M.P. Tanaka

Swelling behavior of austenitic stainless steels in a spectrally tailored reactor experiment: Implications for near-term fusion machines

Abstract Current designs for engineering test reactors such as the International Thermonuclear Experimental Reactor propose to use an austenitic stainless steel for the first wall. Most of the available swelling data have been derived from neutron-irradiation experiments in which helium generation rates are very low (fast breeder reactors) or very high (mixed spectrum reactors). Recently a spectrally tailored experiment was concluded in the Oak Ridge Research Reactor in which the helium generation rate and damage rate were maintained at values typical of a fusion reactor operating at ~ 1 MW/m 2 . It was found that the swelling behavior of a titanium-modified stainless steel (PCA) in both the cold-worked and solution-annealed conditions differed significantly from the behavior observed in earlier experiments in which the He/dpa ratio was either ~ 0.5 or ~ 50. The results suggest that there is a strong dependence of microstructural evolution on the He/dpa ratio. The data are shown to be consistent with earlier theoretical predictions of swelling behavior that is a non-monotonic function of the He/dpa ratio. Finally both the present data set and a larger collection of low-temperature swelling data are discussed in the context of near-term machines.

Tensile properties and microstructure of helium injected and reactor irradiateD V-20 Ti

The ojective of this work was to determine the effect of preinjected helium followed by neutron irradiation on the mechanical properties and microstructure of V-20% Ti. (MOW)

Microstructural development of austenitic stainless steels irradiated in HFIR

Abstract Microstructural developments of neutron irradiated JPCA and USPCA, which are Ti-modified austenitic stainless steels and candidate structural material for fusion reactor first wall, have been examined. The irradiation has been performed in High Flux Isotope Reactor (HFIR) at temperatures ranging between 300 and 600°C to a peak neutron fluence corresponding to approximately 34 dpa and 2500 appm helium. Microstructure of PCAs after irradiation at temperatures of 400°C and below suggests that the mutual stability of the radiation enhanced MC precipitation, and fine bubbles associated with such precipitation, has contributed to the extension of the transient regime of swelling to fluences above 34 dpa. At higher irradiation temperatures of 500°C and above, however, the conversion of some of the helium bubbles to voids has occurred at 34 dpa irradiation. MC precipitation on the radiation-induced dislocation lines is reduced at 500°C and above. This reduces the effective sink strength and increase the number of sites for helium bubble formation, which may lead to a severe reduction in the incubation regime of swelling at these temperatures. The factors controlling the stability of the MC precipitates in PCAs is discussed.

Microstructural development of PCAs irradiated in HFIR at 300 to 400° C☆

Abstract Microstructural developments were determined on solution-annealed (SA) and cold-worked (CW) JPCA and U.S. PCAs irradiated in the High Flux Isotope Reactor (HFIR) at 300 and 400°C. Irradiation produced damage levels of about 10 and 34 dpa and helium concentrations of around 580 and 2500 appm respectively. High concentrations of fine bubbles and MC precipitates, as well as Frank faulted loops, were observed in both SA and CW PCAs. Mutual stability of the MC particles and associated fine bubbles contributed to the extension of the transient regime of swelling to higher fluence. The irradiation responses of JPCA and U.S.-PCA were similar in the HFIR, despite minor compositional differences (P, B) between the two materials. Useful fusion applications of SA-PCA as well as CW-PCA in the 300 to 400° C temperature range are suggested from these data.

Temperature dependence of swelling in type 316 stainless steel irradiated in HFIR

The temperature dependence of swelling was investigated in solution-annealed (SA) and 20% cold-worked (CW) Type 316 stainless steel irradiated to 30 dpa at 300 to 600°C in the High Flux Isotope Reactor (HFIR). At irradiation temperatures ⩽ 400°C, a high concentration (2 to 4 × 1023 m−3) of small bubbles (1.5 to 4.5 nm diameter) formed uniformly in the matrix. Swelling was low ( < 0.2%) in both SA and CW materials irradiated to 30 dpa. In SA 316, cavity size increased but the number density decreased with increasing irradiation temperature above 500°C. At 500°C, there was a mixture of bubbles and voids, but at 600°C, most of the cavities were voids. Maximum swelling ( ~ 5%) occurred at 500°C. By contrast, cavities in 20% CW specimens were much smaller, with diameters of 6 and 9 nm at 500 and 600°C, respectively, suggesting that they were primarily bubbles. The cavity number density in CW 316 at both 500 and 600°C (~1 × 1022 m−3) was about one order of magnitude less than at 400°C. Swelling increased slightly as irradiation temperature increased, peaking at 600°C (0.3%). These results indicate that SA 316 swells more than CW 316 at 500 and 600°C, but both SA and CW 316 are resistant to void swelling in HFIR at 400°C and below to 30 dpa.

