Silicon carbide (SiC)-based ceramic composites have been studied for fusion applications for more than a decade. The potential for these materials have been widely discussed and is now understood to ...be (1) the ability to operate in temperature regimes much higher than for metallic alloys, (2) an inherent low level of long-lived radioisotopes that reduces the radiological burden of the structure, and (3) perceived tolerance against neutron irradiation up to high temperatures. This paper reviews the recent progress in development, characterization, and irradiation effect studies for SiC composites for fusion energy applications. It also makes the case that SiC composites are progressing from the stage of potential viability and proof-of-principle to one where they are ready for system demonstration, i.e., for flow channel inserts in Pb–Li blankets. Finally, remaining general and specific technical issues for SiC composite development for fusion applications are identified.
•We characterize tungsten mono-block components after exposure to ITER relevant heat loads.•We qualify the manufacturing technology, i.e., hot isostatic pressing and hot radial pressing, and repair ...technologies.•We determine the microstructural influences, i.e., rod vs. plate material, on the damage evolution.•Needle like microstructures increase the risk of deep crack formation due to a limited fracture strength.
In order to evaluate the option to start the ITER operation with a full tungsten (W) divertor, high heat flux tests were performed in the electron beam facility FE200, Le Creusot, France. Thereby, in total eight small-scale and three medium-scale monoblock mock-ups produced with different manufacturing technologies and different tungsten grades were exposed to cyclic steady state heat loads. The applied power density ranges from 10 to 20MW/m2 with a maximum of 1000 cycles at each particular loading step. Finally, on a reduced number of tiles, critical heat flux tests in the range of 30MW/m2 were performed.
Besides macroscopic and microscopic images of the loaded surface areas, detailed metallographic analyses were performed in order to characterize the occurring damages, i.e., crack formation, recrystallization, and melting. Thereby, the different joining technologies, i.e., hot radial pressing (HRP) vs. hot isostatic pressing (HIP) of tungsten to the Cu-based cooling tube, were qualified showing a higher stability and reproducibility of the HIP technology also as repair technology. Finally, the material response at the loaded top surface was found to be depending on the material grade, microstructural orientation, and recrystallization state of the material. These damages might be triggered by the application of thermal shock loads during electron beam surface scanning and not by the steady state heat load only. However, the superposition of thermal fatigue loads and thermal shocks as also expected during ELMs in ITER gives a first impression of the possible severe material degradation at the surface during operational scenarios at the divertor strike point.
•Detailed design development plan for the ITER tungsten divertor.•Latest status of the ITER tungsten divertor design.•Brief overview of qualification program for the ITER tungsten divertor and status ...of R&D activity.
In November 2011, the ITER Council has endorsed the recommendation that a period of up to 2 years be set to develop a full-tungsten divertor design and accelerate technology qualification in view of a possible decision to start operation with a divertor having a full-tungsten plasma-facing surface. To ensure a solid foundation for such a decision, a full tungsten divertor design, together with a demonstration of the necessary high performance tungsten monoblock technology should be completed within the required timescale. The status of both the design and technology R&D activity is summarized in this paper.
In order to develop and validate the high performance tungsten monoblock technology, the full-tungsten divertor qualification program was defined. As the first step, small-scale mock-ups were ...manufactured and successfully tested under the required high heat flux loads. The test results demonstrated that the technology is available in Japan and Europe. Post-tests observation of the loaded W monoblocks showed generation of self-castellation – a crack along coolant tube axis. The cause of the self-castellation was discussed and a tungsten material characterization program is being developed with the objective to understand mechanical properties that influence the occurrence of the self-castellation.
In the context of using a full-tungsten (W) divertor for ITER, thermal shock resistance has become even more important as an issue that may potentially influence the long term performance. To address ...this issue a unique series of experiments has been performed on ITER-W monoblock mock ups in three EU high heat flux facilities: GLADIS (neutral beam), JUDITH 2 (electron beam) and Magnum-PSI (plasma beam). This paper discusses the JUDITH 2 experiments. Two different base temperatures, 1200°C and 1500°C, were chosen superimposed by ∼18,000/100,000 transient events (Δt=0.48ms) of 0.2 and 0.6GW/m2.
Results showed a stronger surface deterioration at higher base temperature, quantified by an increase in roughening. This is intensified if the same test is done after preloading (exposure to high temperature without transients), especially at higher base temperature when the material recrystallizes.
•All the tested items sustained the ITER Full W divertor qualification program requirements. This confirms that the technology for the manufacturing of the first set of the ITER Divertor is available ...in Europe.•The surface roughening and local melting of the W surface under high heat flux was proven to be significantly reduced for an armour thickness lower or equal to 6mm.•However, this campaign highlighted some specific areas of improvement to be implemented ideally before the upcoming ITER Divertor IVT serial production.•The issue of the self-castellation of the W monoblocks, which typically appears after a few tenths of cycles at 20MW/m2, is critical because it generates some uncontrolled defects at the amour to heat sink joints. Besides, they create a gap which exposure is almost perpendicular to the magnetic field lines and which might lead to local W melting in the strike point region.•This campaign also evidenced that the minimum IO requirements on the CuCrZr ductility could be revised to avoid the occurrence of rather early fatigue failures. Although the W material characterization program has been set up by the IO, the strategy on the CuCrZr still needs to be defined.
