No abstract available

[en] The fusion power reactor concept has an inherent advantage over fission reactor designs; the radioactivity of the fuel decays rapidly with time. However, this inherent advantage will be lost unless reactor structural materials can be employed which have similar advantages. Elemental additions of Nb, Mo, Ni, and N have half lives which will lead to high levels of radioactivity even after 100 years of storage. Therefore, the Office of Fusion Energy, DOE has funded development of low activation alloys, i.e., alloys with very limited additions of Nb, Mo, Ni and N. This paper describes the design and testing of one series of alloys from one of the classes of materials being studied. The alloys are bainitic or martensitic stainless steels which have tungsten and/or vanadium substituted for molybdenum. Alloy selection criterion, production and testing will be described including microstructural examination and mechanical properties testing following irradiation by fast neutrons in a liquid metal cooled fast breeder reactor. The results demonstrate that martensitic stainless steels in the Fe-9 to 12 Cr range appear most promising for fusion structural materials applications

[en] The objective of this work is to determine the effect of neutron irradiation on microstructural development in simple ferritic alloys. The present experiment also allows comparisons due to spectrum differences between EBR-II and FFTF. Microstructural examinations are reported for six Fe-Cr binary alloys (from 3 Cr to 18 Cr) following irradiation in FFTF at 430/degree/C to 15 dpa. All specimens contained radiation damage such as dislocation loops and voids. Swelling was highest, 0.26 percent, in the Fe-9Cr alloy with a corresponding dislocation network comprising equal parts of α <100> and α/2 <111> Burgers vectors. Less swelling was found in other alloys with corresponding dislocation structures composed mainly of α <100> loops. An example of a deformation slip band containing elongated voids was found which indicates that channel fracture may occur in ferritic alloys that contain uniform void swelling. 17 refs., 11 figs., 2 tabs

[en] Microstructural examinations are reported for nine specimen conditions of 2-1/4Cr-1Mo steel which had been irradiated by fast neutrons over the temperature range 390 to 510⁰C. Two heats of material were involved, each with a different preirradiation heat treatment, one irradiated to a peak fluence of 5.1 x 10²² n/cm² (E > 0.1 MeV) or 24 dpa and the other to 2.4 x 10²³ n/cm² (E > 0.1 MeV) or 116 dpa. Void swelling is found following irradiation at 400⁰C in both conditions and to 480⁰C in the higher fluence conditions. Concurrently dislocation structure and precipitation formed. Peak void swelling, void density, dislocation density and precipitate number density developed at the lowest temperature, approx. 400⁰C, whereas mean void size, and mean precipitate size increased with increasing irradiation temperature. The examination results are used to provide interpretation of in-reactor creep, density change and post irradiation tensile behavior

[en] This paper reviews the ADIP irradiation effects data base on ferritic (martensitic) alloys to provide reactor teams with an understanding of how such alloys will behave for fusion reactor first wall applications. Irradiation affects dimensional stability, strength and toughness. Dimensional stability is altered by precipitation and void swelling. Swelling as high as 25% may occur in some ferritic alloys at 500 dpa. Irradiation alters strength both during and following irradiation. Irradiation at low temperatures leads to hardening whereas at higher temperatures and high exposures, precipitate coarsening can result in softening. Toughness can also be adversely affected by irradiation. Failure can occur in ferritic in a brittle manner and irradiation induced hardening causes brittle failure at higher temperatures. Even at high test temperatures, toughness is reduced due to reduced failure initiation stresses. 39 refs

[en] Microstructural observations are reported for a series of five commercial ferritic alloys, 2 1/4 Cr-1 Mo, H-11, EM-12, 416, and 430F, covering the composition range 2.25 to 17% chromium, following EBR-II irradiation over the temperature range 400 to 650⁰C and to a maximum fluence of 17.6 x 10²² n/cm² (E > 0.1 MeV). These materials were confirmed to be low void swelling with maximum swelling of 0.63% measured in EM-12 following irradiation at 400⁰C to 14.0 x 10²² n/cm². A wide range of precipitation response was found both as a function of alloy and irradiation temperature. Precipitates observed included M₆C, Mo₂C, Chi, Laves, M₂₃C₆, α' and a low temperature phase as yet unidentified. It is predicted, based on these results, that the major impact of irradiation on the ferritic alloy class will be changes in postirradiation mechanical properties due to precipitation

