 In addition, the neutron irradiation affects the material composition by transmutation of tungsten to Re and subsequently osmium (transmutation of W isotopes to Ta and Hf are negligible222) The amount of transmutation strongly depends on the applied neutron wall load and neutron spectrum223 and for the W to Re transmutation reaction reaches values between 0.3 and at.% per dpa.222 The subsequent transmutation of Re to Os is expected to occur faster than the production of Re from W resulting in a steadily proceeding burnup of Re The neutron fluence on the first wall varies strongly with location For the full lifetime of ITER a maximum of $0.3 MWa mÀ2 is achieved224 (%1.35 dpa in tungsten225) As the divertor PFCs will be exchanged times and only the last three will operate in a D–T environment, a neutron fluence of $0.1 MW mÀ2 is expected during the lifetime of each PFC For DEMO, an average neutron wall load of MW mÀ2 is assumed for the main chamber, which would result in $45 dpa after full power operation years These conditions yield a transmutation of 100% W into 75% W, 12% Re, and 13% Os.36 For geometrical reasons, that is, larger surface to angular extension ratio, it will be roughly a factor less in the divertor region Furthermore, neutron irradiation damages the material properties by the formation of vacancies and interstitials (see Chapter 1.03, Radiation-Induced Effects on Microstructure) Their behavior including analysis of displacement cross-sections,226,227 diffusion, mutual recombination, and clustering are being assessed by atomistic modeling.228–231 Both transmutation and defect generation influences the material properties and subsequently the material response to steady state and transient thermal loads 4.17.4.3.1 Thermophysical properties and swelling The influence of neutron irradiation on the thermophysical properties is related to the irradiation temperature and the number of defects generated in the crystal structure At temperatures 1200  C (see Figure 8) In addition to defect generation, material degradation is also related to the formation of transmutation products such as Re and Os, which in general exhibit poorer thermophysical properties Transmutationinduced degradation increases with increasing temperature and irradiation dose, which makes it the most relevant process for the degradation of material properties for future fusion reactors such as DEMO Despite the potential for full recovery of the material defects mentioned above, void-induced swelling occurs The results235,236 of tungsten and tungsten alloys show that the material’s volume increases with increasing irradiation temperature ( 1050  C).237 W–Re alloys exhibit significantly improved swelling behavior compared to pure W, with a local maximum at $750  C However, the swelling only amounts to 1.7% at 9.5 dpa.237 Therefore, a negligible effect of swelling can be expected for the operation of ITER Experimental values not exist at temperatures >1050  C as expected for the operation of DEMO 70 Thermal diffusivity (mm2 s–1) 568 65 Nonirradiated 0.6 dpa at 200 ЊC 60 55 50 45 40 35 30 200 400 600 800 1000 1200 1400 1600 Temperature (ЊC) Figure Thermal diffusivity of W–1% La2O3 in nonirradiated and irradiated condition Tungsten as a Plasma-Facing Material 4.17.4.3.2 Mechanical properties Data in the literature on mechanical properties of neutron-irradiated tungsten are very limited.234,238,239 However, in combination with experimental results obtained for other refractory metals, it has been shown that in metals with a bcc lattice structure, irradiation hardening causes a steep increase in yield stress and a decrease in ductility.110 Consequently, the material fails by brittle cleavage fracture as soon as the yield stress exceeds the cleavage strength Therefore, the increase of the DBTT depends on the neutron fluence, the neutron spectrum (will be addressed by the International Fusion Materials Irradiation Facility, IFMIF), and the irradiation temperature The radiation hardening in bcc alloys at low temperatures (1000  C would be preferred as full or at least partial recovery of defect-induced material degradation is achieved by annealing at 1200  C.234 This implies that the nearsurface part of a W component will retain its ductility, which has a beneficial effect on the crack resistance at the plasma loaded surface However, such temperatures are not feasible at the interface to the heat sink where tungsten will be in contact with copper (ITER) or steel (DEMO), which are limited to significantly lower operational temperatures Hence, better understanding of the irradiation effects on tungsten at temperatures between 700 and 1000  C is needed, particularly related to reactor application in DEMO.109,110,240 In addition to the influencing factors on the DBTT mentioned above, that is, neutron fluence, neutron spectrum, and irradiation temperature, the material’s composition also plays an important role While the addition of Re has a beneficial effect on the material’s ductility in the nonirradiated state, under neutron irradiation it results in more rapid and severe embrittlement than it is observed for pure W.239 Similarly, less mechanical strength and an increased loss of ductility compared to pure W is found for particlestrengthened W alloys (e.g., W–1% La2O3) when tested up to 700  C The only exception among all explored tungsten alloys might be mechanically alloyed W–TiC (see Section 4.17.3.3) that showed no irradiation hardening as measured by Vickers hardness at 600  C.87 569 Finally, the mechanical properties are influenced by neutron-induced He-generation and the transmutation of tungsten While He generation in W is, compared to CFC and Be, very small ($0.7 appm He per dpa) and its influence on the mechanical properties of