AGC-4 Graphite Preirradiation Data Analysis Report
The Advanced Reactor Technology (ART) Graphite R&D program is conducting an extensive graphite irradiation program to provide data for licensing of a high temperature reactor (HTR) design. In past applications, graphite has been used effectively as a structural and moderator material in both research and commercial high temperature gas cooled reactor (HTGR) designs. Nuclear graphite H 451, used previously in the United States for nuclear reactor graphite components, is no longer available. New nuclear graphite grades have been developed and are considered suitable candidates for new HTR reactor designs. To support the design and licensing of new HTR core components within a commercial reactor, a complete properties database must be developed for these current grades of graphite. Quantitative data on in service material performance is required for the physical, mechanical, and thermal properties of each major graphite grade with a specific emphasis on data accounting for the life limiting effects of irradiation creep on key physical properties of the HTR candidate graphite grades. Further details on the research and development activities and associated rationale required to qualify nuclear grade graphite for use within the HTR are documented in the graphite technology research and development plan. Based on experience with previous graphite core components, the phenomenon of irradiation induced creep within the graphite has been shown to be critical to the total useful lifetime of graphite components. Irradiation induced creep occurs under the simultaneous application of high temperatures, neutron irradiation, and applied stresses within the graphite components. Significant internal stresses within the graphite components can result from a second phenomenon—irradiation induced dimensional change—where the graphite physically changes (i.e., first shrinking and then expanding with increasing neutron dose). This disparity in material volume change can produce significant internal stresses within graphite components. Irradiation induced creep relaxes these large internal stresses, thus reducing the risk of crack formation and component failure. Obviously, higher irradiation creep levels tend to relieve more internal stress, thus allowing the components longer useful lifetimes within the core. Determining the irradiation creep rates of nuclear grade graphite is critical for determining the useful lifetime of graphite components and is a major component of the Advanced Graphite Creep (AGC) experiment. The AGC experiment is currently underway to determine the in service behavior of these new graphite grades for HTR and molten salt reactor designs. This experiment will examine properties and behavior of nuclear grade graphite over a large spectrum of temperatures, irradiation fluencies and applied stress levels that are expected to cause irradiation creep strains within a HTR graphite component. Irradiation data are provided through the AGC test series, which is comprised of six planned capsules irradiated in the Advanced Test Reactor (ATR) in a large flux trap located at Idaho National Laboratory (INL). Each irradiation capsule consists of over 400 graphite specimens that are characterized before and after irradiation to determine the irradiation induced material properties changes and life-limiting irradiation creep rate for each graphite grade.
- Research Organization:
- Idaho National Laboratory (INL), Idaho Falls, ID (United States)
- Sponsoring Organization:
- USDOE Office of Nuclear Energy (NE)
- DOE Contract Number:
- AC07-05ID14517
- OSTI ID:
- 2208825
- Report Number(s):
- INL/EXT-16-38044-Rev000
- Country of Publication:
- United States
- Language:
- English
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