Studying silicon carbide for nuclear fuel cladding

19 April 2013



With its high melting point and low oxidation rate, silicon carbide remains stable even in nuclear accident scenarios. A wide-ranging R&D programme in the USA is now underway to assess the feasibility of an SiC nuclear fuel cladding. By Shannon Bragg-Sitton, Kristine Barrett, Isabella van Rooyen, David Hurley and Marat Khafizov.


Prior to March 2011, the emphasis for advanced fuel R&D activities was on improving the fuel performance for waste minimization and increased power density, as well as collaborating with industry on fuel reliability. Following the event at Fukushima in 2011, enhancing the accident tolerance of light water reactors (LWRs) became a topic of serious discussion in the United States.

The US Department of Energy Office of Nuclear Energy (DOE-NE) is conducting research and development on enhanced accident-tolerant fuels (ATF) for light water reactors (LWRs). This work is being carried out through the Light Water Reactor Sustainability (LWRS) Program, Advanced Nuclear Fuels Pathway, and Fuel Cycle Research and Development (FCRD) Program Advanced Fuels Campaign (AFC).

A central goal of these coordinated programmes is to develop novel fuel and cladding concepts to replace the current zirconium alloy-uranium dioxide (UO2) fuel system (Figure 1). ATF concepts should provide improved performance, reliability and safety characteristics during normal operations and accident conditions, while minimizing waste generation, relative to the current UO2-zirconium alloy system. The initial effort focuses on implementation in operating reactors or reactors with design certifications.

The work on accident-tolerant fuels is conducted by the Department of Energy laboratories in conjunction with industry and academia. Currently, there are three industry awards being funded by the DOE, with cost sharing from industry, to develop accident tolerant fuel and cladding. The three industry teams are led by Westinghouse, AREVA, and GE Global Research, but each team includes a large number of research partners from industry, academia and DOE laboratories. Additionally, there are university teams that are funded through the US DOE Office of Nuclear Energy under Integrated Research Proposals to develop accident-tolerant fuel and reactor concepts. These teams are led by the University of Tennessee, the University of Illinois, and the Georgia Institute of Technology, again with a large number of research partners on each team.

Fuels with enhanced accident tolerance would be able to bear loss of active cooling in the reactor core during design-basis and beyond design-basis events for a considerably longer time period than current fuel designs. (No specific grace period target has been defined, nationally or internationally, although the consensus of a recent international meeting on ATF is that this should be measurable on the order of hours rather minutes.) At the same time ATF would also maintain or improve fuel performance during normal operations and operational transients.

Key design objectives for fuels and claddings with enhanced accident tolerance include improved reaction kinetics with steam and reduced hydrogen generation. ATF would also need to demonstrate acceptable thermo-mechanical properties, fuel-cladding interactions and fission-product retention under extreme conditions. Any new fuel concept proposed must also comply with current operational, economic and safety requirements, and fit in with current fuel cycle impacts and current LWR design constraints.

In order to achieve significant operating improvements over standard zirconium alloy clad nuclear fuel while remaining within safety boundaries, significant steps beyond incremental improvements in the current generation of nuclear fuel are required. Fundamental improvements are needed in the areas of nuclear fuel composition, cladding integrity, and the fuel/cladding interaction to allow power uprates and increased fuel burn-up while potentially improving safety margin through the adoption of an 'enhanced accident tolerant' fuel system.

Research and development efforts for top-ranked cladding design, materials and fabrication concepts will include a rigorous series of mechanical, thermal and chemical characterization tests to assess operating potential in a relatively low-cost, nonnuclear test series. Promising options would then proceed to irradiation in a test reactor such as the Idaho National Laboratory (INL) Advanced Test Reactor (ATR) (Figure 2).

Various advanced cladding concepts are currently under investigation. These include:

corrosion-resistant coatings applied to existing zirconium-based cladding

hybrid ceramic/metal cladding

complete replacement of the conventional nuclear fuel cladding material (for example with engineered zirconium alloys, advanced steels, or fully-ceramic designs) [1].

