Seishi Shima

Effects of hydrogen on mechanical properties in irradiated austenitic stainless steels

Abstract To investigate the hydrogen effect on mechanical properties of solution annealed Type 304 stainless steel, tensile tests of neutron irradiated materials were conducted after a hydrogen charging and discharging process (hydrogen treatment). Elongation was less with increasing neutron fluence after hydrogen treatment than that of as-irradiated specimens. Intergranular cracking occurred by the hydrogen treatment in heavier irradiated specimens, in which the Cr depleted zone along grain boundary was observed. Embrittlement and intergranular cracking after the hydrogen treatment were estimated to be attributed to the Cr depleted zone at the grain boundary due to neutron irradiation.

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 Fe19Cr9NiMn alloys and JPCA irradiated to 5.3×1024 n/m2 (E > 1 MeV) at 573 K. In Fe19Cr9NiMn 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 Fe19Cr9NiMn alloys under the present irradiation condition.

Stress corrosion cracking and intergranular corrosion of austenitic stainless steels irradiated at 323 K

The effect of neutron irradiation on intergranular corrosion (IGC) and stress corrosion cracking (SCC) susceptibility was investigated in several austenitic stainless steels irradiated up to 3.0 × 1025 n/m2 (E > 1 MeV) at about 323 K. The IGC of commerical purity (CP) 304 and CP304L in HNO3/Cr6+ solution was enhanced by neutron irradiation, while HP304 was immune to corrosion. The stainless steels except nuclear grade 316 (316NG) failed intergranularly after neutron irradiation by SSRT. HP304 showed higher IASCC susceptibility than CP304 and CP304L. The occurrence of IASCC in both HP and CP 304 indicated that radiation enrichment of impurities might not be main contributor to IASCC. 316NG had better SCC resistance than 304 stainless steels under the present irradiation conditions. IASCC susceptibility could be correlated to the austenite phase stability.

Underwater cutting technology of thick stainless steel with YAG laser

In nuclear power plants, irradiated materials like Control Rod (CR) should be stored underwater after service. Due to reducing the storage space, underwater cutting technology is expected. In this study, we developed underwater cutting technology of thick stainless steel with YAG laser in order to cut used CR. Preliminary tests were performed with flat plate test-pieces to optimize the cutting conditions. Due to creating a local dry area between nozzle and test-piece, high-pressure air was blown from the nozzle. Underwater laser cutting was carried out by laser irradiation power of 4 kW, changing the parameters of cutting speed, distance between the nozzle and test-piece, and thickness of the test-piece. We also investigated the wastes like dross and aerosols by laser cutting. Amount of dross was approximately 0.1 kg/m after cutting a 14 mm thick stainless steel plate, which is estimated to be less than other cutting method. Based on these results, we developed underwater cutting system of CR test-piece with YAG laser as a mock-up test. In the cutting torch, there was tracking system was introduced to keep the distance between the nozzle and the test-piece constant, and cutting monitor was also set-in to detect whether the test-piece was successfully cut or not. We have already tried to cut the CR test-piece with this facility and successfully cut in half.

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.

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.

Hafnium corrosion behavior in high-temperature steam

Abstract Corrosion properties for three kinds of hafnium have been examined in 10.5 MPa steam at 773 K. Nuclear grade hafnium formed a shiny black film with a few white nodules, which were found to be monoclinic HfO 2 . Hydrogen pick-up fraction during the corrosion amounted to ca. 40% to produce hydrogen dissolved in the hafnium matrix with evolving the remaining 60% hydrogen as H 2 gas. Sponge and crystal bar hafnium were also examined and they showed superior and inferior corrosion resistance, respectively, to nuclear grade hafnium. The corrosion resistance for hafnium increased with increasing the iron content involved in hafnium for these three kinds of hafnium and, correspondingly related to the density of iron-containing second phase particles, which were characterized to be face centered cubic Hf 2 Fe. The corrosion mechanism, which was previously proposed for Zircaloy nodular corrosion, was adopted with making minor alterations, to explain the hafnium corrosion properties.

Laser-Based Maintenance and Repair Technologies for Reactor Components

Stress corrosion cracking (SCC) is the major factor to reduce the reliability of aged reactor components. Toshiba has developed various laser-based maintenance and repair technologies and applied them to existing nuclear power plants. Laser-based technology is considered to be the best tool for remote processing in nuclear power plants, and particularly so for the maintenance and repair of reactor core components. Accessibility could be drastically improved by a simple handling system owing to the absence of reactive force against laser irradiation and the flexible optical fiber. For the preventive maintenance, laser peening (LP) technology was developed and applied to reactor components in operating BWR plants. LP is a novel process to improve residual stress from tensile to compressive on material surface layer by irradiating focused high-power laser pulses in water. We have developed a fiber-delivered LP system as a preventive maintenance measure against SCC. Laser ultrasonic testing (LUT) has a great potential to be applied to the remote inspection of reactor components. Laser-induced surface acoustic wave (SAW) inspection system was developed using a compact probe with a multi-mode optical fiber and an interferometer. The developed system successfully detected a micro slit of 0.5mm depth on weld metal and heat-affected zone (HAZ). An artificial SCC was also detected by the system. We are developing a new LP system combined with LUT to treat the inner surface of bottom-mounted instruments (BMI) of PWR plants. Underwater laser seal welding (LSW) technology was also developed to apply surface crack. LSW is expected to isolate the crack tip from corrosive water environment and to stop the propagation of the crack. Rapid heating and cooling of the process minimize the heat effect, which extends the applicability to neutron-irradiated material. This paper describes recent advances in the development and application of such laser-based technologies.© 2004 ASME

Laser de-sensitization treatment for inside surface of SUS304 stainless steel pipe welds in nuclear power plants

A technology to prevent the occurrence of Intergranular Stress Corrosion Cracking (IGSCC) by irradiating a high power Nd:YAG laser beam was developed. Laser Desensitization Treatment (LDT) process was realized by irradiating a laser beam onto the sensitized Heat Affected Zone (HAZ) surface of SUS304 stainless steel. LDT was formed by both a molten layer of approximately 0.2mm depth and a solution heat treated layer. The results of a Creviced Bent Beam (CBB) test showed that no cracks had appeared on the surface of LDT. After LDT was applied in the vicinity of welding joints on the inside surface of pipes, tensile residual stress was measured there. On the other hand, the tensile stress of outside surface of the pipes was decreased. From these results, LDT processing on the inside surface of a pipe can be expected to prevent the occurrence of IGSCC owing to the effect of both metallurgical improvement and decrease of the residual stress on the outside surface of the pipe. We developed the LDT processing system and successfully applied on the pipes of some actual nuclear power plants.A technology to prevent the occurrence of Intergranular Stress Corrosion Cracking (IGSCC) by irradiating a high power Nd:YAG laser beam was developed. Laser Desensitization Treatment (LDT) process was realized by irradiating a laser beam onto the sensitized Heat Affected Zone (HAZ) surface of SUS304 stainless steel. LDT was formed by both a molten layer of approximately 0.2mm depth and a solution heat treated layer. The results of a Creviced Bent Beam (CBB) test showed that no cracks had appeared on the surface of LDT. After LDT was applied in the vicinity of welding joints on the inside surface of pipes, tensile residual stress was measured there. On the other hand, the tensile stress of outside surface of the pipes was decreased. From these results, LDT processing on the inside surface of a pipe can be expected to prevent the occurrence of IGSCC owing to the effect of both metallurgical improvement and decrease of the residual stress on the outside surface of the pipe. We developed the LDT processing sy...