Carbon fiber composites (CFCs) are the first choice as plasma facing materials for the strike points of divertor targets for future nuclear fusion devices like WENDELSTEIN 7-X and ITER. For the ...application in these facilities several potential European 3D-CFCs were compared and qualified: (1) four material batches of NB31 produced by Snecma Propulsion Solide (SNECMA); (2) NB41, SNECMA, the upgraded version of NB31; (3) N31, SNECMA, which is densified by chemical vapor infiltration (CVI) instead of a final liquid pitch infiltration characteristic for NB31; (4) a new developmental 3D-CFC produced by DUNLOP.
The characterization of the composites is comprised of thermo-physical measurements and tensile tests. The results are correlated to density and microstructure and summarized as follows: (1) NB41 provides the highest thermal conductivity in the ex-pitch direction of ∼375
W/(m
K) at room temperature; (2) all material grades are, due to their heterogeneity, characterized by a relatively large scatter of mechanical properties; (3) the different densification process for N31 in comparison to NB31 has no influence on the material properties; (4) NB41 provides in all three directions a comparably high tensile strength with an average in the ex-pitch direction of ∼220
MPa; (5) the 3D-CFC from DUNLOP is comparable to NB41 but yet does not meet the specifications in the needling direction.
A research and development (R&D) programme for the ITER blanket-shield modules has been implemented in Europe to provide input for the design and the manufacture of the full-scale production ...components. It involves in particular the fabrication and testing of mock-ups and full-scale prototypes of shield blocks and first wall (FW) panels. This paper summarises the main achievements obtained so far and presents the latest results of this R&D programme. In particular, it reports the status of the shield fabrication development programme with the manufacture of a full-scale shield prototype. It also reports the latest results of high heat flux and thermal fatigue tests of FW mock-ups. It describes the preparation of irradiation experiments of Be coated FW mock-ups. Finally, it presents the outline of a possible qualification programme that each contributing participant teams should pass prior to the procurement of the blanket-shield modules for ITER.
The lifetime of structural components of spallation targets (beam window, liquid metal container, return hull) is determined by the irradiation-induced changes of the mechanical properties of their ...materials. An extensive test program was initiated using specimens obtained from spent target components from operating spallation facilities (Los Alamos Neutron Science Center, LANSCE and the Spallation Neutron Source at Rutherford–Appleton Laboratory, ISIS). The investigated materials include a nickel-based alloy (IN 718), an austenitic stainless steel (AISI 304L), a martensitic stainless steel (DIN 1.4926) and a refractory metal (tantalum). The materials experienced 800 MeV proton irradiation to maximum fluences of >10
25 p/m
2. The mechanical property changes were investigated by microhardness measurements, three-point bending tests and tensile tests at temperatures ranging from room temperature (RT) to 250 °C. Subsequent scanning electron microscopy was employed to investigate the fracture surfaces. Generally, irradiation hardening and a decrease in ductility with increasing proton fluence was observed. Nevertheless, all materials except IN 718 tested at RT, retained some ductility up to the maximum doses explored. The transmission electron microscopy investigation showed that a high density of ‘black dots’ and dislocation loops appeared in all materials. No effect of long-range radiation-induced segregation at grain boundaries was detected by energy dispersive X-ray investigation on AISI 304L and IN718 which failed by intergranular fracture.
EU R&D on the ITER First Wall Lorenzetto, P.; Peacock, A.; Bobin-Vastra, I. ...
Fusion engineering and design,
02/2006, Letnik:
81, Številka:
1
Journal Article, Conference Proceeding
Recenzirano
This paper presents the last results obtained in the ITER First Wall (FW) development work programme performed in Europe on beryllium (Be), copper (Cu) alloys and 316L(N) stainless steel (SS) joining ...by Hot Isostatic Pressing (HIPping). It reports first the performances under high heat flux testing of CuCrZr/316L(N) SS mock-ups and compares them with those obtained earlier with CuAl25/316L(N) SS mock-ups. It reports then the results of the last developments done on Be/Cu alloy HIPped joints. In particular it presents the results of high heat flux testing of Be/CuCrZr FW mock-ups and compares them with those obtained earlier with Be/CuAl25 FW mock-ups. It gives the status of on-going thermal fatigue tests of Be/Cu alloy FW mock-ups and presents the result of a neutron irradiation experiment performed on Be/CuAl25 mock-ups. The results obtained so far with FW mock-ups made from CuCrZr alloys have exceeded those obtained earlier with CuAl25 alloys.