Several leading candidates for advanced nuclear fuel cladding incorporate silicon carbide (SiC) (see also NEI January 2010, pp.14-16). Although a potentially long-range development timeline may be required, SiC has the potential to offer significant performance improvements over zirconium-alloy cladding.

Research into SiC concepts is widespread, including US DOE national laboratories (namely INL and Oak Ridge National Laboratory) [2-8], industry [9-13], and international organizations [14-18].

Concepts include:

Fully-ceramic SiC/SiC cladding that might include SiC fibre ceramic matrix composite (CMC) and monolithic SiC layers

SiC CMC, metal hybrid cladding, for example layering of CMC and metallic components.

Some researchers are also investigating the use of SiC CMC for channel boxes in a boiling water reactor [19] as a nearer-term LWR application of SiC relative to fuel cladding.

For any of these concepts, additional data on the specific SiC material and fabrication technique is necessary to better characterize its potential performance in a nuclear application. SiC design concepts should be considered high-risk, (potentially) high-reward options for nuclear fuel cladding given the current status of technology gaps associated with use in nuclear environments.

Why SiC?

Significant data is available on SiC from advanced nuclear fuel research, fusion research and the aerospace industry (where it is used in sensors and electronic circuits). SiC has demonstrated exceptionally low oxidation rates up to 1700°C and has been shown to withstand temperatures exceeding 2500°C (SiC does not melt and has a very high sublimation temperature of ~2700°C). Currently available data on monolithic SiC samples indicates oxidation rates that are two to three orders of magnitude lower than that of zirconium-based alloys. SiC oxidation kinetics, coupled with lower heat of oxidation, translates to lower maximum temperatures under postulated accident conditions and significant reduction in hydrogen generation.

Monolithic SiC can be fabricated into a hermetically sealed tube, but it is very brittle. A composite structure, on the other hand, would provide improved strength and fracture toughness, but maintaining hermeticity is a current challenge. Hence, layered options are under investigation to allow a cladding design to take advantage of each of these properties.

SiC CMC is a very strong, high-temperature ceramic material that is also chemically nonreactive. It is made up of very fine filaments (~10 micron) in the form of multi-filament "tows." They are woven or braided into a tubular form to enhance the SiC cladding structural strength while mitigating the brittle nature of monolithic SiC.

The process of forming a thin-walled CMC tube uses textile methods of continuous fibre braid lay-up (preforming) or filament winding over a mandrel. This is followed by formation of a very thin (sub-micron) interface layer between the fibres and adjacent ceramic matrix, and subsequently by the process to form the SiC ceramic matrix.

A properly-engineered interface allows those cracks propagating through the matrix to cause fibre/matrix debonding, thus arresting the crack, but is strong enough to take full advantage of the significant strength inherent to SiC fibres. The interface layer can consist of a number of materials such as pyrolytic carbon (PyC), oxide ceramics, or boron nitride (BN). Performance of such layers under prototypic LWR fluence and temperature must be demonstrated.

There are multiple industrial processes for forming the SiC ceramic matrix surrounding the continuous ceramic fibres. The most common processes under investigation for nuclear applications include:

Chemical Vapor Infiltration (CVI) of the SiC. This can be done using isothermal or temperature-gradient and forced-flow, isobaric or pulsed-flow methods

Pre-ceramic liquid Polymer Impregnation and Pyrolysis (PIP) formation followed by elevated temperature conversion to SiC

Nano-Infiltration and Transient Eutectic-phase (NITE) formation of the SiC matrix using the transient liquid phase-assisted pressure sintering process.

An excellent review of these processing methods has been published [20]. In each process the resulting local chemical bond between Si to C is the same. However, each process needs to be controlled to the desired crystalline phase (beta or alpha) and to achieve a Si/C ratio equal to 1. Variations in the local Si/C ratio (greater than or less than 1, free Si or free C) can negatively affect the final material properties.

