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Title: Report on fundamental modeling of irradiation-induced swelling and creep in FeCrAl alloys

Abstract

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.

Authors:
 [1];  [2];  [1];  [1]
  1. Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
  2. Univ. of Tennessee, Knoxville, TN (United States)
Publication Date:
Research Org.:
Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
Sponsoring Org.:
USDOE Office of Nuclear Energy (NE)
OSTI Identifier:
1337855
Report Number(s):
ORNL/TM-2016/569
AF5810000; NEAF278; TRN: US1701425
DOE Contract Number:
AC05-00OR22725
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
36 MATERIALS SCIENCE; CREEP; SWELLING; IRON BASE ALLOYS; CHROMIUM ALLOYS; ALUMINIUM ALLOYS; TERNARY ALLOY SYSTEMS; TEMPERATURE RANGE 0400-1000 K; FUEL CANS; WATER MODERATED REACTORS; PHYSICAL RADIATION EFFECTS; NEUTRONS; DISLOCATIONS; RADIATION HARDENING; COMPUTERIZED SIMULATION; WATER COOLED REACTORS; STABILITY; ACCIDENT-TOLERANT NUCLEAR FUELS

Citation Formats

Kohnert, Aaron A., Dasgupta, Dwaipayan, Wirth, Brian, and Linton, Kory D. Report on fundamental modeling of irradiation-induced swelling and creep in FeCrAl alloys. United States: N. p., 2016. Web. doi:10.2172/1337855.
Kohnert, Aaron A., Dasgupta, Dwaipayan, Wirth, Brian, & Linton, Kory D. Report on fundamental modeling of irradiation-induced swelling and creep in FeCrAl alloys. United States. doi:10.2172/1337855.
Kohnert, Aaron A., Dasgupta, Dwaipayan, Wirth, Brian, and Linton, Kory D. 2016. "Report on fundamental modeling of irradiation-induced swelling and creep in FeCrAl alloys". United States. doi:10.2172/1337855. https://www.osti.gov/servlets/purl/1337855.
@article{osti_1337855,
title = {Report on fundamental modeling of irradiation-induced swelling and creep in FeCrAl alloys},
author = {Kohnert, Aaron A. and Dasgupta, Dwaipayan and Wirth, Brian and Linton, Kory D.},
abstractNote = {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.},
doi = {10.2172/1337855},
journal = {},
number = ,
volume = ,
place = {United States},
year = 2016,
month = 9
}

Technical Report:

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  • In order to use data from surrogate neutron spectra for fusion applications, it is necessary to analyze the impact of environmental differences on property development. This is of particular importance in the study of irradiation creep and its interactions with void swelling, especially with respect to the difficulty of separation of creep strains from various non-creep strains. As part of an on-going creep data rescue and analysis effort, the current study focuses on comparative irradiations conducted on identical gas-pressurized tubes produced and constructed in the United States from austenitic steels (20% CW 316 and 20% CW D9), but irradiated inmore » either the Prototype Fast Reactor (PFR) in the United Kingdom or the Fast Flux Test Facility in the United States. In PFR, Demountable Subassemblies (DMSA) serving as heat pipes were used without active temperature control. In FFTF the specimens were irradiated with active ({+-}{degrees}5C) temperature control. Whereas the FFTF irradiations involved a series of successive side-by-side irradiation, measurement and reinsertion of the same series of tubes, the PFR experiment utilized simultaneous irradiation at two axial positions in the heat pipe to achieve different fluences at different flux levels. The smaller size of the DMSA also necessitated a separation of the tubes at a given flux level into two groups (low-stress and high-stress) at slightly different axial positions, where the flux between the two groups varied {le}10%. Of particular interest in this study was the potential impact of the two types of separation on the derivation of creep coefficients.« less
  • A general theory of the effects of point defect trapping on radiation-induced swelling and creep deformation rates is developed. The effects on the fraction of defects recombining, and on void nucleation, void growth and creep due to the separate processes of dislocation climb-glide and dislocation climb (the so-called SIPA mechanism) are studied. Trapping of vacancies or interstitials increases total recombination and decreases the rates of deformation processes. For fixed trapping parameters, the reduction is largest for void nucleation, less for void growth and creep due to dislocation climb-glide, and least for creep due to dislocation climb. With this formation, themore » effects of trapping at multiple vacancy and interstitial traps and of spatial and temporal variation in trap concentrations may be determined. Alternative pictures for viewing point defect trapping in terms of effective recombination and diffusion coefficients are derived. It is shown that previous derivations of these coefficients are incorrect. A rigorous explanation is given of the well-known numerical result that interstitial trapping is significant only if the binding energy exceeds the difference between the vacancy and interstitial migration energies, while vacancy trapping is significant even at small binding energies. Corrections which become necessary at solute concentrations above about 0.1% are described. Numerical results for a wide range of material and irradiation parameters are presented.« less