K. Fukuya

Post-irradiation annealing effects on microstructure and helium bubbles in neutron irradiated type 304 stainless steel

Abstract Helium bubble nucleation and growth processes were studied at elevated temperatures on type 304 stainless steel which had been neutron irradiated to 1.4 × 1026 n / m 2 ( E > 1 MeV) and annealed at 400, 550, 650 and 900°C for 1 h. After annealing, specimens for microstructural observation were prepared and observed (by transmission electron microscope (TEM).) Radiation defects were present at high density in the as-irradiated specimens and annealed out with increasing annealing temperature. Helium bubbles became visible where defects were annealed out (above 650°C). The bubbles grew preferentially at dislocations and grain boundaries. At the grain boundaries, most of the helium bubbles formed on grain boundary dislocations. It was concluded that during post-irradiation annealing, the dislocations which remained after annealing played an important role in helium bubble growth, even in the grain boundaries.

Effect of carbon and nitrogen on grain boundary segregation in irradiated stainless steels

SUS304 stainless steels with carbon contents of 0.052%, 0.019% and 0.004% and SUS316L stainless steels with nitrogen contents of 0.095%, 0.032% and 0.003% were irradiated with 12 MeV Ni ions at 573 K to a dose of 1 dpa at 1 μm depth. Microstructure and grain boundary chemical composition were investigated using a transmission electron microscope with a field-emission-gun (FE-TEM) at the probe size of 0.5 nm. The number density of dislocation loop was higher as the carbon content was higher and was almost independent of nitrogen content. With increasing carbon and nitrogen content, the degree of Cr depletion and Si/Ni segregation was decreased. Both carbon and nitrogen suppressed the Cr depletion and Si/Ni segregation. The suppression effect of carbon was larger than that of nitrogen.

Stress corrosion cracking and intergranular corrosion of neutron irradiated austenitic stainless steels

Abstract The effects of irradiation on stress corrosion cracking (SCC) and intergranular corrosion (IGC) susceptibility were investigated in solution-treated Feue5f819Crue5f89Niue5f8Mn alloys and JPCA irradiated to 5.3×1024 n/m2 (E > 1 MeV) at 573 K. In Feue5f819Crue5f89Niue5f8Mn alloys, the irradiation enhanced IGC i n boiling HNO3 + Cr6+ solution when the alloys contained phosphorus and silicon and induced SCC in all the alloys with strain rate tensile tests in 571 K water containing 32 ppm oxygen. With increasing phosphorus and silicon contents. IGC was promoted but IGSCC was suppressed after irradiation. The results indicated that these elements are not the main contributors to irradiation-assisted SCC, although they affect SCC behavior. The Japanese Prime Candidate Alloy (JPCA) had better SCC resistance than Feue5f819Crue5f89Niue5f8Mn alloys under the present irradiation condition.

Effects of phosphorus, silicon and sulphur on microstructural evolution in austenitic stainless steels during electron irradiation

Fe-18Cr-9Ni-1.5Mn austenitic alloys containing phosphorus, silicon and sulphur were irradiated by 1 MeV electrons at 573–773 K. Phosphorus increased the intersitial loop nucleation and decreased the void swelling by increasing void number density and suppressing void growth. Silicon had a similar effect to phosphorus but its effect was weaker than phosphorus. Sulphur enhanced void swelling through increasing the void density. Nickel enrichment at grain boundaries was suppressed only in the alloy containing phosphorus. These phosphorus effects may be explained by a strong interaction with interstitials resulting in a high density of sinks for point defects.

Effect of neutron flux on low temperature irradiation embrittlement of reactor pressure vessel steel

Abstract Miniature Charpy V-notch impact test specimens of commercial reactor pressure vessel (RPV) steels having high and low copper contents were irradiated at the different irradiation positions with neutron flux levels of ∼6xa0×xa010 14 , ∼7xa0×xa010 15 , and ∼8xa0×xa010 16 n m −2 s −1 ( E xa0>xa01 MeV) to fluence levels ranging from ∼6xa0×xa010 21 to ∼7xa0×xa010 22 n m −2 ( E xa0>xa01 MeV) at temperatures of about 50°C to 150°C in the Japan Materials Testing Reactor (JMTR). The results showed that the radiation-induced increases in ductile-to-brittle transition temperature (ΔDBTT) at a neutron flux level of ∼6xa0×xa010 14 n m −2 s −1 were greater than those for neutron flux level of ∼7xa0×xa010 15 n m −2 s −1 . The neutron flux effect on embrittlement tended to be more pronounced in the lower neutron fluence range of ∼6xa0×xa010 21 –∼1xa0×xa010 22 n m −2 than in the higher fluence level of ∼7xa0×xa010 22 n m −2 , and also to be larger for the low copper steel than for the high copper steel, although the ΔDBTT for the high copper steel was larger than that for the low copper steel regardless of neutron fluence or flux. The displacement dose rate effect identified by the data converted to the ΔDBTT for the full size Charpy specimens from those for the miniature Charpy specimens was consistent with that based on the comparison of the results in the literature.

