Isao Masaoka

Solute segregation along non-migrated and migrated grain boundaries during electron irradiation in austenitic stainless steels

Solute distribution and precipitation in the vicinity of the grain boundary in Type 316 steels were studied during electron irradiation up to about 50 dpa at temperatures from room temperature to 873 K. Undersized solute atoms, such as nickel, silicon and phosphorus, segregate toward the grain boundary, and oversized solutes, chromium and molybdenum, segregate away from the grain boundary during irradiation in the temperature range between 623 and 873 K. Enrichment of silicon and phosphorus along the grain boundary occurs after the irradiation at room temperature. The segregation of solute atoms increases with irradiation temperature except for silicon and phosphorus; the concentration of silicon and phosphorus along the grain boundary exhibits a maximum at 773 K. Remarkable depletion of chromium with enrichment of nickel, silicon and phosphorus occurs in the area swept by the migrating grain boundary. Massive M23C6 type carbide precipitates in front of the migrating grain boundary during irradiation in the temperature range from 723 to 873 K in the steels.

Electrical resistivity recovery in Fe-Cr-Ni alloys after neutron irradiation at low temperature

Isochronal and isothermal annealing processes of austenitic Fe-Cr-Ni alloys and Type 316 steel are studied by electrical resistivity after fast neutron irradiation up to 2.9 × 1021n/m2 at 5 K and after quench from 1473 K. Little resistivity change is observed below 80 K in the irradiated specimens and below 280 K in the quenched ones. The resistivity variation of the annealing curves in the irradiated specimens is formed by the resistivity decrease due to defect annihilation and the resistivity increase due to structural change produced by the self-interstitial and vacancy migration. Self-interstitials and vacancies migrate above 100 K and 300 K, respectively. The radiation-induced reesistivity consists of two contributions; the resistivity increase due to defect formation and the resistivity decrease due to disordering of solute atoms by cascade collision during neutron irradiation.

Effects of surface finishing on stress corrosion cracking of austenitic stainless steels in high temperature water

The intergranular stress corrosion cracking (IGSCC) susceptibility of surface finished type 304 and 304L stainless steels has been studied using a constant load tensile specimen in 288 °C water containing 26 ppm dissolved oxygen. The study was to define conditions to prevent IGSCC in piping used in a boiling water reactor (BWR). The results are as follows:1)IGSCC susceptibility of type 304 stainless steel increased markedly by surface finishing, such as grinding, when in a sensitized condition; increasing surface roughness shortened the time to failure.2)Type 304 stainless steel specimens, which are solution heat treated, do not show IGSCC susceptibility even if they are ground.3)Type 304L stainless steels do not show IGSCC susceptibility even if they are ground and sensitized at 621 °C for 2 h.

Void formation and precipitation during electron-irradiation in austenitic stainless steels modified with Ti, Zr and V

Abstract Solution-annealed type 316 stainless steels modified with one or two elements of titanium, zirconium and vanadium were electron-irradiated up to a dose of about 50 dpa at temperatures of 773 to 823 K in a high voltage electron microscope. Addition of 0.3 wt% of titanium or zirconium to 316 steel remarkably reduce the void density at 823 K, compared with that in the standard 316 steel. Conversely, addition of 0.3 wt% of vanadium enhances void formation at 823 K; the void density is two orders of magnitude higher than that in the 316 steel. The enhancement is related to the radiation-induced V-carbides in the early stage of irradiation. Further addition of 0.15 wt% titanium to the V- and Zr-modified steels completely suppresses void formation up to 50 dpa at 823 K, because fine Ti-carbides precipitated along dislocations beyond about 10 dpa change dislocation bias to reduce the vacancy supersaturation rate. A beneficial effect of zirconium on void formation is disappeared by the addition of 0.15 wt% vanadium to the Zr-modified steel irradiated at 773 to 823 K.

Effects of nitrogen and carbon on void swelling in electron-irradiated austenitic stainless steel

Abstract Swelling behavior of solution annealed Type 316L stainless steel with 0.005 wt% carbon and 0.018 wt% nitrogen during electron irradiation at temperatures of 673 to 873 K was continuously observed with a high voltage electron microscope. Peak swelling of 4.2% occurred at 823 K in the 30 dpa irradiation for 316L steel. Addition of 0.08 wt% nitrogen to the steel reduced both the void number density and growth rate below 873 K; therefore, the resultant swelling was less than half that in the 316L at 30 dpa. Addition of 0.08 wt% carbon to the 316L steel drastically suppressed void formation at 823 K and above; no swelling was observed up to a dose of 30 dpa at the high temperatures. The swelling at 823 K was much less in the steel with both 0.08 wt% carbon and 0.08 wt% nitrogen than in the 316L steel or the steel with 0.08 wt% nitrogen alone.