Testing needs

Technology development testing will have a strong focus on closing the technology gaps and fully characterizing material properties for each candidate cladding technology. This will be done through a suite of nonnuclear material property measurements, characterization tests, and limited irradiation tests of sample coupons or short cylindrical sections.

All environmental conditions relevant to an operating LWR (such as water flow rate, temperature, chemistry, etc.) can be simulated in a nonnuclear environment. This will allow early characterization of material corrosion behaviour, strength, conductivity, etc. in the absence of radiation. Similarly, an environment simulating steam conditions during a loss of coolant accident (LOCA) can be established in a nonnuclear test laboratory. Small-scale irradiation experiments, such as irradiation testing of material coupons or sealed cladding in a drop-in capsule, will be used for preliminary investigation of irradiation performance of candidate cladding materials [2].

Tests conducted to determine material properties and performance must be performed in accordance with existing standards set by the American Society for Testing and Materials (ASTM). If a standard is not available, as is the case for some of the options under consideration, investment will be required to establish such a standard. Initial survey of the test and characterization facilities across the DOE complex suggests that all the equipment and laboratories necessary for preliminary characterization of the advanced cladding designs (such as hot cells) are currently available. Many of the test techniques are also available in-cell to allow for post-irradiation materials characterization.

Characterization of specific SiC fabrication methods and assembled cladding designs is necessary to determine their applicability for nuclear applications. Specific recommendations for SiC characterization and testing activities are described here. Some efforts are currently underway, while others are in the planning stages.

Thermal properties

It is well known that the thermal conductivity of SiC degrades rapidly due to the production of point defects created by neutron irradiation [21-24]. For advanced cladding designs, a reduction in thermal conductivity leads to decreased fuel thermal margin and exacerbates issues associated with creep deformation. It is thus important to engineer SiC cladding materials that have a high baseline thermal conductivity (unirradiated). To assist in this task there is need for a characterization approach that relates the fundamental response of the constituents to the overall performance of the composite. This will not only help reveal how different fabrication processes influence thermal conductivity, but will also provide crucial information for mesoscopic-scale thermo-mechanical and microstructure evolution models currently being developed.

For composites, one must consider the thermal properties of both the matrix and fibres. SiC fibres, even stoichiometric fibres, have relatively low thermal conductivity compared to monolithic SiC. As a consequence, the matrix is expected to transport the majority of the thermal energy. The matrix material is typically deposited using either CVI or PIP, as discussed above. Both deposition processes can result in matrix material having a thermal conductivity that is substantially lower than that of the bulk material. INL researchers have applied a spatially resolved laser-based technique, termed modulated thermo-reflectance (MTR), to isolate and quantify thermal transport in the matrix of a CMC sample [25, 26].

Baseline laser flash thermal diffusivity measurements were performed on two samples -- monolithic SiC and a SiC CMC composed of Nicalon Type S fibres -- to initiate the thermal transport study. As shown in Figure 3, the monolithic sample had a considerably higher thermal diffusivity than the CMC. Results for a room temperature measurement of diffusivity for the CMC matrix is also reported in Figure 3. An optical micrograph of the CMC sample surface is also shown (Figure 4). The region probed (~20 × 20 µm) is demarcated by the dashed square. The thermal diffusivity of th­e matrix material, 14.5 mm2/s, is considerably lower than the room temperature diffusivity for the monolithic sample. The relatively low value of the CMC diffusivity can be explained by thermal resistance caused by the interface between the fibre and matrix and reduced diffusivity of the matrix material in comparison to the monolithic sample.

Oxidation kinetics

Zirconium alloys undergo significant reaction with reactor coolant leading to materials loss, growth of a low-conductivity oxide phase on the cladding surface, hydriding of the cladding interior, and related loss of materials ductility under normal operating conditions. Under LOCA conditions, zirconium alloys can undergo phase transition, loss of strength, exothermic reaction with steam, and associated hydrogen production.

