journal article·2016-12Open access

Motivation for utilizing new high-performance advanced materials in nuclear energy systems

S.J. Zinkle, K.A. Terrani, L.L. Snead

Abstract

Abstract Despite the very demanding operational environment in nuclear reactors, there have been relatively few advanced high-performance materials introduced into fission reactors during the past 50 years. Some of the regulatory and operational barriers to the introduction of high performance materials are briefly discussed, and several examples of potential improvement in current and planned fission reactor systems that could be enabled by advanced structural materials for in-core applications are outlined. Enhanced non-proprietary public-private research and development on advanced structural materials could yield numerous performance, economic, environmental and safety benefits for current boiling water and pressurized water reactors as well as future Generation IV reactor systems such as sodium cooled fast reactors or very high temperature gas cooled reactors.
Source: openaire · Quality: candidate unverified · Source record ↗

Research context

OpenAlex topics: Nuclear Materials and Properties · Nuclear reactor physics and engineering · Fusion materials and technologies

SciVX discovery tags are terminology-based. OpenAlex topics are source-provided classifications.

References (100+)

Show source-provided references
1. Greater tolerance for nuclear materials10.1038/nmat2266 ↗
2. Challenges in materials research for sustainable nuclear energy10.1557/mrs2008.67 ↗
3. Structural materials challenges for advanced reactor systems10.1016/j.jnucmat.2008.11.026 ↗
4. Structural materials for fission and fusion energy10.1016/S1369-7021(09)70294-9 ↗
5. Materials challenges in nuclear energy10.1016/j.actamat.2012.11.004 ↗
6. Structural materials for next generation nuclear systems: challenges and the path forward10.1016/j.nucengdes.2015.09.015 ↗
7. Hankey · 2016
8. Advanced oxidation-resistant iron-based alloys for LWR fuel cladding10.1016/j.jnucmat.2013.06.041 ↗
9. Embrittlement of nuclear reactor pressure vessels10.1007/s11837-001-0081-0 ↗
10. Rensman · 2005
11. Developing structural, high-heat flux and plasma facing materials for a near-term DEMO fusion power plant: the EU assessment10.1016/j.jnucmat.2014.06.014 ↗
12. Reactor pressure vessel (RPV) design and fabrication: the case of the USA
13. US program on materials technology for ultra-supercritical coal power plants10.1007/s11665-013-0717-6 ↗
14. Materials considerations for aerospace applications10.1557/mrs.2015.278 ↗
15. Composite structures: the first 100 years
16. Computational design of hierarchically structured materials10.1126/science.277.5330.1237 ↗
17. THERMO-CALC & DICTRA, computational tools for materials science10.1016/S0364-5916(02)00037-8 ↗
18. Precipitate design for creep strengthening of 9% Cr tempered martensitic steel for ultra-supercritical power plants10.1088/1468-6996/9/1/013002 ↗
19. Effect of alloying additions on phase equilibria and creep resistance of alumina-forming austenitic stainless steels10.1007/s11661-009-9886-1 ↗
20. Alloy design of creep- and oxidation-resistant 9%Cr steel for high efficiency USC power plant10.4028/www.scientific.net/MSF.706-709.3 ↗
21. Design of 10%Cr martensitic steels for improved creep resistance in power plant applications10.1179/0267083612Z.000000000218 ↗
22. Computational thermodynamics, Gaussian processes and genetic algorithms: combined tools to design new alloys10.1088/0965-0393/21/4/045013 ↗
23. Designing ultrahigh strength steels with good ductility by combining transformation induced plasticity and martensite aging10.1002/adem.200900061 ↗
24. A new ultrahigh-strength stainless steel strengthened by various coexisting nanoprecipitates10.1016/j.actamat.2010.03.005 ↗