W negligible,73,83,224 the transmutation of W into Re and subsequently Os significantly alters the material structure and its properties The generation of significant amounts of ternary a and subsequently s-phases results in extreme material embrittlement and will cause shrinkage In combination with thermally induced strains, this might produce high tensile stresses causing the extremely brittle material to extensively crack and perhaps even crumble to powder.36 4.17.4.3.3 Thermal shock on irradiated W The simulation of disruptions (700 K.45,262 Another material parameter that increases hydrogen retention is the number of dislocations,266,286 particularly those introduced during deformation processes used for material densification However, the recrystallization of the material removes not only dislocations but also vacancies and vacancy clusters, which have been introduced by the impinging H-ions286,287 and as grain boundaries This effectively reduces the trapping sites for hydrogen retention and, consequently, the lowest retention is observed for high-purity SC materials, particularly due to the low diffusion rate compared to polycrystalline tungsten.275,288,289 This low diffusion rate results in a near-surface accumulation of hydrogen, which acts as a diffusion barrier and leads to a saturation of hydrogen retention with increasing fluence.290 Such saturation is not observed for pure polycrystalline tungsten due to the possible hydrogen migration along grain boundaries.291 Finally, the hydrogen retention is influenced by impurities277 and dopants The addition of La2O3 and TiC particles as well as the formation of pores, for example, by potassium doping, not only introduce traps and increase hydrogen retention,293 but also decrease the diffusion rate.291 In contrast, alloying with up to 10% Re has no measurable effect on the H retention properties of the material,279 as it only creates a slightly deformed crystal lattice structure but no additional hydrogen traps In addition to hydrogen retention, material damage and particularly blistering is influenced by the material’s microstructure Blistering occurs preferentially when the crystal is oriented with the h111i direction perpendicular to the surface292 and the blisters develop in different shapes from low, large, and spherical to high, small, and dome or coneshaped.45,262,267 The blisters in recrystallized materials are mainly plateau-shaped, often multilayer structures, which indicate a step-wise build-up, and in few cases also small blisters on top of large ones are formed.263,293 However, it is significant that the blister size is commonly limited by the grain size45,294 indicating that the grain boundaries play an important role in the formation of blisters Accordingly, SCs and nanostructured materials such as W–TiC provide the strongest resistance against blistering, although the particular reason is different For SCs, the hydrogen diffusion and accumulation is limited and there is a fast desorption from lowenergy traps at elevated temperatures In contrast, for nanostructured materials the size of individual grains is extremely small and so is the volume for blister formation Furthermore, the migration of hydrogen is significantly increased by the large number of grain boundaries.292 Further material parameters that reduce blister formation are open porosity and the surface finish, particularly the number of random or artificially introduced scratches that might act similar to grain boundaries.45 In contrast, the introduction of Tungsten as a Plasma-Facing Material impurities and dopants in commercially available grades of tungsten increases the number of blisters and exfoliation in both their stress relieved and recrystallized states.277,293 4.17.4.4.3 Combined loading conditions As described above, the damage mechanisms of hydrogen and He-irradiation are rather similar, although they occur in different temperature ranges Accordingly, their mutual interaction is also strongly influenced by the implantation temperature Therefore, the testing sequence plays a role in the behavior, as for He preirradiation followed by hydrogen implantation, the implantation temperature of He determines the amount and kind of produced damage and the He-retention, which subsequently influences the hydrogen uptake occurring as described in Section 4.17.4.4.2 For example, He-implantation at RT either does not change the retention or may increase it due to the formation of additional trapping sites295–297 and the lower diffusion rate of He compared to H With increasing He implantation temperature up to 800 K, hydrogen retention significantly decreases compared to pure hydrogen irradiation.261,292,296 This may be attributed to the occupation of trap sites by He as a result of its increasing mobility.298 Potential trap sites are the numerous He-induced nanosized bubbles acting as a diffusion barrier.292 A further increase in temperature to 1600 K does create significant material damage by He due to pore and bubble formation or even blistering This tremendously increases the number of trap sites in the material and leads to He desorption during implantation and accordingly increases the hydrogen retention.299 For simultaneous loading of He and hydrogen, the fraction of He should reach at least at.% to observe significant changes in the material’s response.261,292 Furthermore, for implantation temperatures below 900 K, results similar to those described above are observed for sequential ion beam loading.299 However, due to desorption of hydrogen at high temperatures >1000 K, no hydrogen retention takes place and the damage mechanisms are dominated by the He-irradiation during such temperature excursions Correlated with hydrogen retention, blister formation at temperatures