Section 8. Dispersion-strengthened alloys
Perspective of ODS alloys application in nuclear environments

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Abstract

Oxide dispersion strengthened (ODS) steels are the most promising class of materials with a potential to be used at elevated temperature under severe neutron exposure environment. Leading technology development of ODS steels has been conducted at the Japan Nuclear Cycle Development Institute (JNC) particularly emphasizing fuel cladding application for fast reactors. This paper reviews the JNC’s activities on ODS steel development as ‘nano-composite materials’. Martensitic 9Cr-ODS and ferritic 12Cr-ODS steels have been successfully developed; Y2O3 oxide particles can be controlled on a nano-scale and high-temperature properties were noticeably improved through controlling the grain boundary structure on an atomic scale. The ODS-technology development achieved in the field of fast reactors should be effectively spun off to the fusion reactor first wall and blanket structural materials to allow for safe and economical reactor design.

Introduction

Ferritic/martensitic steels (FMS) are a primary candidate for the advanced fast reactor cladding/duct materials as well as fusion DEMO plant first wall and blanket structural materials because of their advantage to radiation resistance up to high neutron dose as high as 200 dpa [1], [2]. Their utilization is, however, limited to around 600 °C, which is due to inferior tensile and creep strength at higher temperatures. To achieve higher plant operation temperature for improved thermal efficiency, efforts have been made to improve high-temperature properties by means of controlling alloying elements and heat-treatment with stabilized carbide precipitates in FMS, especially for application in the power-generation industry [3]. Oxide dispersion strengthened (ODS) FMS are promising materials with a potential to be used at elevated temperatures due to the addition of extremely thermally stable oxide particle dispersion into the ferritic/martensitic matrix. The development of ODS FMS has been conducted in the field of fast reactor fuel cladding application [4], [5], [6], [7], [8] and fusion reactor materials application [3], [9], [10], [11], [12].

A leading technology development of ODS FMS has been conducted in the Japan Nuclear Cycle Development Institute (JNC) particularly emphasizing fuel cladding application for fast reactors. This technological R&D is believed to extend the performance of reduced activation ferritic steels as a system applicable in fusion structural materials. In this paper, JNC’s activities on the development of ODS FMS are reviewed. The underlying guideline for the processing method in the shape of panel and pipe will be provided on the basis of comparison of the current baseline properties with the requirements from the tentative fusion reactor materials design. Future work needed for ODS development also is presented.

Section snippets

JNC’s activity and progress

The research and development of the ODS FMS, as a prospective cladding material for the advanced fast reactor, are being conducted since 1987 in JNC. Fundamental studies concerning optimization of mechanical milling (MM) processing as well as effects of alloying elements on the high-temperature mechanical strength had been carried out in cooperation with fabrication vendors [13], [14]. Based on the results of those studies, the manufacturing of thin-walled cladding had been tried with

Target mechanical properties

In the first wall and breeder blanket structural materials for the DEMO fusion reactor, the tentative design requires an UTS of 500–550 MPa at 650 °C after 15 MW/m2 to maintain their integrity for the thermal stress, when reduced-activation ODS ferritic steels are applied [34]. Fig. 8 compares the tensile strength behavior of as-manufactured F82H (8Cr–0.1C–2W–0.2V–0.04Ta)[3] and martensitic 9Cr-ODS steel tubes. The design requirement for the tensile strength at 650 °C are just located at the

Future work

The future work toward realizing ODS fuel pins for advanced fast reactors is schematically represented in Fig. 11. We are currently starting the stage of materials system engineering, where cladding manufacturing dimension accuracy, joining with plug, inspection by ultrasonic inspection, large scale production and construction of engineering data-base are included. For technology demonstration as a fuel pin system, irradiation tests of ODS fuel pins, that are under preparation, will be

Conclusion

The martensitic 9Cr-ODS and ferritic 12Cr-ODS steels have been successfully developed as promising fuel cladding materials in fast reactors. Essentially in the ODS steels, the distribution of stable Y2O3 oxide particles can be controlled on a nano-scale that serves as a strong block for mobile dislocations and as a sink for radiation defects at the particle-matrix interfaces. Furthermore, high-temperature strength and ductility are far advanced through controlling the grain boundary structure.

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