[en] This report presents the results of the collection and subsequent analysis of fracture toughness data on the austenitic stainless steels used in PWR internals. These data were made available by CEA Saclay and ECN Petten in the framework of a CEC study contract on PWR internals. The data include austenitic stainless steel plate 304L and 316L, as well as the weld metals 308L, 19-12-2, 17-13-2 and 16-8-M. Test specimens of these materials were irradiated in earlier programs, up to doses of 30 dpa at temperatures of approximately 400 C. Test specimens include several CT-type and impact test specimens. The results show a degradation of the fracture toughness with increasing irradiation dose. The degradation in terms of the ratio of the fracture toughness at 1 mm crack extension of irradiated to unirradiated material, turns out to be remarkably similar for all materials, a common trend curve can be fitted through the data points. In general there is a quick degradation of the fracture toughness during the first 5 dpa's, at higher doses the degradation seems to saturate. (orig.)

[en] Within the framework of the European Fusion Technology Programme and the RandD for International Thermonuclear Experimental Reactor (ITER), ECN is investigating the irradiation behaviour of the structural materials for ITER. The main structural material for ITER is austenitic stainless steel Type 316L(N)-ICI. The operating temperatures of (parts of) the components are envisaged to range between 350 and 700K. A significant part of the dose-temperature domain of irradiation conditions relevant for ITER has already been explored. There is, however, very little data at about 600K. Available data tend to indicate a maximum in the degradation of the mechanical properties after irradiation at this temperature, e.g. a minimum in ductility and a maximum of hardening. Therefore an irradiation program for plate material 316L(N)-IG, its Electron Beam (EB) weldments and Tungsten Inert Gas (TIG) weld metal, and also including Hot Isostatically Pressed (HIP) 316L(N) powder and solid-solid joints, was set up in 1995. Irradiations have been carried out in the High Flux Reactor (HFR) in Petten at a temperature of 600K, at dose levels from 1 to 10 dpa. The paper presents the currently available post-irradiation test results. Next to tensile and fracture toughness data on plate, EB and TIG welds, first results of powder HIP-ed material are included. 15 refs

[en] The low cycle fatigue properties of unirradiated and neutron irradiated Type 316LN-SPH stainless steel plate and EB welded material have been measured at a temperature of 350K. Miniaturized test specimens of the European reference heat 1 for fusion (ERH1) have been irradiated in the High Flux Reactor (HFR) of Petten, Netherlands, simulating the first wall conditions by a combination of high displacement damage with high amounts of helium. This work was performed as part of the European Fusion Technology Programme for ITER as 'Irradiation testing of stainless steel'. The report contains the experimental conditions and summarizes the test results. The stress strain response of EB welded material was found to be identical to that of plate material. For both plate and weldments a large increase in first cycle tensile stress is observed, which saturates within 0.7 dpa. The irradiated material shows continuous cyclic softening during low cycle fatigue testing, decreasing to about the stress level of unirradiated material at fatigue life. EB welded material tends to have longer mean fatigue lives than plate material, but the data have overlapping scatter bands. Material irradiated to 0.7 dpa and material irradiated to 5 dpa show similar fatigue life behaviour as unirradiated material. It is demonstrated that the RCC-MR curve is sufficiently conservative for use as design curve in case of application of Type 316LN-SPH plate and EB- welded material at the temperature of 350K. Furthermore, it is also demonstrated that for irradiation damage dose levels up to 5 dpa at 350K there is no need to apply additional safety factors for the number of allowable cycles up to 105 cycles. 15 figs., 6 tabs., 18 refs

[en] In the framework of the European Fusion Technology Programme ECN participates in a NET task PSM-8 'Coatings and Surface Effects on Stainless Steel 316L'. High emissivity coatings were developed for enhanced heat transfer from graphite tiles to a Stainless Steel First Wall. Four candidate materials, Cr₂O₃, Black Cr, Al₂O₃/TiO₂ and TiC were tested as candidate high emissivity coatings. These coatings were manufactured by atmospheric and vacuum plasma spraying technique and the Black Chromium coatings were manufactured by a galvanic coating technique. The tests included total emissivity measurements and Low Cycle Fatigue (LCF) experiments. The total emissivity of two TiC coatings at 525 K appeared to be 0.62 and 0.64. The total emissivity of the TiC and 5 wt% TiO₂/Al₂O₃ coating was about 0.7. (orig.)

[en] The tensile properties of unirradiated and neutron irradiated type 316L(N)-SPH stainless steel plate, EB weldments, 16-8 TIG-weldments, and full 16-8 TIG-deposits have been measured. Miniature 4 mm diameter test specimens of the European Reference Heat 1 and 2 (ERH), and 4 mm and some 8 mm diameter specimens of the weldments mentioned above, were irradiated in the High Flux Reactor (HFR) in Petten, The Netherlands, simulating the first wall conditions by a combination of high displacement damage with high amounts of helium. The irradiation conditions were 0.5 and 5 displacements per atom (dpa) at 350K and 0.5 and 5 dpa at 500K. Testing temperatures ranged from 300K to 850K. This work was performed as part of the European Fusion Technology Programme for ITER as 'Irradiation testing of stainless steel' The report contains the experimental conditions and summarises the results. The tensile properties of the unirradiated ERH's 1 and 2 plate materials were found to differ slightly but significantly: ERH2 has a lower UTS, but higher yield strength and ductility than ERH1. The plate materials have lower yield strength in the unirradiated condition than all of the weldments (EB, TIG-weld and TIG-deposit), accompanied by a higher ductility of the plate materials. When irradiated at 350K the differences in strength between the plate and weld materials decrease, but the ductility of the plate remains higher than that of the weldments. A saturation of irradiation damage has taken place already at about 0.5 dpa. When irradiated at 500K the plate material continuously hardens up to 5 dpa, where it has lost all uniform plastic ductility. The weldments show similar but less dramatic hardening and loss of ductility as the plate material for both irradiation conditions. 54 figs., 17 tabs., 21 refs