It is accepted that SiC will react more slowly than zirconium-based alloys with steam under LOCA or beyond-design-basis-accident conditions. Near atmospheric pressure, the reaction of steam with zirconium-based alloys and SiC [27] has been extensively studied and is well understood. While it is well known that metallic materials have a linear pressure dependence of mass loss, this dependence is generally not important over the pressure range associated with reactor transients (for zirconium-based alloys) and can be ignored. However, as SiC has the potential for substantially greater performance than zirconium-based alloys, it becomes more important to understand the projected pressure/temperature of any beyond-design-basis-accident and the physical mechanism of SiC reactions.

The relative attractiveness and benefit of SiC cladding (the ultimate economic driver) will depend on the quantitative determination of cladding performance. For this reason an understanding of the cladding performance under LOCA and/or design basis accident conditions is required. Oak Ridge National Laboratory (ORNL) is currently conducting studies of the oxidation kinetics of SiC exposed to high-temperature steam (1200-1700°C) at 0.34-2.07 MPa at both low flow (~1 cm/s) and high flow (~2 m/s) conditions.

Significant oxidation data has been collected for monolithic chemical vapor deposition (CVD) SiC, but only limited data is currently available for composite SiC/SiC. As the current cladding concepts are overcoated with CVI SiC, it is expected that their behaviour in steam will be very similar to that observed for monolithic samples. Recent measurements for SiC/SiC CVI composites at ORNL have shown a pressure dependence of the oxidation rate (material recession rate increases with increasing pressure), but the oxidation kinetics are 2-3 orders of magnitude slower than for Zr alloys [28]. Additional testing in steam will be used to develop correlations for steam oxidation kinetics for SiC composites.

Fatigue testing

Behaviour of composites under cyclic loading (that is, experiencing regularly-recurring stresses) in oxidizing conditions must be understood in order to use these materials in nuclear fuel cladding tubes. It is expected that composite degradation will be further increased by cyclic loading, especially at conditions near the matrix microcracking regime, as shown in previous aerospace industry research [29, 30]. The effects of cyclic oxidation exposure are of specific interest due to the opening and contraction of micro-cracks in the matrix during a fatigue cycle. This cycling may increase the rate of degradation of a composite tube.

A fatigue study is currently proposed to investigate the fatigue properties already available for composites, with reference to the fabrication method, fatigue cycle parameters and test environments. This study (currently in the planning stages, with testing expected in the October 2013-September 2015 timeframe) will provide a baseline to assess any gaps in fatigue data available for some of the candidate materials considered for use as cladding.

Additionally, a fatigue testing method for short mock-up sample tubes will be designed to study the potential cladding system performance under cyclic conditions; only the inherent material properties can be determined from testing of mini-material coupons. Design parameters will include varying loading frequencies, temperature and test environments (for example in steam, air, argon, etc.). Testing may include both compression-tension loading and tension-tension loading experiments. Experimental data may then be used to validate computational models and for comparison of candidate cladding designs for accident-tolerant fuels. It is expected that the composite microstructure, damage and failure mechanisms will be identified for each test condition. Results will further show the possible impact of fabrication method on the failure mechanisms, further assisting design and modelling efforts.

Material characteristics such as surface roughness may also influence the oxidation reaction kinetics and fatigue properties. Preliminary plans are in place to measure macro-surface roughness for PIP fabricated SiC-CMC sleeves intended for a hybrid cladding design before October 2013.

Joining technology

Cladding designs for nuclear fuels require a hermetic structure and end-cap seals that can withstand the radiation, temperature and chemical environment inherent to an operating LWR. The end-cap seal for the fully ceramic system requires sealing of the SiCf CMC to itself. A reliable, reproducible technique to join and hermetically-seal silicon carbide composites has been identified as a critical technology gap for SiC-based cladding systems.

