24. MATERIALS FOR ADVANCED NUCLEAR ENERGY SYSTEMS

The Generation IV nuclear energy initiative is an international collaboration to identify, assess, and develop sustainable nuclear energy technologies that are competitive in most markets, while further enhancing nuclear safety, minimizing the nuclear waste burden, and further reducing the risk of proliferation (reference 1). Many nuclear energy systems have been proposed to advance the goals of the Generation IV program (see references 2-8), including designs that use liquid-metal coolants such as sodium and lead, gas coolants such as helium, water coolants such as supercritical water, and molten salt coolants. For these systems, operation at higher temperature has been identified as a means to improve economic performance and to support the thermochemical production of hydrogen. However, the move to higher operating temperatures will require the development and qualification of advanced materials to perform in the more challenging environment. As part of the process of developing advanced materials for these reactor concepts, a fundamental understanding of materials behavior must be established and a database that defines the critical performance limitations of these materials under irradiation must be developed. A recent workshop details many of the research challenges for higher temperature materials associated with proposed Generation IV systems (reference 9). Grant applications are sought only in the following subtopics:

a. Advanced Radiation Resistance Ferritic-Martensitic Alloys—Because of their resistance to void swelling, 9 Cr and 12 Cr ferritic-martensitic steels are considered prime candidates for intermediate temperature reactors such as the proposed liquid metal and supercritical water concepts operating in the temperature range of 400-750°C. However, many ferritic-martensitic steels are limited by poor higher temperature creep strength, typically degrading at temperatures greater than 550-600°C (reference 10). Grant applications are sought to improve the creep strength of 9 Cr and 12 Cr ferritic-martensitic steels through alloying, dispersion strengthening, or precipitation hardening. Innovative alloys with protective coatings are also of interest. Proposed approaches must provide for (1) isotropic creep properties with strength greater than that of Sandvik HT9 steel, (2) a ductile to brittle transition temperature less than room temperature, and (3) a minimum plane-strain fracture toughness of 0.25σ_y. Alloying elements that act as neutron poisons (e.g., boron) or that become highly activated in a neutron spectrum (e.g, cobalt) must be minimized or eliminated. Because the ferritic-martensitic steels likely would be used in conjunction with sodium-cooled, lead- or lead-bismuth-cooled, or supercritical water-cooled reactor concepts, approaches that optimize corrosion performance while achieving improved high temperature strength would be considered high priority. Lastly, approaches that also address irradiation performance are strongly encouraged.

b. Advanced Refractory, Ceramic, Ceramic Composite, or Coated Materials—Some Generation IV concepts aim for very high temperature (>900°C) operation. However, with the exception of limited data on SiC-based systems, the radiation resistance of construction materials subjected to very high temperatures has not been identified or proven. Grant applications are sought to develop advanced refractory, ceramic, ceramic composite, or coated materials that can meet the very demanding conditions required to operate at temperatures greater than 900°C in a fast spectrum nuclear energy system. For these conditions, the materials should have low thermal expansion coefficients, excellent high temperature strength, excellent high temperature creep resistance, and good thermal conductivity. For post-irradiation handling at lower temperatures, sufficient room temperature fracture toughness must be maintained. Additionally, the materials need to be easily fabricated and capable of being joined. Because the reactors operating in this temperature regime are expected to be helium cooled, the materials must have low erosion properties in flowing helium and be able to survive an air ingress condition. Because sustainable nuclear energy systems are likely to be based on fast spectrum systems, the materials must avoid low atomic mass components such as hydrogen and carbon. Because the high temperature strength and corrosion resistance may be difficult to achieve with a single material, composite or coated systems may be required. Finally, because sustainable nuclear energy systems may be based on fast spectrum (i.e., fast flux) designs, materials intended for fast reactor concepts should minimize the use of low atomic mass components such as hydrogen and carbon.

References:

1. Moving Forward: Generation IV Nuclear Energy Systems, U.S. DOE Office of Nuclear Energy, Science and Technology, http://gen-iv.ne.doe.gov

2. Sekimoto, H., et al., “Small Lead-Bismuth-Eutectic (LBE)-Cooled Fast Reactor for Expanding Market,” Proceedings of the Tenth International Conference on Nuclear Engineering (ICONE 10), Arlington, VA, April 14-18, 2002, American Society of Mechanical Engineers (ASME), 2002. (Paper No. ICONE10-22049)*

3. Wade, D. C., et al., “Status of the Encapsulated Nuclear Heat Source (ENHS) Reactor Concept,” Proceedings of ICONE 10, ASME, 2002. (Paper No. ICONE10-22202)*

4. Hejzlar, P., et al., “Design Strategies for a Lead-Bismuth-Cooled Reactor for Actinide Burning and Low-Cost Electricity,” Proceedings of ICONE 10, ASME, 2002. (Paper No. ICONE10-22377)*

5. Kiryushin, A. I. et al., “BN-800—Next Generation of Russian Sodium Fast Reactors,” Proceedings of ICONE 10, ASME, 2002. (Paper No. ICONE10-22405)*

6. Hittner, D., “The Programme and First Results of the European High Temperature Reactor (HTR) Technology Network (HTR-TN),” Proceedings of ICONE 10, ASME, 2002. (Paper No. ICONE10-22423)*

7. King, R. L. and Porter, D. L., “Performance of Key Features of [the] Experimental Breeder Reactor (EBR)-II and the Implications for Next-Generation Reactor Systems,” Proceedings of ICONE 10, ASME, 2002. (Paper No. ICONE10-22524)*

8. Oka, Y. and Koshizuka, S., “Design Concept of Once-Through Cycle Supercritical-Pressure Light-Water-Cooled Reactors,” Proceedings of SCR-2000:International Symposium on Supercritical Water-Cooled Reactors, Design and Technology, Tokyo, Japan, November 6-9, 2000, Tokyo: Tokyo University, July 1, 2000. (ISBN: 4-901332-00-4) (OSTI ID: 20218877) (Abstract available at: http://www.osti.gov/doeecd. Using “Basic Search,” search “Bibliographic Info for “20218877.”)

9. Allen, T., et al., Workshop on Higher Temperature Reactor Materials, La Jolla, CA, March 18-21, 2002, Sponsored by U. S. DOE Office of Nuclear Energy, Science, and Technology and DOE Office of Basic Energy Sciences, August 12, 2002. (Report No. ANL-02/12) (Full text available at: http://www.osti.gov/doeecd/. Using “Advanced Search”, search “Title” and “Author” with information above.)

10. Klueh, R. L. and Harries, D. L., High Chromium Ferritic and Martensitic Steels for Nuclear Applications, West Conshohocken, PA: American Society for Testing and Materials, 2001. (ISBN: 0-8031-2090-7)

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* Abstracts of papers and ordering information available through ASME at: http://store.asme.org/category.asp?catalog%5Fname=Conference+Papers&category%5Fname=Tenth+International+Conference+on+Nuclear+Engineering&Page=1. In SEARCH boxes in upper right corner, search “Conference Papers” for Paper No. given in citation above.

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