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Material Selection For Long Term Application used in Heat Exchangers in High Temperature Reactors

Material Selection For Long Term Application used in Heat Exchangers in High Temperature Reactors. Catherine Bartholomae MANE6980 - Engineering Project Advisor: Sudhangshu Bose 9/30/10. Abstract.

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Material Selection For Long Term Application used in Heat Exchangers in High Temperature Reactors

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  1. Material Selection For Long Term Application used in Heat Exchangers in High Temperature Reactors Catherine Bartholomae MANE6980 - Engineering Project Advisor: Sudhangshu Bose 9/30/10

  2. Abstract • Very High-Temperature Reactors (VHTR) are being designed to provide new energy options for the future. • The heat exchanger sees temperatures around 1000oC . • Material selection is one of the main challenges in the new concept because of : • Creep Behavior • Fatigue Properties • Environmental Resistance • Important factors besides just material properties: • It is also important to consider fabricability and component behavior: • Workability • Weldability • Non-destructive Testing • And more..

  3. Introduction • This project will focus on the heat exchanger of the VHTR and materials being researched for its high temperature application as well as other requirements such as being environmentally resistant to the highly corrosive coolant that passes through it. • Helium serves as a primary coolant and can contain traces of reactive impurities such as hydrogen, methane, carbon monoxide and water vapor.

  4. Problem Description and Expected Outcomes • To perform a critical review of literature regarding materials expected to be used in the VHTR heat exchanger. • The expected outcome for the project is to gather a detailed analysis of materials found and compile them into one technical paper including recommendations for use in industry.

  5. Current Design • Currently in use are Generation II and Generation III Reactors. Generation III reactors are upgrades or improvements to Generation II designs that have evolved over time. • In this figure: • 1. Charge Tubes • 2. Control Rods • 3. Graphite Moderator • 4. Fuel Assemblies • 5. Concrete pressure • vessel and radiation shielding • 6. Gas circulator • 7. Water • 8. Water Circulator • 9. Heat Exchanger • 10. Steam Advanced Gas-Cooled Reactor

  6. Current Design Continued • The AGR, which uses graphite as a neutron moderator and carbon dioxide as its coolant, was developed from the Magnox Reactor mainly comprised of Magnesium. • Outlet temperature was designed at 648oC. • Figure shows differences in temperatures.

  7. Proposed Design The concept for the VHTR is a graphite-moderated core with a once through uranium fuel cycle. It is expected to have outlet temperatures of ~1000 oC.

  8. Proposed Design Continued • This is one of 6 proposed designs for the Generation IV reactors and is unique because of its capability for hydrogen production as well as energy. • High outlet temperature and helium coolant change material requirements from current design.

  9. Requirements for Proposed System • Melting Point -Material must withstand high temperatures; at least greater than 1000oC outlet temperature. • Creep Resistance -High creep strength is required in this environment. Material properties become weak in the high temperature and failure becomes more likely. • Environmental Resistance- Impurities in the helium can cause high levels of corrosion. Any cracking or pitting resulting from fatigue will only increase oxidation or carburization.

  10. Materials Investigated • Materials that have been researched for Generation IV reactors include: • Nickel Based Alloys • Specifically Inconel 617,Hastelloy X, and Incoloy 800H • Ceramic Materials • Specifically Silicon Carbide/Silicon Nitride Ceramics

  11. Suggested Material Inconel 617 It is widely read that Alloy 617 exhibits high levels of creep strength at high temperatures It has already been approved for applications with temperatures up to 982oC by ASME VIII-Div.1 Annealing becomes in issue if system is started up or shut down. Inconel 617 shows a good combination of thermal conductivity and thermal expansion at 1000oC.

  12. Suggested Material Continued… • As expected maximum allowable stresses decrease with temperature which is expected of a materials mechanical properties.

  13. Suggested Material Continued… • Inconel Corrosion Resistance: • As discussed corrosion is mainly caused by impurities in the coolant, helium. It has been tested at 950oC, Inconel 617 forms an Al2O3 layer which prevents carburization and internal oxidation.

  14. Suggested Material Continued… • Hastelloy X:Similar to Inconel 617 this alloy exhibits high creep strength, exceptional strength, and oxidation resistance. • Typical applications now are in jet engine tailpipes or other aircraft components. It has been proven to maintain it’s properties well for elongated periods of time at high temperatures.

  15. Creep curves in helium environment show similar and lower strain rates to that of Inconel 617 and at much smaller time intervals. Suggested Material Continued…

  16. Suggested Material Continued… • Corrosion Resistance: • Research states that corrosion also affects the ductility and creep rupture life of the material. • Surface Corrosion on all 3 alloys after 500 hrs.

