Report on fundamental modeling of irradiation-induced swelling and creep in FeCrAl alloys [electronic resource]
In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in h...
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Online Access (via OSTI) |
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| Format: | Government Document Electronic eBook |
| Language: | English |
| Published: |
Washington, D.C. : Oak Ridge, Tenn. :
United States. Office of the Assistant Secretary for Nuclear Energy ; distributed by the Office of Scientific and Technical Information, U.S. Department of Energy,
2016.
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| Subjects: |
MARC
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| 245 | 0 | 0 | |a Report on fundamental modeling of irradiation-induced swelling and creep in FeCrAl alloys |h [electronic resource] |
| 260 | |a Washington, D.C. : |b United States. Office of the Assistant Secretary for Nuclear Energy ; |a Oak Ridge, Tenn. : |b distributed by the Office of Scientific and Technical Information, U.S. Department of Energy, |c 2016. | ||
| 300 | |a 17 p. : |b digital, PDF file. | ||
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| 500 | |a "NEAF278" | ||
| 500 | |a Aaron A. Kohnert; Dwaipayan Dasgupta; Brian Wirth; Kory D. Linton. | ||
| 520 | 3 | |a In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, the material response must be demonstrated to provide suitable radiation stability, in order to ensure that there will not be significant dimensional changes (e.g., swelling), as well as quantifying the radiation hardening and radiation creep behavior. In this report, we describe the use of cluster dynamics modeling to evaluate the defect physics and damage accumulation behavior of FeCrAl alloys subjected to neutron irradiation, with a particular focus on irradiation-induced swelling and defect fluxes to dislocations that are required to model irradiation creep behavior. | |
| 536 | |b AC05-00OR22725. | ||
| 650 | 7 | |a Creep. |2 local. | |
| 650 | 7 | |a Swelling. |2 local. | |
| 650 | 7 | |a Iron Base Alloys. |2 local. | |
| 650 | 7 | |a Chromium Alloys. |2 local. | |
| 650 | 7 | |a Aluminium Alloys. |2 local. | |
| 650 | 7 | |a Ternary Alloy Systems. |2 local. | |
| 650 | 7 | |a Temperature Range 0400-1000 K. |2 local. | |
| 650 | 7 | |a Fuel Cans. |2 local. | |
| 650 | 7 | |a Water Moderated Reactors. |2 local. | |
| 650 | 7 | |a Physical Radiation Effects. |2 local. | |
| 650 | 7 | |a Neutrons. |2 local. | |
| 650 | 7 | |a Dislocations. |2 local. | |
| 650 | 7 | |a Radiation Hardening. |2 local. | |
| 650 | 7 | |a Computerized Simulation. |2 local. | |
| 650 | 7 | |a Water Cooled Reactors. |2 local. | |
| 650 | 7 | |a Stability. |2 local. | |
| 650 | 7 | |a Accident-Tolerant Nuclear Fuels. |2 local. | |
| 650 | 7 | |a Materials Science. |2 edbsc. | |
| 710 | 2 | |a Oak Ridge National Laboratory. |4 res. | |
| 710 | 1 | |a United States. |b Office of the Assistant Secretary for Nuclear Energy. |4 spn. | |
| 710 | 1 | |a United States. |b Department of Energy. |b Office of Scientific and Technical Information. |4 dst. | |
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