Hydrogen analysis and slow strain rate test in Ar gas for irradiated austenitic stainless steel

Abstract Hydrogen concentration in austenitic stainless steel irradiated with neutrons in boiling water reactors (BWRs) was measured and the effect of hydrogen in the austenitic stainless steel on intergranular cracking was investigated by the slow strain rate test (SSRT) in Ar gas. The hydrogen concentration decreased at low neutron fluences and increased at high neutron fluences. The decrease was attributed to the effect of heating or γ-ray irradiation at the early stage of reactor operation. The increase at high fluences was considered mainly due to the generation of hydrogen by nuclear transmutation. Intergranular cracking was not found for the specimen irradiated to a high fluence ( 1.4×10 26 n / m 2 ) in the SSRT at a very low strain rate ( 1.0×10 −8 s −1 ). This meant that the hydrogen concentration was too small to induce cracking, or hydrogen could not diffuse because of being trapped in irradiation defects at the test temperature.

Effects of neutron irradiation on corrosion and segregation behavior in austenitic stainless steels

Corrosion behavior and compositional changes at grain bounadries were examined in Type 304 and 304L stainless steels irradiated up to 6 × 1025n/m2 (E > 1 MeV) at 323 and 573 K. The rate of intergranular attack in boiling HNO3 +Cr6+ solution increased with increasing fluence in commercial purity stainless steels. High purity 304L (low phosphorus and silicon) was immune to corrosion in this solution both before and after irradiation. The enhancement of intergranular attack in commercial purity stainless steels due to irradiation was related to the increase of phosphorus and silicon at grain boundaries, as detected by STEM/EDX analysis.

The effect of hydrogen on microstructural changes using dual ion irradiation

Abstract The effect of hydrogen on cavity formation in vanadium was investigated using dual ion irradiation at 773,873 and 973 K with hydrogen injection of 0, 15, 30 and 60 appm/dpa doses up to 50 dpa. In some of the samples after dual beam irradiation with H and Ni ions, bimodal cavity size distribution was observed. The injected hydrogen may affect cavity nucleation in a similar manner as helium. There is a critical radius of the cavity under H dual irradiation which depends on the irradiation temperature, like He dual irradiation. The dependence of the critical radius on the amount of injected hydrogen is not clear, but it is suggested that the amount of hydrogen more than a certain value is necessary to promote the nucleation of cavities by injected hydrogen.

Effect of weld thermal cycle on helium bubble formation in stainless steel

Helium bubble structure was examined on a helium-implanted stainless steel after applying two kinds of heat input. Helium ions were implanted on Type 304 stainless steel at 573 K from 2 to 200 appm to a peak depth of 0.5 μm from the surface. After that, weld thermal history was applied by an electron beam. The cooling rates were selected to be 370 and 680 K/s from 1023 to 773 K. TEM observation revealed that nucleation and growth of helium bubbles were strongly dependent on the cooling rate after welding and the helium concentration.

Weldability of helium-containing stainless steels using a YAG laser

Bead-on-plate welding experiments using a 400 W YAG laser were conducted on SUS304 stainless steels implanted with helium ions of 0.5, 5 and 50 appm uniformly to a depth of 0.25 mm. High heat input welding at 20 kJ/cm caused surface grain boundary cracking in the heat-affected zone at 50 appm He. Cross-sectional observations after etching in oxalic acid solution revealed that bubble growth at grain boundaries in the heat-affected zone was enhanced at higher heat input and at higher helium concentrations. Bubble growth was negligible for the laser welding condition of 1 kJ/cm even at 50 appm He. The results suggest that YAG laser welding is a promising welding technique for stainless steels containing high amounts of helium.