25. From high-entropy alloys to high-entropy steels10.1002/srin.201500133 ↗
26. New nano-particle-strengthened ferritic/martensitic steels by conventional thermo-mechanical treatment10.1016/j.jnucmat.2007.03.001 ↗
27. Characterization of a boron alloyed 9Cr3W3CoVNbBN steel and further improvement of its high-temperature mechanical properties by thermomechanical treatments10.1016/j.jnucmat.2013.05.018 ↗
28. Effect of thermomechanical treatment on 9Cr ferritic–martensitic steels10.1016/j.jnucmat.2013.01.323 ↗
29. Stability of MX-type strengthening nanoprecipitates in ferritic steels under thermal aging, stress and ion irradiation10.1016/j.actamat.2014.03.015 ↗
30. Development of new generation reduced activation ferritic-martensitic steels for advanced fusion reactors10.1016/j.jnucmat.2016.05.037 ↗
31. Alvin · 2015
32. Additive manufacturing of metal-ceramic-composites by thermoplastic 3D-printing (3DTP)
33. Perspectives on additive manufacturing
34. Light Water Reactor Sustainability Program · 2012
35. Economies of scale in United States electric-power generation10.1086/260470 ↗
36. Economies of scale in nuclear-power generation10.2307/1058857 ↗
37. The economics of nuclear-power - further evidence on learning, economies of scale, and regulatory effects10.1016/0165-0572(88)90009-6 ↗
38. Economic features of integral, modular, small-to-medium size reactors10.1016/j.pnucene.2009.09.003 ↗
39. Financial case studies on small- and medium-size modular reactors10.13182/NT12-A13561 ↗
40. Bainitic chromium-tungsten steels with 3 pct chromium10.1007/s11661-997-0136-0 ↗
41. Creep-rupture behavior of 3Cr-3W-V bainitic steels10.1016/j.ijpvp.2006.09.004 ↗
42. Ferritic/martensitic steels for next generation reactors10.1016/j.jnucmat.2007.05.005 ↗
43. Design challenges and activation concerns for ARIES vacuum vessel10.13182/FST64-449 ↗
44. Stability of ferritic steel to higher doses: survey of reactor pressure vessel steel data and comparison with candidate materials for future nuclear systems10.1016/j.ijpvp.2014.06.001 ↗
45. Irradiation-assisted stress corrosion cracking10.1515/corrrev.2011.020 ↗
46. Irradiation-assisted stress corrosion cracking and impact on life extension10.5006/0803 ↗
47. Mechanism of dislocation channel-induced irradiation assisted stress corrosion crack initiation in austenitic stainless steel10.1016/j.cossms.2015.04.001 ↗
48. 2015 Frank Newman Speller award: stress corrosion cracking and nuclear power10.5006/1898 ↗
49. Pathania · 2012
50. Rickover · 1975
51. Adamson · 2010
52. Corrosion behavior of Zircaloy-4 cladding with varying tin content in high-temperature pressurized water reactors
53. The evolution of microstructure and deformation stability in Zr-Nb-(Sn, Fe) alloys under neutron irradiation
54. Corrosion of zirconium alloys used for nuclear fuel cladding10.1146/annurev-matsci-070214-020951 ↗
55. Development of zirconium-barrier fuel cladding
56. Hydrogen absorption mechanism of zirconium alloys based on characterization of oxide layer
57. G. Ledergerber, S. Valizadeh, J.L. Wright, M. Limbäck, L. Hallstadius, D. Gavillet, et al., Fuel performance beyond design–Exploring the limits, LWR Fuel Performance Meeting Top Fuel/WRFPM American Nuclear Society, 2010, pp. 513–24.
58. Effects of hydride precipitate localization and neutron fluence on the ductility of irradiated Zircaloy-4
59. Hydrogen pickup and redistribution in alpha-annealed Zircaloy-4
60. Hydride formation in zirconium alloys10.1007/s11837-012-0479-x ↗
61. Mechanisms of hydride reorientation in Zircaloy-4 studied in situ
62. 10CFR Part 50–Domestic licensing of production and utilization facilities, US Nuclear Regulatory Commission.
63. Taleb · 2007
64. Accident tolerant fuels for LWRs: a perspective10.1016/j.jnucmat.2013.12.005 ↗
