Masami Shindo

Corrosion behaviour of high temperature alloys in impure helium environments

Abstract Corrosion tests with Ni-base high temperature alloys were carried out at 900 and 950°C in simulated high temperature reactor helium environments. It is shown that the carburization and decarburization behaviour is strongly affected by the Cr and Ti (Al) contents of the alloys. In carburizing environments, additions of Ti, alone or in combination with Al, significantly improve the carburization resistance. In oxidizing environment, the alloys with high Cr and Al (Ti) contents are the most resistant against decarburization. In this environment alloys with additions of Ti and Al show poor oxidation resistance. The experimental results obtained are compared with a recently developed theory describing corrosion of high temperature alloys in high temperature reactor helium environments.

Creep rupture properties of a NiCrW superalloy in air environment

The creep rupture properties in air environment were studied at 900, 1000 and 1050°C for bar, plate and seamless tube materials of a NiCrW superalloy developed for use at service temperatures around 1000°C. The long-term creep rupture strength is estimated by applying the time-temperature parameter method to the creep rupture data. Data of anomalous behaviour due to oxidation strengthening, which is observed in creep curves with rupture times above ∼ 10 000 h at 1000°C, are corrected for the application of the time-temperature parameter method. The Larson-Miller parameter method is better than the Orr-Sherby-Dorn parameter method in respect of curve fitting to the present creep rupture data. The creep rupture strength of the bar material with grain sizes above ASTM No. 2 and of the plate material is above 9.8 MPa for a 1 × 105 h life at 1000°C, which is the final target of this program. The creep rupture strength increases with increasing grain size and heat treatment temperature.

Effect of minor elements on irradiation assisted stress corrosion cracking of model austenitic stainless steels

Abstract A low impurity Fe 18Cr 12Ni (HP) and its heats doped with Si and C (HP + Si and HP + C) were irradiated to 6.7 × 10 24 n/m 2 ( E ≫ 1 MeV) at 513 K. The slow strain rate tensile (SSRT) tests were carried out at a constant strain rate of 1.7 × 10 −7 s −1 in high purity, 573 K water. Scanning electron microscopy on the fracture surface revealed that HP and HP + Si failed mainly by the intergranular stress corrosion cracking (SCC), while the major failure mode in HP + C was the transgranular SCC. All alloys exhibited radiation hardening. HP + Si exhibiting the smallest hardening showed uniform elongation of 17%, while HP and HP + C did not. Transmission electron microscopy was also carried out. Frank loops and unidentified small clusters were formed in HP and HP + C, while only small clusters were observed in HP + Si.

Improvement of the oxidation resistance of a graphite material by compositionally gradient SiC/C layer

Abstract For the improvement of the oxidation resistance of a graphite material, a preparation study of compositionally gradient SiC/C layer was performed. The compositionally gradient material of SiC/C, which is a graphite material with compositionally gradient SiC/C layer, was produced by a combination of the reaction between gaseous SiO and graphite and chemical vapour deposited SiC coating. The compositionally gradient material showed the same excellent oxidation resistance as the SiC coated graphite material at high temperatures in air. Against severe thermal cycle, furthermore, the compositionally gradient material exhibited more stable characteristic compared with the SiC coated graphite.

Functionally gradient material of silicon carbide and carbon as advanced oxidation-resistant graphite

For advanced graphite materials, functionally gradient material (FGM) of silicon carbide and carbon (SiC/C) has been prepared in order to improve the oxidation resistance. The FGM of SiC/C prepared by the reaction between graphite and gaseous silicon monoxide, 2C(solid) + SiO(gas) → SiC(solid)+ CO(gas), at high temperatures of around 1350°C has a gradient in the concentration of SiC in the graphite matrix, i.e. the ratio of SiC to C in the matrix gradually decreases toward inside, and exhibites better oxidation resistance compared with virgin graphite material.

Safety characteristics of the high temperature engineering test reactor

Abstract Various safety evaluations had been performed to confirm the validity of the design of High Temperature engineering Test Reactor (HTTR) facility taking into account the inherent safety features and characteristics in the design of the HTTR. It is shown that the reactor facility is so designed that (1) the integrity of fuel and reactor coolant pressure boundary is not damaged against the trouble of equipments, etc., during operation, (2) the influence of accidents including the rupture of reactor coolant pressure boundary, the reactivity initiated accident, etc. is not spread and (3) the release of radioactive materials under accidents is well mitigated.

Effects of carburization and aging on the tensile properties of an experimental Ni-Cr-W superalloy

Abstract Carburization tests with experimental fine- and coarse-grained Ni-Cr-W alloys strengthened by /ga-W phase were carried out at 950°C in heavily carburizing environment ( 90% Ar + 10% CH 4 ) to investigate the carburization behaviour and its effect on room temperature tensile properties; the aging effect was also examined. The results were compared with those of Hastelloy XR which is modified Hastelloy X. It is shown that for the determination of the effect of carburization on room temperature tensile properties of high temperature alloys the relationship between the tensile properties and the carbon content, the carburization resistance and the aging effect need to be considered.

Evaporation behavior of Hastelloy-X alloys in simulated very high temperature reactor environments

A sequential analysis was made on the material degradations during exposure of nickel-base corrosionresistant austenitic alloys to simulated very high temperature reactor environments. The materials tested were two modified versions of Hastelloy-X in terms of both increased manganese content for improved compatibility and decreased manganese content for possible adverse effects. Quantitative analysis of the specimens after exposure for 1000 h at several temperature steps from 850 to 1050/sup 0/C have revealed the temperature-dependent aspects of the processes including the depletion of chromium and manganese due to oxidation, evaporation, and carbon transfer into and/or from the materials. The material with enriched manganese, developed and specified as Hastelloy-XR, showed enhanced resistance to loss of chromium in terms of both oxidation and evaporation.

In-pile and post-irradiation creep of type 304 stainless steel under different neutron spectra

In addition to post-irradiation creep tests, in-pile creep tests were performed using newly developed technology with in situ measurement under different neutron spectra. The in-pile creep properties of type 304 stainless steel at 550°C appear to depend on neutron spectrum, but a spectral effect on post-irradiation creep properties is not clearly seen. The rupture time of in-pile creep under a high thermal neutron flux condition is the shortest. The order of the rupture time following the high thermal flux condition is post-irradiation creep, in-pile creep with a thermal neutron shield condition and finally creep of unirradiated material, all in increasing order. It is suggested that the acceleration of creep deformation and fracture observed in irradiation creep tests may be related to enhancement of thermal creep in terms of FMD increased under a high thermal neutron flux in addition to increased helium embrittlement.

A Distributed Database System for Mutual Usage of Materials Information (Data-Free-Way)

A distributed database system named ``Data-Free-Way`` for advanced nuclear materials has been developed by the National Research Institute for Metals (NRIM), the Japan Atomic Energy Research Institute (JAERI), and the Power Reactor and Nuclear Fuel Development Corporation (PNC) under a cooperative agreement. In this paper, features and functions of the system, including input data, are described together with a method to share the database among the three organizations as well as examples of the easy accessible search of material properties provided by the system. Results of an analysis of tensile and creep properties data on Type 316 stainless steel collected by the different organizations and stored in the present system are introduced as an example of attractive utilization of the system. Moreover, in order to known how to use the system, some trails of the WWW server of several sites in Data-Free-Way to supply the information on nuclear materials are introduced.