A Review on Current Status of Alloys 617 and 230 for Gen IV Nuclear Reactor Internals and Heat Exchangers
暂无分享,去创建一个
[1] Robert I. Jetter,et al. DOE-ASME Generation IV Materials Tasks , 2006 .
[2] P. J. Ennis,et al. Effect of carburizing service environments on the mechanical properties of high-temperature alloys , 1984 .
[3] Weiju Ren,et al. Assessment of Existing Alloy 617 Data for Gen IV Materials Handbook , 2005 .
[4] R. Klueh,et al. The resistance of 9 Cr-1 MoVNb and 12 Cr-1 MoVW steels to helium embrittlement☆ , 1983 .
[5] H. McCoy,et al. Mechanical properties of Inconel 617 and 618 , 1985 .
[6] E. Lara‐Curzio,et al. Advanced Alloys for Compact, High-Efficiency, High-Temperature Heat-Exchangers , 2007 .
[7] G. O. Hayner,et al. Next Generation Nuclear Plant Materials Research and Development Program Plan , 2005 .
[8] Akira Ohtomo,et al. Morphological changes of carbides during creep and their effects on the creep properties of inconel 617 at 1000 °C , 1980 .
[9] Weiju Ren,et al. Gen IV Materials Handbook Implementation Plan , 2005 .
[10] D. Mosedale,et al. Irradiation Creep in Several Metals and Alloys at 100° C , 1967, Nature.
[11] J. Shingledecker,et al. U.S. program on materials technology for ultra-supercritical coal power plants , 2005 .
[12] R. Holt,et al. Impurities and trace elements in nickel-base superalloys , 1976 .
[13] R. M. Boothby,et al. Modelling grain boundary cavity growth in irradiated nimonic PE16 , 1990 .
[14] W. Mankins,et al. Microstructure and phase stability of INCONEL alloy 617 , 1974, Metallurgical and Materials Transactions B.
[15] F. Schubert,et al. Multiaxial creep of tubes from incoloy 800 H and inconel 617 under static and cyclic loading conditions , 1989 .
[16] L. Mansur,et al. Mechanical property changes induced in structural alloys by neutron irradiations with different helium to displacement ratios , 1988 .
[17] J. K. Wright. Next Generation Nuclear Plant Intermediate Heat Exchanger Materials Research and Development Plan (PLN-2804) , 2008 .
[18] E. W. C. Wilkins,et al. Cumulative damage in fatigue , 1956 .
[19] John P. Shingledecker,et al. BOILER MATERIALS FOR ULTRASUPERCRITICAL COAL POWER PLANTS , 2004 .
[20] P. Shewmon. Radiation-Induced Swelling of Stainless Steel , 1971, Science.
[21] D. Harries. Neutron irradiation-induced embrittlement in type 316 and other austenitic steels and alloys , 1979 .
[23] D. Harries. NEUTRON IRRADIATION EMBRITTLEMENT OF AUSTENITIC STAINLESS STEELS AND NICKEL- BASE ALLOYS , 1966 .
[24] W. Ren,et al. A Review of Aging Effects in Alloy 617 for Gen IV Nuclear Reactor Applications , 2006 .
[25] T. Totemeier,et al. Procurement and Initial Characterization of Alloy 230 and CMS Alloy 617 , 2006 .
[26] Weiju Ren,et al. Construction of Web-Accessible Materials Handbook for Generation IV Nuclear Reactors , 2005 .
[27] H. Nickel,et al. Precipitation Behavior of Ni-Cr-22 Fe-18 Mo (Hastelloy-X) and Ni-Cr-22 Co-12 Mo (Inconel-617) After Isothermal Aging , 1984 .
[28] W. Ren. Development of a Controlled Material Specification for Alloy 617 for Nuclear Applications , 2005 .
[29] Weiju Ren,et al. GEN IV MATERIALS HANDBOOK BETA RELEASE FOR STRUCTURAL AND FUNCTIONAL EVALUATION , 2006 .
[30] Jeries Abou-Hanna,et al. Applicability of Simplified Methods to Alloy 617 in Excess of 649°C , 2007 .
[31] R. Cook,et al. Creep rupture behavior of candidate materials for nuclear process heat applications , 1984 .
[32] M. G. Burke,et al. Irradiation-assisted stress corrosion cracking of HTH Alloy X-750 and Alloy 625 , 1995 .