Post irradiation tensile and fatigue behavior of austenitic PCA stainless steels irradiated in HFIR

Abstract Mechanical properties were determined on solution annealed (SA) and cold worked (CW) JPCA (Ti-modified austenitic stainless steel) irradiated in the High Flux Isotope Reactor (HFIR) at temperatures ranging from 300 to 600°C. The irradiation produced damage levels from 16 to 56 dpa and helium concentration from 1020 to 4100 appm. The improved stability of MC precipitates which formed in the matrix during irradiation prevent loss of ductility at 500°C and below. Application of solution annealed JPCA is recommended for structural components of fusion reactors to be operated at 500°C and below.

Program of United states-Japan collaborative testing in HFIR and ORR☆

Abstract The objective of the program of US-Japan collaborative testing in HFIR and ORR is to investigate the response of austenitic stainless steels to levels of neutron radiation damage anticipated in fusion reactors. Emphasis is placed on United States and Japanese Type 316 stainless steels and prime candidate alloys (PCA), but improved alloys are also included. The program consists of eight HFIR target capsules and two ORR spectral tailoring capsules with test temperatures of 60 to 600°C and fluences of 30 to 50 dpa. Helium contents after irradiation match or exceed those produced in a fusion reactor at the same dpa level. Types of data to be obtained include tensile, fatigue, crack growth, irradiation creep, swelling, and microstructural evolution. Five HFIR capsules have been irradiated, and testing is under way. Less than 1% swelling is observed for Type 316 stainless steel and PCA alloys irradiated to about 30 dpa at 300 to 500°C.

The endurance limit of JPCA alloy at 703 K in vacuum

Abstract The high-cycle fatigue behavior of the Japanese Primary Candidate Alloy (JPCA) was determined at 703 K in vacuum. Four different thermomechanical treatments produced the changes in dislocation structure and dispersion of Ti-rich MC type carbides. Applying load controlled sine waves of 30 Hz in tension-compression mode, ductile fractures with striations and dimples were observed for both solution annealed and subsequently aged alloys at 1073 K, while brittle facets and wavy patterns with tearing by shear were observed for 15% cold worked alloy. The fatigue life was shown to be improved by the cold work in the relatively high-stress and short-life regime. But the fatigue endurance limit at 10 7 cycles for cold worked JPCA was 550 MN/m 2 , slightly lower than that for solution annealed JPCA, and was furthermore decreased by aging for the precipitation of MC type carbides.

Microstructural Development of Titanium-Modified Austenitic Stainless Steel Under Neutron Irradiation in HFIR up to 57 dpa

The Japanese prime candidate alloy (JPCA), a titanium modified austenitic stainless steel, has shown good performance after irradiation in HFIR up to 34 dpa at 300 C to 600 C, but considerable void swelling develops in solution annealed (SA) JPCA after irradiation 57 dpa at 500 C. However, cold worked (CW) or cold worked and aged (CW+A) JPCA still demonstrates good performance after similar irradiation. Swelling resistance appears to strongly depend on the behavior of fine titanium-rich MC precipitates. This paper describes the microstructural evolution process observed in the JPCA steel during HFIR irradiation. The onset of rapid void swelling was related to MC precipitate dissolution, and the instability of the MC was interpreted in terms of a model involving the build of and effects of a solute segregation zone in the matrix surrounding the precipitate particle. 8 refs., 9 figs., 3 tabs.

Neutron damage of austenitic stainless steels as candidate materials of blanket-components in a fusion reactor

Base materials of type 316 and Ti-modified austenitic steel (JPCA) and weld-joints of these materials were irradiated in HFIR at 55 °C to fluences up to 1.12×10 27 neutrons/m 2 (> 0.1 MeV), which produced up to 56 dpa and 3630 appm He. Tensile tests at room temperature showed a large increase in strength properties and that the weld-joints are weaker than the base-materials. The ductility of all materials was reduced by irradiation. For the irradiated weld-joint, deformation and fracture occurred mainly in part of the weldment. This data shows that the benefit of a cold-worked material (which is high strength) could no longer be expected in the weld-joint. The increase in strength and work-hardening rate by MC-precipitates, with the addition of Ti and Nb to the filter metal, is necessary to improve the deformation behavior of the irradiated weld-joints in type 316 stainless steel and JPCA.