9.  MATERIALS FOR NUCLEAR ENERGY SYSTEMS

 

Nuclear power provides over 20 percent of the current U.S. electricity supply without harmful greenhouse gases or air pollutants, including those that may cause adverse global climate changes.  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 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 2-5), including designs that use liquid-metal coolants such as sodium and gas coolants such as helium.  For these reactor concepts, operation at higher temperature has been identified as a means to improve economic performance and/or 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.   Furthermore, in situ materials monitoring sensors that can operate in the high neutron fluence irradiation damage environment for sodium fast reactors, and also can perform well at extreme temperatures (> 900 C) for high-temperature gas-cooled reactors for many years, will be needed to provide non-destructive evaluation methods for continuous or periodic surveillance during normal plant  operations and accident conditions.

 

In addition, to maintain the security of the nation’s 104 existing nuclear power plants, research on materials aging and degradation is needed.  The safe and reliable operation of nuclear power plants operating in an extended lifetime will require a high degree of confidence in the reliability of nuclear power plants systems, structures, and components.  The Light Water Reactor Sustainability Program (reference 7) seeks to develop the fundamental scientific knowledge basis to understand and predict changes in the systems, structures, and components and their materials as they age in a nuclear power plant environment. 

 

Applications that require the handling of radioactive specimens may propose to use the Idaho National Laboratory Advanced Test Reactor National Scientific User Facility and its hot cells, or the Oak Ridge National Laboratory High Flux Isotope Reactor and its radiological examination facilities.  Hot cell facilities at Argonne National Laboratory, Pacific Northwest National Laboratory, and Los Alamos National Laboratory also may be considered.  Grant applications are sought only in the following subtopics.

 

a. Advanced Radiation Resistant Ferritic-Martensitic Alloys and Oxide Dispersion Strengthened (ODS) Steels—Because of their resistance to void swelling, ferritic-martensitic and ODS steels are considered prime candidates for intermediate temperature applications, such as the proposed liquid metal reactor concept operating in the temperature range 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 6), and ODS steels are difficult to form and weld.  Grant applications are sought to improve the creep strength of ferritic-martensitic steels through alloying, dispersion strengthening, or precipitation hardening.

 

Grant applications also are sought to improve the weldability and formability of ODS steels.  Innovative alloys with protective coatings also are 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 0.25σy.  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 and ODS steels likely would be used in conjunction with a sodium-cooled reactor concept, approaches that optimize corrosion performance while achieving improved high-temperature strength would be considered high priority. 

 

Lastly, grant applications are sought to develop approaches for monitoring these important materials properties in ferritic-martensitic and ODS steels when they are used as in-reactor materials and core barrel/vessel components.  Of particular interest are methods that can (1) measure in situ irradiation performance; and (2) provide data, in conjunction with non-destructive evaluation techniques, that could potentially yield in situ monitoring capability for core/vessel materials performance and detect incremental changes in mechanical properties. 

 

Questions – contact Sue Lesica (sue.lesica@hq.doe.gov)  

 

b. Advanced Refractory, Ceramic, Ceramic Composite, Graphitic, or Coated Materials—Generation IV Advanced Gas Cooled Reactors (Next Generation Nuclear Plant (NPNG), reference 4. ) 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, graphitic, or coated materials that can meet the very demanding conditions required to operate at temperatures greater than 900°C in a thermal 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 the high temperature strength and corrosion resistance may be difficult to achieve with a single material, composite or coated systems may be required. 

 

Furthermore, use of these advanced irradiation-resistant high temperature materials for gas reactor (NGNP) applications also will require advanced methods for periodic and, eventually, real-time monitoring capability during extreme temperature and flux service conditions.  Grant applications are sought to develop advanced methods that can measure the in situ irradiation performance of these NGNP refractory, ceramic, graphitic, and coated composite materials.  Of particular interest are grant applications for sensors that can (1) monitor the mechanical properties of NGNP in-core/in-vessel materials as they change during their service lifetime and (2) provide accurate and reliable measurements of material mechanical properties during the large temperature changes that occur as the plant operates.

 

Questions – contact Sue Lesica (sue.lesica@hq.doe.gov)  

 

c. Advanced Technologies for the Assessment and Mitigation of Materials Degradation for Light Water Reactor Systems and Components—Extending the service-life of the current light water reactor fleet will require a high degree of confidence in understanding and predicting materials performance.  New technologies and advances are needed for the long term (e.g., beyond current licensing periods of 60 years) characterization and repair of materials systems experiencing a nuclear power plant environment, and to develop improved methodologies for assessing risk and uncertainty.  Therefore, grant applications are sought to develop and demonstrate (1) advanced in situ techniques for fundamental phenomenological aging characterization of nuclear-related materials, such as swelling in stainless steel, hardening of reactor pressure vessels and the degradation of concrete; (2) advanced welding techniques for component repair; (3) techniques and processes to mitigate or predict irradiation effects, or other aging phenomena experienced in nuclear reactor components; (4) advanced nuclear fuel cladding materials; and (5) databases and methodologies for assessing risk and uncertainty associated with materials degradation of Light Water Reactor components. 

 

Questions – contact Robert Jordan (robert.jordan2@hq.doe.gov) 

 

References:

 

1.      “Generation IV Nuclear Energy Systems,” U.S. DOE Office of Nuclear Energy, Science and Technology Website.  (URL:  http://nuclear.energy.gov/genIV/neGenIV1.html)

 

2.      “Global Nuclear Energy Partnership,” U.S. DOE Office of Nuclear Energy, Science and Technology Website  (URL:  http://www.gnep.energy.gov)

 

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

 

4.       Hayner, G. O., et al, Next Generation Nuclear Plant Materials Research and Development Program Plant,” Idaho National Laboratory, INL/EXT-06-11701,  Revision 3 August 2006 (URL:  http://nuclear.inl.gov/deliverables/docs/ngnp_materials_program_plan.pdf)

 

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

 

6.      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)

 

7.      “Life Beyond 60 Workshop Summary Report”, Nuclear Regulatory Commission/Department of Energy, 2008

(URL: http://nuclear.energy.gov/pdfFiles/LifeAfter60WorkshopReport.pdf )

 

 

* 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. 

 Search by Paper No. in citation above.)