There are a number of conventional and advanced techniques to join SiC (or SiC/SiC) to itself or other materials (see [31-33] for a summary of these techniques). Successfully-demonstrated techniques include pre-ceramic polymer joining, glass-ceramics, reaction bonding, active metal/pre-ceramic polymers, and active metal solid state displacement techniques. While the strength of the joints produced by these methods appears to be adequate for LWR applications, there is currently a lack of standards for testing ceramics, and a variety of tests have been used to measure the strength of the bonds created using each technique.

There is currently limited irradiation data on SiC/SiC joints and materials used to fabricate the joints, and the joint fabrication techniques that have been tested under irradiation have demonstrated poor irradiation stability. Hence, a reliable SiC/SiC joining technique for reactor structural materials has yet to be developed and is considered the leading technology gap for SiC/SiC composite application to nuclear fuel cladding. Given the functional requirement of hermeticity for cladding, necessary to retain helium and gaseous fission products, the SiC/SiC joining technique must be radiation stable for the relevant conditions of applied stress (to be defined), temperature (~400-500°C) and neutron damage (~6 dpa).

Several methods of joining SiC ceramic composites are considered promising for general applications; however, not all are expected to hold promise for in-reactor applications. Primary considerations for nuclear applications include resistance to neutron irradiation; mechanical properties, such as strength and reliability during mechanical loading; compatibility of the processing condition with the design requirement; chemical compatibility with the operating environment for the intended application; and the ability to satisfy the hermeticity requirement.

Methods for joining SiC CMC materials for application in LWR nuclear fuel cladding are currently being developed at ORNL [34] and via industry collaboration with General Atomics and HyperTherm HTC under the LWRS program. nonnuclear characterization of SiC/SiC joints for end-cap sealing is being performed in 2013; initial irradiation of promising SiC/SiC joints under LWR conditions may begin in early 2014 under the FCRD Advanced Fuels Campaign.

Irradiation performance

The INL, through the Department of Energy Idaho Operations Office, has been assigned the responsibility of irradiating candidate accident-tolerant fuels in the advanced test reactor under the FCRD LWR Fuels program. The ATF irradiation experiments will be drop-in capsule experiments scheduled to start irradiation testing in mid-2014. SiC cladding is just one of several ATF concepts that will be tested under the LWR fuels programme.

The team aims to achieve 10 dpa to verify performance of SiC joints, but it plans to pull out and examine samples at multiple points to better understand the material's irradiation behaviour. Full demonstration would require multiple irradiation cycles (on the order of years).

The ATR is the highest-power test reactor operating in the US and has larger test volumes in high flux areas than any other reactor. It is controlled using a combination of control cylinders outside the core and neck shim rods within the core. The control cylinders rotate hafnium plates toward and away from the core, and the shim rods are either fully inserted or fully withdrawn, allowing reactor power within a cycle to be controlled with very little axial perturbation. Within bounds, the power level in each corner lobe of the reactor can be controlled independently [35]. The ATR's unique design and control devices permit large power differences to be maintained among the nine flux traps, allowing the irradiation environment within each to be uniquely tailored to experiment requirements.

The ATF experiment assembly consists of an aluminium experiment basket, a 316L stainless steel pressure boundary capsule, and fuel or material test specimens encapsulated inside a rodlet. The experiment basket is designed to hold multiple vertically stacked capsule assemblies per ATR position. Each stainless steel capsule will contain a single fuel rodlet or stack of material specimens. The capsules will be exposed to a near prototypical thermal neutron flux and will remain in the test positions until sufficient burnup is reached (near lifetime of typical LWR bundles), at which time they will be removed and shipped to the INL Materials and Fuels Complex (MFC) for post-irradiation examination [36].

Conclusions

SiC-based cladding designs may hold significant potential for use in accident tolerant nuclear fuel concepts, but significant development is still required, making it a 'high-risk, (potentially) high-reward' technology. Available material properties data, testing under steam exposure and post-steam exposure, and performance under irradiation have shown promising results. Additional data and testing are necessary to close technology gaps. Once feasibility is established, qualification for use as nuclear fuel cladding will require much more substantial testing and characterization.