  17. Suggested Material Continued… • Incoloy 800H • High creep strength as well however not as susceptible to the high temperatures Inconel 617 can sustain. • Incoloy 800 - Well characterized and readily available and also in consideration for other proposed reactor designs at lower design temperatures. • Similar to other alloys including Hastelloy X, Incoloy 800 was altered to Incoloy 800H to provide higher creep strength

  18. Compared to other alloys it has the lowest rupture strength. Suggested Material Continued…

  19. Suggested Material Continued… • Environmental Resistance: • At 850oC this alloy form carbides and oxides due to reaction with the coolant. Internal oxidation of Al2O3 is formed about 40 mm below the surface.

  20. Suggested Material Continued… • Ceramic Matrix Composites (CMC) • These composites have been evaluated for use in high temperature applications due to their strength, resistance to high temperatures, and light weight. Most common for high temperature use would be Silicon Carbide (SiC) or Silicon Nitride (SiN) • Creep Rates: • In composites creep occurs by deformation of the grains themselves.

  21. Here we see tensile creep behavior of 3 SiC composites and one SiN composite. Hexaloy (SiC) has highes creep resistance and can be used in materials up to 1500oC. SiN Material has second highest creep resistance and can be used in materials up to 1390oC. Suggested Material Continued…

  22. CMC-When exposed to oxidizing gasses will corrode as shown. CMC Coatings are being researched and have been successful. Suggested Material Continued…

  23. Recommendations • These materials are currently being studied for use in high temperature nuclear application. Through research I have seen that once a researcher recognizes a component of a material that may cause increase creep rupture strength or higher corrosion resistance they are then able to alter the alloy’s composition to increase or decrease a specific element. It is my recommendation that this continue to occur and that these elements become stronger and even surpass the requirements for the harsh environment of a VHTR. • Incoloy 800-Incolloy 800H • Hastelloy X- Hastelloy XR- Hastelloy XR2

  24. Recommendations • Similar to the coatings that were produced for CMC materials I recommend further research into possible coatings for nickel-based allows that will increase corrosion resistance.

  25. Nickel-Based Alloys are the most researched material for this application so far. Based on the knowledge from the research done I’d recommend Inconel 617 as a material to be used in these high temperature nuclear reactors however further research should be done to enhance the alloy to make it stronger. Summary

  26. Milestones • Submit Draft Proposal 9/23/10 • Submit Proposal and Brief Presentation 9/30/10 • Begin research and obtain 12 technical papers 9/30/10 • Attend The Cole Library in Hartford 10/7/10 • Based on prior research continue research of papers found in Hartford 10/14/10 • Submit 1st Progress Report 10/21/10 • Compile all pros and cons thus far for each technical paper 10/28/10 • Begin writing final report and continue research 11/4/10 • Submit 2nd Progress Report and Presentation 11/24/10 • Continue writing report between 11/24/10 and 12/2/10 • Submit Final Draft 12/2/10 • Finalize paper between 12/2/10 and 12/16/10 • Submit Final Report and Presentation 12/16/10

  27. References • "Advanced Gas-cooled Reactor." Wikipedia, the Free Encyclopedia. Web. 01 Oct. 2010. <http://en.wikipedia.org/wiki/Advanced_gas-cooled_reactor>. • Aquaro, D., and M. Pieve. "High Temperature Heat Exchangers for Power Plants: Performance of Advanced Metallic Recuperators." Applied Thermal Engineering 27.2-3 (2007): 389-400. Print. • Cabet, C., and F. Rouillard. "Corrosion of High Temperature Metallic Materials in VHTR." Journal of Nuclear Materials 392.2 (2009): 235-42. Print. • Gan, J., J. Cole, T. Allen, S. Shutthanandan, and S. Thevuthasan. "Irradiated Microstructure of Alloy 800H." Journal of Nuclear Materials 351.1-3 (2006): 223-27. Print. • "Generation III Reactor." Wikipedia, the Free Encyclopedia. Web. 01 Sept. 2010. <http://en.wikipedia.org/wiki/Generation_III_reactor>. • "Generation IV Reactor." Wikipedia, the Free Encyclopedia. Web. 01 Sept. 2010. <http://en.wikipedia.org/wiki/Generation_IV_reactor>. • "Hastelloy X Technical Data." High Temp Materials. Web. 22 Nov. 2010. <www.hightempmaterials.com/techdata/hitempHastXdata.php>. • "High Temp Metals 800-500-2141 - Incoloy 800/800AT/800H Technical Data." High Temp Metals. Web. 02 Dec. 2010. <http://www.hightempmetals.com/techdata/hitempIncoloy800data.php>. • "High Temp Metals 800-500-2141 - Inconel 617 Technical Data." High Temp Metals. Web. 22 Nov. 2010. <http://www.hightempmetals.com/techdata/hitempInconel617data.php>. • Hirano, T., H. Araki, and H. Yoshida. "Carburization and Decarburization of Superalloys in the Simulated Htgr Helium." Journal of Nuclear Materials 97.3 (1981): 272-80. Print. • Kurata, Y., and Y. Ogawa. "Internal Stress during High-temperature Creep of Special Grade Hastelloy X Alloys." Journal of Nuclear Materials 158 (1988): 42-48. Print. •  Li, Xiuqing, David Kininmont, Renaud Le Pierres, and Stephen John Dewson. "Alloy 617 for the High Temperature Diffusion-Bonded Compact Heat Exchangers." (2008): 282-88. Print • Nickel, H., F. Schubert, and H. Schuster. "Evaluation of Alloys for Advanced High-temperature Reactor Systems." Nuclear Engineering and Design 78.2 (1984): 251-65. Print. • Nickel, H., F. Schubert, and H. Schuster. "Very High Temperature Design Criteria for Nuclear Heat Exchangers in Advanced High Temperature Reactors." Nuclear Engineering and Design 94.3 (1986): 337-43. Print.

