Tatsuo Kondo

Research and development on heat-resistant alloys for nuclear process heating in Japan

The developments of the last decade are reviewed on a technical basis for heat-resistant alloys in application to the high-temperature structural components of the process heating high-temperature gas-cooled reactor. The major activities have fallen into two categories: the near-term development for the experimental reactor and the long-term R and D second generation applications, i.e., for the materials to be used in the second-stage heat exchanger installation in the experimental reactor and those for advanced-stage reactors with very high outlet temperatures. In both categories of programs, significant advances have been made, respectively, in providing and testing a modified commercial alloy with enhanced compatibility with the service environments and in selecting potential high performance alloys from the new developmental candidate alloys. Modification of the existing commercial alloy was achieved through the application of the finding on enhanced oxidation resistance by controlling the common impurities in the material, while the enhanced creep rupture strength recognized in the best performing new alloys has been attributed to the precipitation of a tungsten-rich phase (..cap alpha../sub 2/) during holding at test temperatures. The new alloy development program currently under way is also introduced.

IFMIF, its facility concept and technology

The historical background and the current status of the high energy, flux neutron irradiation test facility for fusion materials development are described focusing on the recent progress of the International Energy Agency (IEA) Conceptual Design Activity (CDA) on IFMIF. The reference design incorporated the results of the past international collaboration intended to select an optimum source concept, validate relevant technologies and identify requirements for the mission. Throughout the design period, the international design team closely followed the requirements with the guide of an authorized users group. The entire process was promoted by an international steering group based on an assumed material development plan and schedule to be phased with the design and construction of the DEMO reactor. The facility is able to satisfy the requirements for hypothetical test matrices using current candidate materials and includes a flexibility for future upgrading. Options for the next required steps are also discussed.

Oxidation rate of advanced heat-resistant steels for ultra-supercritical boilers in pressurized superheated steam

Oxidation kinetics of recently developed ferritic heat-resistant steels, HCM12A, NF616, and HCM2S, were investigated in a superheated steam to evaluate the effects of chemical composition of the steels, testing temperature (560-700°C), steam pressure (1-10 MPa), and degrees of microstructural evolution by aging on oxidation. The contribution of alloyed Cr to oxidation resistance was pronounced above 600°C, while no material dependency was found at 600°C or lower. The apparent activation energy of the oxidation rate clearly changed at around 600°C for NF616 and HCM12A. In contrast, HCM2S showed single activation energy over the range of temperatures. Although temperature and chemical composition were the major factors, steam pressure also showed a clear negative effect on the oxidation rate in the lower temperature range, 570-600°C.

Materials development and testing aspects of IFMIF in the conceptual design stage

Abstract The conceptual design activity for the International Fusion Materials Irradiation Facility (IFMIF-CDA, 1995–1997) has recently attained its systems integration step incorporating three major subsystems, i.e. remote handling cell/irradiation test assembly, liquid metal target and accelerators. The mission is to provide a facility for testing candidate materials for fusion power reactors to full lifetime performance and also to provide scientific bases for calibration and validation of data from other radiation sources. The reference design considered to overcome the technical barriers has remained since the time of the early Fusion Materials Irradiation Test Facility (FMIT) design, and followed closely the ‘Users Requirements’ in terms of the test volume, flux, flux spatial distribution, energy spectrum, attainable annual fluence, etc. Implications of the interfaces between the requirements and the design approaches are discussed, and a perspective of the engineering validation to be done prior to construction and operation is presented.

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.

An evaluation of potential material–coolant compatibility for applications in advanced fusion reactors

Abstract In assessing possible potential issues for fusion applications, the compatibility of several metallic structural materials was examined using high temperature/pressure steam as test environment. High corrosion resistance associated with protective oxide film formation was regarded as essential for the function of protecting from tritium permeation and corrosion damage. A Ti–Al-based intermetallic compound with V addition, recently developed, showed excellent performance. A low-activation ferritic/martensitic steel, F82-H, was comparable with the current advanced materials for modern supercritical fossil boilers, while some potential vanadium alloys, although not intended for use in steam, were found less compatible.

Postirradiation Tensile and Creep Properties of Heat-Resistant Alloys

The effect of neutron irradiation on hightemperature tensile and creep properties of austenitic heat-resistant alloys was studied. The effect, which appeared in the loss of ductility at elevated te...

Low-Cycle Fatigue of Heat-Resistant Alloys in High-Temperature Gas-Cooled Reactor Helium

Strain controlled low-cycle fatigue tests were conducted on four nickel-base heat-resistant alloys at 900/sup 0/C in simulated high-temperature gas-cooled reactor (HTGR) environments and high vacuums of about 10/sup -6/ Pa. The observed behaviors of the materials were different and divided into two groups when tests were made in simulated HTGR helium, while all materials behaved similarly in vacuums. The materials that have relatively high ductility and compatibility with impure helium at test temperature showed considerable resistance to the fatigue damage in impure helium. On the other hand, the alloys qualified with their high creep strength were seen to suffer from the adverse effects of impure helium and the trend of intergranular cracking as well. The results were analyzed in terms of their susceptibility to the environmentenhanced fatigue damage by examining the ratios of the performance in impure helium to in vacuum. The materials that showed rather unsatisfactory resistance were considered to be characterized by their limited ductility partly due to their coarse grain structure and susceptibility to intergranular oxidation. Moderate carburization was commonly noted in all materials, particularly at the cracked portions, indicating that carbon intrusion had occurred during the crack growth stage.

Creep and Rupture Behavior of a Special Grade Hastelloy-X in Simulated HTGR Helium

Creep and rupture tests were conducted for Hastelloy-XR (a modified version of the conventional Hastelloy alloy X) at 800, 900, and 1000/sup 0/C in simulated high-temperature gas-cooled reactor helium. Creep testing machines with special control of helium chemistry were used. As a result, the scatter of creep-rupture data could be reduced, and the variability of creep-rupture behavior due to manufacturing history could be resolved. Results of metallography and carbon analysis of ruptured specimens showed that the material improved resistance to corrosion in the helium environment, and carbon intrusion during the steady-state creep stage was suppressed to a negligible level. Under refined test conditions combined with the quality controlled material, it was demonstrated that there was little significant difference between helium and air in the creep-rupture results obtained at 800 to 1000/sup 0/C up to about 10/sup 4/h. The importance of maintaining the protective function of the surface oxide film of alloys was stressed in securing reproducibility and predictability of long-time creep performance.

Crack growth behavior of ferritic steel for USC boilers in pressurized superheated steam

Abstract In this study, effects of superheated steam on cyclic crack propagation behavior of a heat resistant steel were investigated. Crack propagation experiments were carried out on NF616 (9Cr-0.5Mo-2WVNb) in pressurized superheated steam (600°C/10MPa) under cyclic loading either with or without holding time at constant load. Superheated steam environment has two opposing effects on cyclic crack growth, acceleration and retardation. A modified tarnish rupture (TR) model has been proposed to explain the crack propagation behavior. The crack propagation rate estimated based on the TR-type model well agreed with the experimental data.