 

References

1. K. Barrett, S. Bragg-Sitton, D. Galicki, Advanced LWR Nuclear Fuel Cladding System Development Trade-off Study, INL/EXT-12-27090, September 2012.

2. S. Bragg-Sitton, Advanced LWR Nuclear Fuel Cladding System Development: Technical Program Plan, INL/MIS-12-25696 Rev. 1, December 2012. (See www.inl.gov/lwrs, Program Documents)

3. I.J. van Rooyen, Pre-Irradiation Testing and Analysis to Support the LWRS Hybrid SiC-CMC-Zircaloy-4 Unfueled Rodlet Irradiation, INL/EXT-12-27189, January 2013.

4. K. M. McHugh et al., High Temperature Steam Corrosion of Cladding for Nuclear Applications: II. Experimental, paper ICACC-S13-018-2013, 37th International Conference & Exposition on Advanced Ceramics and Composites, Daytona Beach, FL, Jan 27 - Feb 1, 2013.

5. Snead, L. L., T. Nozawa, Y. Katoh, T.-S. Byun, S. Kondo and D.A. Petti, "Handbook of SiC properties for fuel performance modeling," Journal of Nuclear Materials. 2007. 371(1-3): p. 329-377.

6. Newsome, G. A., L.L. Snead, T. Hinoki, Y. Katoh and D. Peters, "Evaluation of neutron irradiated silicon carbide and silicon carbide composites," Journal of Nuclear Materials. 2007. 371(1-3): p. 76-89.

7. Cheng, T., J.R. Keiser, M.P. Bradfy, K.A. Terrani and B.A. Pint, "Oxidation of fuel cladding candidate materials in steam environments at high temperature and pressure," J. Nucl. Mat. 2012. 427(1-3): p. 396-400.

8. Katoh, Y., L. L. Snead, T. Nozawa, S. Kondo and J. T. Busby, "Thermophysical and mechanical properties of near-stoichiometric fiber CVI SiC/SiC composites after neutron irradiation at elevated temperatures," Journal of Nuclear Materials. 2010. 403(1-3): p. 48-61.

9. Hallstadius, L., S. Johnson and E. Lahoda, "Cladding for high performance fuel," Progress in Nuclear Energy. 2012. 57: p. 71-76.

10. Deck, C. P., H. E. Khalifa, and C. S. Back, 2012. Effects of Structure and Processing on the Thermal Conductivity of SiC-SiC Composites, Transactions of the American Nuclear Society, Vol. 106, pp 1123-1125, June 2012.

11. H. Serizawa, et al., Numerical analysis of mechanical testing for evaluating shear strength of SiC/SiC composite joints, Journal of Nuclear Materials. 2007. 367(b): p. 1223-1227. (Ceramatec)

12. E.D. Herderick, K. Cooper, N. Ames, New Approach to Join SiC for Accident-Tolerant Nuclear Fuel Cladding, Advanced Materials & Processes. 2012. 170 (1): p. 24-27.

13. PWR Cores with Silicon Carbide Cladding. Palo Alto, CA : EPRI, 2011. 1022908 (export controlled document).

14. C. Chateau, et al., In situ X-ray microtomography characterization of damage in SiCf/SiC minicomposites, Composites Science and Technology. 2011. 71: p. 916-924.

15. W. Kim, D. Kim, J. Park, Environmental Effect on the Oxidation of CVD SiC Ceramics and Composites, paper ICACC-S13-032-2013, 37th International Conference & Exposition on Advanced Ceramics and Composites, Daytona Beach, FL, Jan 27 - Feb 1, 2013.

16. K. Ozawa, et al., Tensile and Interfacial Properties of Unidirectional Advanced SiC/SiC Minicomposites, paper ICACC-S13-032-2013, 37th International Conference & Exposition on Advanced Ceramics and Composites, Daytona Beach, FL, Jan 27 - Feb 1, 2013.

17. Hirayama,H., T. Kawakubo, A. Goto and T. Kaneko, "Corrosion Behavior of Silicon Carbide in 290°C Water," J. Amer. Ceram. Soc. 1989. 72(11): p. 2049-2053.