Grain boundary migration during electron irradiation in austenitic stainless steels

Abstract The grain boundary migration in solution-annealed Fe-17% Cr-12% Ni alloy and Type 316 stainless steels during electron-irradiation at temperatures of 673 to 873 K was continuously observed with a high voltage electron microscope. The grain boundary migration begins after irradiation to about 3 dpa, prior to the visible void formation. Large voids are formed in the area swept by the migrating grain boundary, which results in large swelling in that area compared with the rest of the matrix. Remarkable depletion of chromium occurs in the swept area of the Fe-Cr-Ni ternary alloy, and massive M 23 C 6 type carbide precipitates are found in front of the migrating grain boundary in the 316 steel. The carbide impedes the grain boundary migration after irradiation beyond about 20 dpa.

Electrical resistivity change in Cu and AI stabilizer materials for superconducting magnet after low-temperature neutron irradiation

Abstract The change of electrical resistivity in magnetic field has been studied at 4.2 K in both copper and aluminum irradiated by fast neutrons at 5 K, together with the isochronal annealing behaviors and the effects of the irradiation-anneal cycles (cyclic irradiation). The increasing rate of the resistivity by the irradiation in aluminum is about three times as large as that in copper in zero magnetic field. The rates in the high purity copper with R 294k R 4,2k (RRR) of 1400 and aluminum with RRR of 1500 increase remarkably with an increase in applied magnetic field to the low irradiation doses, while those in the impure copper with RRR of 280 and 300 are independent of magnetic field. The radiation-induced resistivity in copper cannot be recovered even by annealing at 300 K. and the retained resistivity decreases in the cold-worked samples. The retained resistivity is accumulated by the cyclic irradiation. The radiation-induced resistivity of aluminum is completely annihilated by the annealing at 300 K.

Evaluation of hydrogen-assisted cracking behavior of low-alloy steel in the range 95 °C to 350 °C

In this report, hydrogen-assisted cracking (HAC) behavior of low-alloy steel was evaluated using a constant elongation rate tensile test, and the temperature and crack tip strain rate effects were observed. It was found that temperature dependence of the threshold condition (Cσmc) of HAC above about 100 °C followed the relation Cσmc = Kexp(−41,300/Rr) whereK is a constant andT is absolute temperature. The relationship between HAC growth rate and crack tip strain rate was established as almost linear, irrespective of temperature and hydrogen concentration at the crack tip. Hydrogen heat release tests were also performed. From these tests, formation and growth of microcracks which are trap sites of hydrogen were thought to be the mechanism of HAC in the steel. From this mechanism, HAC behavior of the low-alloy steel could be qualitatively explained.

Effect of chemical composition on corrosion resistance of zircaloy fuel cladding tube for BWR.

Effects of Fe and Ni contents on nodular corrosion susceptibility and hydrogen pick-up of Zircaloy were investigated. Total number of 31 Zr alloys having different chemical compositions; five Zr-Sn-Fe-Cr alloys, eight Zr-Sn-Fe-Ni alloys and eighteen Zr-Sn-Fe-Ni-Cr alloys, were melted and processed to thin plates for the corrosion tests in the environments of a high temperature (510°C) steam and a high temperature (288°C) water.In addition, four 450kg ingots of Zr-Sn-Fe-Ni-Cr alloys were industrially melted and BWR fuel cladding tubes were manufactured through a current material processing sequence to study their producibility, tensile properties and corrosion resistance.Nodular corrosion susceptibility decreased with increasing Fe and Ni contents of Zircaloys. It was seen that the improved Zircaloys having Fe and Ni contents in the range of 0.30[Ni]+0.15[Fe]≥0.045 (w/0) showed no susceptibility to nodular corrosion.An increase of Fe content resulted in a decrease of hydrogen pick-up fraction in both steam and water environments. An increase of Fe and Ni content of Zircaloys in the range of Fe≤0.25w/0 and Ni≤0.1w/0 did not cause the changes in tensile properties and fabricabilities of fuel cladding tube. The fuel cladding tube of improved Zircaloy, containing more amount of Fe and Ni than the upper limit of Zircaloy-2 specification showed no susceptibility to nodular corrosion even in the 530°C steam test.