  28. References Continued… • Nickel, H., F. Schubert, H. Penkalla, and H. Over. "Mechanical Design Methods for High Temperature Reactor Components." Nuclear Engineering and Design 76.3 (1983): 197-206. Print. • Penkalla, H., H. Nickel, and F. Schubert. "Multiaxial Creep of Tubes from Incoloy 800 H and Inconel 617 under Static and Cyclic Loading Conditions." Nuclear Engineering and Design 112 (1989): 279-89. Print. • Peterson, Per, Haihua Zhao, Fenglei Niu, and Wensheng Huang. "Development of C-SiC Ceramic Compact Plate Heat Exchangers for High Temperature Heat Transfer Applications." Print. • Sommers, A., Q. Wang, X. Han, C. T'Joen, Y. Park, and A. Jacobi. "Ceramics and Ceramic Matrix Composites for Heat Exchangers in Advanced Thermal Systems-A Review." Applied Thermal Engineering (2010): 1277-291. ELSEVIER. Web. <http://www.elsevier.com/locate/apthermeng>. • "The Story of the "Incoloy Alloys Series," from 800, through 800H, 800HT." Special Metals. Sept. 2004. Web. 02 Dec. 2010. <www.specialmetals.com>. • Tachibana, Y., H. Sawahata, T. Iyoku, and T. Nakazawa. "Reactivity Control System of the High Temperature Engineering Test Reactor." Nuclear Engineering and Design 233.1-3 (2004): 89-101. Print. • Tan, L., K. Sridharan, and T. Allen. "The Effect of Grain Boundary Engineering on the Oxidation Behavior of INCOLOY Alloy 800H in Supercritical Water." Journal of Nuclear Materials 348.3 (2006): 263-71. Print. • Tsuji, H., and H. Nakajima. "Creep-fatigue Damage Evaluation of a Nickel-base Heat-resistant Alloy Hastelloy XR in Simulated HTGR Helium Gas Environment." Journal of Nuclear Materials 208.3 (1994): 293-99. Print. • Tsuji, H., and T. Kondo. "Strain-time Effects in Low-cycle Fatigue of Nickel-base Heat-resistant Alloys at High Temperature." Journal of Nuclear Materials 150.3 (1987): 259-65. Print. • Tsuji, H., T. Tanabe, Y. Nakasone, and H. Nakajima. "Applicability of Creep Damage Rules to a Nickel-base Heat-resistant Alloy Hastelloy XR." Journal of Nuclear Materials 199.1 (1992): 43-49. Print. • "Very High Temperature Reactor." Wikipedia, the Free Encyclopedia. Web. 30 Sept. 2010. <http://en.wikipedia.org/wiki/Very_high_temperature_reactor>. • Wiederhorn, S., B. J. Hockey, and J. D. French. "Mechanisms of Deformation of Silicon Nitride and Silicon Carbide at High Temperatures." Journal of the European Ceramic Society 19.13-14 (1999): 2273-284. Print. • Yvon, P., and F. Carré. "Structural Materials Challenges for Advanced Reactor Systems." Journal of Nuclear Materials 385.2 (2009): 217-22. Print. • Zhu, S., M. Mizuno, Y. Kagawa, J. Cao, Y. Nagano, and H. Kaya. "Creep and Fatigue Behavior of SiC Fiber Reinforced SiC Composite at High Temperatures." Materials Science and Engineering A 225.1-2 (1997): 69-77. Print.

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