18. Kim, W.J., H.S. Hwang, J.Y. Park and W.S. Ryu, "Corrosion behaviors of sintered and chemically vapor deposited silicon carbide ceramics in water at 360 ?C," Journal of Materials Science Letters. 2003. 22: p. 581-584.

19. Yueh, K., et al., Silicon Carbide Composite for BWR Channel Applications, in proceedings of Top Fuel 2012, Manchester, United Kingdom, Sept 2-6 2012.

20. Naslain, R., "Design, preparation and properties of non-oxide CMCs for application in engines and nuclear reactors: an overview," Composites Sci. and Tech. 2004. 64(2): p. 155.

21. M. Rohde, Reduction of the thermal conductivity of SiC by radiation damage, J. Nucl. Mater. 1991. 182: p. 87-92.

22. L.L. Snead, Limits on irradiation-induced thermal conductivity and electrical resistivity in silicon carbide materials, J. Nucl. Mater. 2004. 329-333(A): p. 524-529.

23. L.L. Snead, S.J. Zinkle, D.P. White, Thermal conductivity degradation of ceramic materials due to low temperature, low dose neutron irradiation, J. Nucl. Mater. 2005. 340: p. 187-202.

24. Y. Katoh et al., Thermophysical and mechanical properties of near-stoichiometric fiber CVI SiC/SiC composites after neutron irradiation at elevated temperatures, J. Nucl. Mater. 2010. 403: p. 48-61.

25. M. Khafizov, D. Hurley, Measurement of thermal transport using time-resolved thermal wave microscopy, J. Appl. Phys. 2011. 110: 83525.

26. D.H. Hurley, M. Khafizov, S. Shinde, Measurement of thermal transport using time-resolved thermal wave microscopy, J. Appl. Phys. 2011. 109: 83504.

27. E.J. Opila, "Variation of the Oxidation Rate of Silicon Carbide with Water-Vapor Pressure," Journal of the American Ceramic Society. 1999. 82(3): p. 625-636.

28. L.L. Snead, K. Terrani and T. Cheng, Oak Ridge National Laboratory, not yet published.

29. M.B. Ruggles-Wrenn, TP Jones, Tension-compression fatigue of a SiC/SiC ceramic matrix composite at 1200C in air and in steam, International Journal of Fatigue. 2013. 47: p. 154-160.

30. M.B. Ruggles-Wrenn, DT Christensen, AL Chamberlain, JE Lane, TS Cook, The effect of frequency and environement on fatigue behavior of a CVI SiC/SiC ceramic matrix composite at 1200C, Composite Science and Technology. 2011. 71: p. 190-196.

31. Y. Katoh, L.L. Snead, C.H. Henager, T. Hinoki, M. Ferraris, S.T. Gonczy, Joining Silicon Carbide for Advanced LWR Fuel Cladding, Trans. Ameri. Nucl. Soc., (2012).

32. M. Ferraris, M. Salvo, V. Casalegno, S. Han, Y. Katoh, H.C. Jung, T. Hinoki, A. Kohyama, Joining of SiC-based materials for nuclear energy applications, J. Nucl. Mater., 417 (2011) 379-382.

33. Y. Katoh, et al., (to be) submitted to J. Nucl. Mater.

34. Y. Katoh, T. Cheng, J. O. Kiggans, Jr., J. L. McDuffee, and L. L. Snead, Status of Irradiation Test Preparation Activities for Silicon Carbide Joining and Irradiation Studies, ORNL/TM-2012/597, Oak Ridge National Laboratory, Oak Ridge, 2012.

35. INL/EXT-08-14709, "FY2009 Advanced Test Reactor National Scientific User Facility User's Guide".

36. K.E. Barrett, "Project Execution Plan for the Fuel Cycle Research and Development LWR Fuels Irradiation Experiments in the ATR", INL document number PLN-4401, February 2013.

 

Figure 3
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Figure 1


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