This paper summarises the European R&D efforts for the manufacture of shield modules and divertor cassettes for the International Thermonuclear Experimental Reactor (ITER), including their plasma ...facing components. The various development steps are described as they had to be taken to resolve the fabrication issues, and to keep track with the evolving design requirements and solutions. For all components, the manufacturing feasibility has been demonstrated on prototype scale which puts Europe in the position to start the procurement as soon as the decision about ITER construction is taken. The time period remaining until then is used to optimise the fabrication processes and to develop more cost effective alternatives.
A broad spectrum of high heat flux test facilities are being used worldwide to investigate the thermal response of plasma facing materials and components to fusion relevant thermal loads. These tests ...cover both normal operation scenarios with cyclic thermal loads and power densities in the range of several MW
m
−2 and transient heat load tests to simulate short events such as edge localized modes, plasma disruptions and vertical displacement events. There is an urgent need for reliable quality control methods and non-destructive analyses during the procurement phase of ITER; most of these techniques are based on heat flux methods. To quantify irradiation induced property changes and to evaluate the overall performance of neutron irradiated components, miniaturized plasma facing components have been irradiated in fission reactors and post-irradiation tested. Furthermore, in-pile tests have been carried out in order to study the synergetic effects of heat loads and neutron loads.
This paper presents the main results obtained so far from the development work performed in Europe to define the joining conditions between beryllium (Be) tiles and the dispersion strengthened copper ...alloy (DS-Cu) heat sink material for the ITER primary first wall (PFW). Two Be/DS-Cu joining techniques were investigated: hot isostatic pressing and furnace brazing. Six PFW mock-ups have been thermal fatigue tested so far. One PFW mock-up with HIPped Be tile was tested at 2.5 MW/m
2 for 1000 cycles without any indication of failure. On two other mock-ups, Be tiles detached at or above 2.7 MW/m
2. Two others were tested at 0.7 MW/m
2 for 13 000 cycles also without any indication of failures. A first PFW mock-up with a furnace brazed Be tile was tested at 1.6 MW/m
2 for 1000 cycles. These results should be compared with the operation conditions of the ITER PFW, namely 0.5 MW/m
2 peak heat flux and off-normal events up to 1.4 MW/m
2. Thermal fatigue testing of other mock-ups is still in progress and the development programme is continuing to further increase the engineering margins while decreasing the fabrication cost of the PFW panels.
The study of melting behaviour and crack formation, which occurs during short transient events, is of significant importance for the qualification of materials for future fusion devices. Heat load ...simulations at room temperature with a pulse duration of 5
ms have been performed on beryllium (S65C) and various tungsten grades with unidirectionally elongated grains, ITER candidate materials for the first wall (Be) and the divertor region (W), at several power loads below and above the melting threshold (∼50
MW
m
−2·s
1/2 for W, ∼28
MW
m
−2
s
1/2 for Be). Crack formation and surface roughening have been studied at single and in case of tungsten also multiple shots.
Significant differences in the crack resistance and the crack pattern of the various tungsten grades below the melting threshold have been determined and further material degradation has been found after multiple shots. In microstructural and metallographic studies, the material damage has been qualified and quantified. Doing so the material's ability to avoid premature failure as well as plasma contamination (by evaporating tungsten particles below the melting threshold) during ITER operation has been characterised.
In recent years, material development for ITER blanket components has been fervent; yet, a lack of materials data has been identified by the ITER International Organization (IO) and other ITER ...parties. Therefore, extensive work on assessment of materials data and qualification of the materials with all the relevant interfaces (joints) has been performed to ensure that requirements are fulfilled for the ITER operational conditions with acceptable margins for the foreseen lifetime of the ITER project. This paper will provide an overview of this qualification program with examples of recent results within the scope of the European contribution of blanket components to ITER. Ongoing actions to achieve a comprehensive understanding of the three main materials, beryllium, CuCrZr and stainless steel grade 316L(N)-IG, and their joints included in the blankets are discussed.
A first primary wall small scale mock up with beryllium as the armor material was manufactured by hot isostatic pressing (HIPing) by CEA and tested at the electron beam test facility JUDITH. The mock ...up consisted of a 9.4 mm thick beryllium armor joint onto a 20 mm thick dispersion strengthened copper (DS-Cu) alloy plate. Type 316L Stainless Steel tubes, 10/12 mm in diameter were embedded in the DS-Cu, which was subsequently joined onto a 30 mm thick 316LN stainless steel plate. The mock up was tested under a surface heat flux of 2.5 MW m
−2 for 100 preparatory cycles and 900 cycles of 30 s heating and 30 s cooling time. At the end of the 1000 cycles, the surface and the Be/DS-Cu joint of the mock up did not show any damage due to the fatigue test.
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