[en] Reduced activation Ferritic/Martensitic (F/M) alloys are being developed for application in future thermonuclear fusion reactors by various partners in the fusion community. One of the general drawbacks of these F/M alloys is the low temperature irradiation embrittlement. The irradiation behaviour of the reduced activation F/M steel IEA heat F82H, and of a laboratory heat specified by ECN and produced by British Steel was investigated by irradiation experiments in the High Flux Reactor (HFR) in Petten. Mechanical test specimens, such as tensile, miniature Charpy (KLST-type) impact specimens, and Compact Tension (CT) specimens, were neutron irradiated up to a dose level of 2.3-3.8 dpa at a temperature range of 280 to 370 deg. C. The results show moderate irradiation hardening (200 MPa at 27 deg. C) and a radiation induced shift of the DBTT for KLST's of approx. 150 deg. C for the IEA Heat F82H irradiated at approx. 300 deg. C. The ECN-BS laboratory heat shows similar irradiation hardening but a much smaller shift in DBTT, i.e. approx. 50 deg. C when irradiated at 350 deg. C. This lower radiation induced shift of the DBTT indicates that it might be possible to develop F/M steels with low irradiation embrittlement. (authors)

[en] A methodology for the structural integrity assessment of irradiated flawed components is provided. The disciplines required for the assessment are fracture mechanics, applied mechanics, materials research, and experimental and analytical neutron metrology. A description of the disciplines is given including the input information and resulting output data of each discipline. The deterministic assessment procedure for fracture initiation is based on the LEFM (Linear Elastic Fracture Mechanics) K-concept (K is the stress intensity factor) and the EPFM (Elastic Plastic Fracture Mechanics) J-integral concept (J is the parameter to characterize the singular stress and strain fields around the crack tip in case of yielding) resulting for instance in a failure assessment diagram. The uncertainties in the material data, loading conditions, and neutron metrology results can be addressed by performing a probabilistic analysis. This approach enables to perform a sensitivity and uncertainty analysis. The fracture assessment methodology described in the report can be elaborated, detailed, and validated by performing a case study on flawed irradiated component material, preferably from a real structure component which has operated in a nuclear environment. 11 figs., 14 refs

[en] The irradiation behaviour of electron beam (EB) and tungsten inert gas (TIG) welded joints of the reduced-activation martensitic steel IEA heat F82H-mod. was investigated by neutron irradiation experiments in the high flux reactor (HFR) in Petten. Mechanical test specimens, such as tensile specimens and KLST-type Charpy impact specimens, were neutron irradiated up to a dose level of 2-3 dpa at a temperature of 300 deg. C in the HFR reactor in Petten. The tensile results for TIG and EB welds are as expected with practically no strain hardening capacity left. Considering impact properties, there is a large variation in impact properties for the TIG weld. The irradiation tends to shift the DBTT of particularly the EB welds to very high values, some cases even above +250 deg. C. PWHT of EB-welded material gives a significant improvement of the DBTT and USE compared to the as-welded condition

[en] Type 316 stainless steel is the primary candidate austenitic structural material for fusion first wall constructions. Here, type 316L(N) stainless steel plate material has been irradiated up to 10 dpa at temperatures of 80, 225, 325, and 425 C in the High Flux Reactor (HFR) of Petten. Tensile tests have been performed in the temperature range from RT to 575 C at a conventional strain rate of 5 x 10^-4 s^-1. The results of the tensile tests are analyzed in terms of irradiation hardening and loss of ductility due to irradiation. Tensile properties saturate in the early stage (within 0.65 dpa) at the lowest applied irradiation temperature. It is indicated that the most severe degradation of tensile ductility occurs in the temperature range of 275 to 350 C. Comparison with literature data revealed a large scatter in irradiation hardening at irradiation temperatures above 325 C

[en] The tensile properties of neutron irradiated type 316L(N) plate and EB welded material have been measured between 300 and 900 K. The test specimens were irradiated in the Petten High Flux Reactor up to 10 dpa at 500, 600, and 700 K. The tensile tests were performed at the conventional strain rate of 5x10^-4 s^-1. Additional experiments have been performed at strain rates of 10^-6 and 10^-2 s^-1 at 850 and 900 K respectively. An increase in tensile strength of about 600 MPa was observed for irradiation temperatures of 500 and 600 K, with a corresponding loss of ductility. Less irradiation hardening was observed for an irradiation temperature of 700 K. The irradiation-induced increase in tensile strength saturated at 5 dpa. The irradiated material appeared to be strain-rate sensitive at a test temperature of 900 K. ((orig.))