With the aim to assess the option to start the ITER operation with a full tungsten divertor, an R&D program was launched in order to evaluate the performances of tungsten (W) armoured plasma facing components (PFCs) under high heat flux. The F4E program consisted in the manufacturing and high heat flux (HHF) testing of W monoblock mock-ups and medium scale prototypes up to 20MW/m2. During the test campaign, 26 W mock-ups and two medium scale prototypes manufactured by Plansee SE (Austria) and by Ansaldo Nucleare (Italy) have been tested at the FE200 (AREVA, Le Creusot, France) and ITER Divertor Test Facility (IDTF) (Efremov Institute Saint Petersburg, Russian Federation) electron beam test facilities. The high heat flux (HHF) testing program foresaw the performance of 5000 cycles at 10MW/m2 and 300+700 cycles at 20MW/m2, 10s power on and 10s dwell time with ITER relevant hydraulic parameters.
The test results fulfilled the ITER qualification requirements, although a few items did not sustain the extended test program (additional 700 cycles at 20MW/m2), the analysis of the results gave indications on potential improvements, in particular concerning the W material itself with the objective to remove the self-castellation of the W monoblocks and concerning the thermo-mechanical fatigue performances of the CuCrZr heat sink. In addition, some critical heat flux experiments, whose results confirmed those previously obtained were also performed.
Indentation of metals by a flat-ended cylindrical punch Riccardi, B; Montanari, R
Materials science & engineering. A, Structural materials : properties, microstructure and processing,
09/2004, Letnik:
381, Številka:
1
Journal Article
Recenzirano
Odprti dostop
The paper reviews the fundamentals of indentation theory for punches with cylindrical geometry, presents a deep-indentation finite element (FE) simulation and discusses an experimental technique for ...flat-ended cylindrical indentation. This technique is based on the use of cylindrical punches with diameters up to 1
mm and allows pressure–penetration curves to be drawn from which yield stress and elasticity modulus can be determined. Several materials have been tested including pure metals, steels and refractory alloys; yield stress has been determined and compared with literature values. By testing at different temperatures it was also possible to determine the ductile-to-brittle transition temperature (DBTT) for some alloys that show such phenomenon.
Recent Critical Heat Flux (CHF) experimental results in the range of 40 MW/m2 obtained on small scale tungsten (W) monoblock mock-ups with reduced minimum W thickness above the coolant tube compared ...to the nominal 6 mm planned for ITER are presented. The CHF margin during ITER operations is further assessed, taking into account design and manufacturing aspects, demonstrating that an absorbed top surface heat flux of 20 MW/m2 should not be exceeded during ITER operations if the factor 1.4 margin to CHF, which must be respected for ITER monoblocks, is to be preserved.
Fabrication and qualification of a representative full-scale prototype of the International Thermonuclear Experimental Reactor (ITER) divertor inner vertical target is the subject of the contract ...assigned to Ansaldo Nucleare S.p.A (ANN) by Fusion for Energy, the EU-Domestic Agency. ENEA, as major partner of this contract activities, was in charge of the plasma facing units (PFU) fabrication by means of Hot Radial Pressing.
In parallel the other main fabrication steps were completed by ANN and its industrial partners based on the F4E technical requirements. Finally, the feasibility of the prototype assembly by integrating the PFUs onto the steel support structure has been demonstrated and the technical capability to reach the tight geometrical tolerances required for the plasma surface was checked.
High heat flux thermal fatigue testing on the 8 full-W PFUs installed on the testing frame will take place at the ITER Divertor Test Facility at Efremov Institute in Saint-Petersburg to finally demonstrating the PFUs thermal removal capability.
The reliability level reached by the HRP joining process and the industrial consolidation of the fabrication route, paves the way with confidence to a forthcoming series production of the ITER full-W divertor. The paper reports the full scale prototype main production achievements.
In order to evaluate the option to start the ITER operation with a full tungsten (W) divertor, the EU-DA launched an extensive R&D program. It consisted in its initial phase in the high heat flux ...(HHF) testing of W mock-ups and medium scale prototypes up to 20
MW/m
2 in the AREVA FE 200 facility (F). Critical heat flux (CHF) experiments were carried out on the items which survived the above thermal fatigue testing.
After 1000 cycles at 10
MW/m
2, the full W Plasma Facing Components (PFCs) mock-ups successfully sustained either 1000 cycles at 15
MW/m
2 or 500 cycles at 20
MW/m
2.
However, some significant surface melting, as well as the complete melting of a few monoblocks, occurred during the HHF thermal fatigue testing program representative of the present ITER requirements for the strike point region, namely 1000 cycles at 10
MW/m
2 followed by 1000 cycles at 20
MW/m
2.
The results of the CHF experiments were also rather encouraging, since the tested items sustained heat fluxes in the range of 30
MW/m
2 in steady-state conditions.