[en] The contribution of anisotropy in Burgers vector distribution to irradiation creep behavior has been largely ignored in irradiation creep models. However, findings on Frank loops suggest that it may be very important. Procedures are defined to identify the orientations of a/2<110> Burgers vectors for dislocations in face-centered cubic crystals. By means of these procedures the anisotropy in Burgers vector populations was determined for three Nimonic PE16 pressurized tube specimens irradiated under stress. Considerable anisotropy in Burgers vector population develops during irradiation creep. It is inferred that dislocation motion during irradiation creep is restricted primarily to a climb of a/2<110> dislocations on 100 planes. Effect of these results on irradiation creep modeling and deformation induced irradiation growth is considered

[en] Work on microstructure in ferritic alloys following irradiation in FFTF is presented: Fe-Cr binary alloys ranging from 3 to 18 Cr, and substitutional ferritic alloys based on Fe-10Cr with minor additions of Si, Mn, V, Ta, and Zr. Irradiations were performed to fluences as high as 30 dpa at 425/degree/C. The simple Fe-Cr binary alloy behavior following irradiation to 15 dpa was similar to that observed following irradiation in EBR-II. All specimens contained radiation damage such as dislocation loops and voids. Swelling was highest, 0.26%, in the Fe-9Cr alloy with a corresponding dislocation network comprising equal parts of a<100> and a/2 <111> Burgers vectors. Less swelling was found in other alloys with a corresponding dislocation structure composed mainly of a<100> loops. A deformation slip band containing elongated voids indicated that channel fracture may occur in ferritic alloys that contain uniform void swelling. The behavior of substantial Fe-10Cr ferritic alloys following irradiation to 30 dpa demonstrated that different solutes affect microstructural development. Void and dislocation development were observed in all conditions but the amount of swelling and the dislocation character (a<100> versus a/2 <111>) and structure (network versus individual loops) varied both with solute type and amount of solute (0.1 or 1.0 wt %). Swelling was lowest in Fe-10Cr-1Si due to delayed void nucleation; the dislocation structures were loop dominated in both Fe-10Cr-1Si and Fe-10Cr-1V. Void shape was found to vary between dodecahedral with /011/ faces and /100/ truncations and cubic with /100/ faces and /011/ truncations. 26 refs., 21 figs., 4 tabs

[en] A broad range of ferritic alloys is possible which satisfy the low activation requirement for near-surface burial of fusion reactor materials after decommissioning. Low activation bainitic alloys in the Fe-2Cr composition range, martensitic alloys in the Fe-7 to 9Cr range and stabilized martensitic alloys in the Fe-12Cr range have been successfully fabricated and are undergoing testing as demonstrated by efforts in Europe, Japan and the United States. However, it is found that irradiation significantly degrades the properties of bainitic and stabilized martensitic alloys. Bainitic alloys containing vanadium develop severe hardening due to irradiation-induced precipitation at temperatures below 450/degree/C and extreme softening due to carbide coarsening at temperatures above 500/degree/C. Stabilized martensitic alloys which rely on manganese additions to provide a fully martensitic microstructure are embrittled at grain boundaries following irradiation leading to severe degradation of impact properties. The most promising composition regime appears to be the FE-7 to 9 Cr range with tungsten additions in the 2% range where high temperature mechanical properties and microstructural stability are retained and impact properties are relatively unaffected by irradiation. 25 refs., 9 figs. , 3 tabs

[en] A series of specimens of HT-9 heat 91354 have been examined following irradiation in HFIR to 39 dpa at 300, 400, 500 and 600⁰C and following irradiation in EBR-II to 29 dpa at 390 and 500⁰C. HFIR irradiation was found to have promoted helium bubble formation at all temperatures and voids at 400⁰C. Cavitation had not been observed at lower fluence, nor was it found in EBR-II irradiated specimens. The onset of void swelling in HFIR is attributed to helium generation. The observations provide an explanation for saturation of ductile-brittle transition temperature shifts with increasing fluence