Coolant Environments Research Articles

Design of Wrought Nickel-Base Alloys for Advanced High-Temperature Gas-Cooled Reactor Applications

The utilization of the high-temperature gas-cooled reactor (HTGR) for advanced or process heat applications will require the use of wrought components operating at temperatures up to 1000°C (1832°F...

数種の不純物を含むヘリウム中におけるハステロイXの高温すきま腐食

Interaction between a Ni-base heat-resistant alloy and simulated HTR primary coolant environment, which contained controlled concentrations of both oxidizing and carburizing impurities, was examined with emphasis placed on the reactions inside narrow crevice gaps. A test method of providing artificial crevice using a Mo container was developed to obtain quantitative results with reasonable reproducibility. The test environment was helium of low oxidizing potential with trace amount of gaseous impurities: H2, H2O, CO2 and CH4. The state of oxidation on the inner surfaces of the crevice was examined by measuring the change in concentration of Cr on the specimen surface using EPMA. Significant difference in the chenge of the surface was seen along with the distance from the crevice entrance. There was a preferential consumption of oxidizing species in the outer part of the crevice due to the formation of continuous oxide film, leading to porous or discontinuous oxide film formation, and the resultant extensive carburizations in the inner part of the crevice. Vaporization was noted far inside the crevice where oxide film was scarecely formed. The zone of continuous oxide film formation was found to extend into the crevice as a function of the 1/4 power of exposure time, and its reaching distance in a given time was proportional to the square root of opening of crevice gap. These relations were consistent with a model with an assumption that the diffusion of reactants through the gas was a rate determining step. The model based on the results obtained will be applied to the crevices with the similar geometrical effects such as the internals of fatigue and creep cracks and also actual crevice formed on the surface of heat exchanger tubes.

Fatigue crack growth rates of irradiated pressure vessel steels in simulated nuclear coolant environment

Fatigue crack growth rates, in a simulated pressurized water reactor primary loop environment, are presented for A508-2 and A533B steels in both the irradiated and unirradiated conditions. The initial results, for specific experimental conditions, constant load amplitude waveforms, and load ratio = 0.2, indicate that irradiation does not enhance fatigue crack growth rates more than the enhancement due to the environment alone. This article outlines the experimental techniques involved in this testing, and describes the results and their possible significance with respect to the existing safety analysis codes.

Technical Basis for Revised Reference Crack Growth Rate Curves for Pressure Boundary Steels in LWR Environment

Reference fatigue crack growth rate curves have been contained in Appendix A of Section XI since its inception. The curves have been designed to be applicable to carbon and low alloy pressure vessel steels exposed to either air or light water reactor coolant environments. Data obtained over the past several years have shown a different behavior of these steels in the LWR environment than that predicted by the present reference curve. A revised set of reference curves have been formulated, incorporating a new curve shape as well as a dependency of growth rate on R ratio (minimum load/maximum load). This article provides the background and justification for such a revision, details the methodology used to develop the revised curves, and includes an evaluation of the adequacy and impact of the revised curves as compared with the single curve which they replace.

Authors and Papers

Authors and Papers