65. Continuous SiC fiber, CVI SiC matrix composites for nuclear applications: properties and irradiation effects10.1016/j.jnucmat.2013.06.040 ↗
66. Oxidation of advanced zirconium cladding alloys in steam at temperatures in the range of 600–1200 °C10.1007/s11085-011-9249-3 ↗
67. Material selection for accident tolerant fuel cladding
68. K.R. Robb, Analysis of the FeCrAl accident tolerant fuel concept benefits during BWR station blackout accidents, in: Proc 16th Intern Topical Meeting on Nuclear Reactor Thermal Hydraulics, NURETH-162015, pp. 1183–1195.
69. Microencapsulated fuel technology for commercial light water and advanced reactor application10.1016/j.jnucmat.2012.05.021 ↗
70. Forward for special JNM issue on accident tolerant fuels for LWRs Preface10.1016/j.jnucmat.2014.03.030 ↗
71. Hydrothermal corrosion of SiC in LWR coolant environments in the absence of irradiation10.1016/j.jnucmat.2015.06.019 ↗
72. Uniform corrosion of FeCrAl alloys in LWR coolant environments10.1016/j.jnucmat.2016.06.047 ↗
73. A comparative study on the wear behaviors of cladding candidates for accident-tolerant fuel10.1016/j.jnucmat.2015.05.017 ↗
74. Thermo-mechanical analysis of LWR SiC/SiC composite cladding10.1016/j.jnucmat.2014.01.006 ↗
75. Stress analysis and probabilistic assessment of multi-layer SiC-based accident tolerant nuclear fuel cladding10.1016/j.jnucmat.2015.08.001 ↗
76. Generation IV nuclear reactors: current status and future prospects10.1016/j.enpol.2013.06.101 ↗
77. A Technology Roadmap for Generation IV Nuclear Energy Systems, Generation IV International Forum; 2002.
78. GIF R&D Outlook for Generation IV Nuclear Energy Systems, Generation IV International Forum; 2009.
79. Material performance in sodium
80. An approach for evaluating the integrity of fuel applied in innovative nuclear energy systems10.1016/j.jnucmat.2013.11.031 ↗
81. Materials requirements for liquid metal fast breeder reactors10.1007/BF02646695 ↗
82. Sodium fast reactor evaluation: core materials10.1016/j.jnucmat.2009.03.021 ↗
83. French and United Kingdom experience of high-burnup mixed-oxide fuel in sodium-cooled fast breeder reactors10.13182/NT08-A3911 ↗
84. A once-through fuel cycle for fast reactors10.1115/1.4000898 ↗
85. 9Cr-1Mo-V-Nb steel
86. Creep behaviour of 14Cr-15Ni-Ti stainless steel at 923 K10.1016/j.msea.2010.04.043 ↗
87. Development of next generation tempered and ODS reduced activation ferritic/martensitic steels for fusion energy applications
88. ODS ferritic/martensitic alloys for sodium fast reactor fuel pin cladding10.1016/j.jnucmat.2011.10.037 ↗
89. Oxide dispersion strengthened steels
90. Recent progress in developing and qualifying nanostructured ferritic alloys for advanced fission and fusion applications10.1007/s11837-014-1207-5 ↗
91. Damage assessment in structural metallic materials for advanced nuclear plants10.1007/s10853-010-4236-7 ↗
92. A perspective on creep and fatigue issues in sodium cooled fast reactors10.1007/s12666-010-0011-3 ↗
93. Evaluation of creep damage in heat affected zone of thick welded joint for Mod.9Cr-1Mo steel10.1016/j.ijpvp.2009.04.008 ↗
94. Ferritic steels for sodium-cooled fast reactors: design principles and challenges10.1007/s11837-010-0143-2 ↗
95. Numerical analysis of thermal striping induced high cycle thermal fatigue in a mixing tee10.1016/j.nucengdes.2008.06.014 ↗
96. Thermal hydraulic issues and challenges for current and new generation FBRs10.1016/j.nucengdes.2015.09.012 ↗
97. In-reactor creep rupture of 20% cold-worked AISI 316 stainless steel10.1007/BF00542729 ↗
98. In-beam mechanical testing of CuCrZr10.1016/j.jnucmat.2004.04.011 ↗
99. Overview of recent European materials R&D activities related to ITER10.1016/j.jnucmat.2004.04.008 ↗
100. Critical questions in materials science and engineering for successful development of fusion power10.1016/j.jnucmat.2007.02.007 ↗