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Title: Experimental studies of U-Pu-Zr fast reactor fuel pins in the Experimental Breeder Reactor 2

Abstract

Argonne National Laboratory's Integral Fast Reactor (IFR) concept has been under demonstration in the Experimental Breeder Reactor II (EBR-II) since February 1985. Irradiation tests of U-Zr and U-Pu-Zr fuel pins to {gt}15 at. pct burnup have demonstrated their viability as driver fuel prototypes in innovative design liquid metal reactors. A number of technically challenging irradiation effects have been observed and are now under study. Microstructural changes in the fuel are dominated early in exposure by grain boundary cavitation and fission gas bubble growth, producing large amounts of swelling. Irradiation creep and swelling of the austenitic (D9) and martensitic (HT-9) candidate cladding alloys have been measured and correlate well with property modeling efforts. Chemical interaction between the fuel and cladding alloys has been characterized to assess the magnitude of cladding wastage during steady-state irradiation. Significant interdiffusion of the uranium and zirconium occurs producing metallurgically distinct zones in the fuel.

Authors:
; ;  [1];  [2]
  1. Argonne National Lab., Idaho Falls, ID (US)
  2. Argonne National Lab., Argonne, IL (US)
Publication Date:
OSTI Identifier:
5906011
Resource Type:
Journal Article
Journal Name:
Metallurgical Transactions, A (Physical Metallurgy and Materials Science); (USA)
Additional Journal Information:
Journal Volume: 21:7; Journal ID: ISSN 0360-2133
Country of Publication:
United States
Language:
English
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; 36 MATERIALS SCIENCE; 46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND TECHNOLOGY; EBR-2 REACTOR; FUEL PINS; MATERIALS TESTING; PLUTONIUM ALLOYS; PHYSICAL RADIATION EFFECTS; URANIUM ALLOYS; ZIRCONIUM ALLOYS; ANL; AUSTENITIC STEELS; CAVITATION; FUEL-CLADDING INTERACTIONS; GRAIN BOUNDARIES; IFR REACTOR; MICROSTRUCTURE; PHASE TRANSFORMATIONS; REACTOR COMPONENTS; ACTINIDE ALLOYS; ALLOYS; BREEDER REACTORS; CRYSTAL STRUCTURE; EPITHERMAL REACTORS; EXPERIMENTAL REACTORS; FAST REACTORS; FBR TYPE REACTORS; FUEL ELEMENTS; IRON ALLOYS; IRON BASE ALLOYS; LIQUID METAL COOLED REACTORS; LMFBR TYPE REACTORS; NATIONAL ORGANIZATIONS; POWER REACTORS; RADIATION EFFECTS; REACTORS; RESEARCH AND TEST REACTORS; SODIUM COOLED REACTORS; STEELS; TESTING; US AEC; US DOE; US ERDA; US ORGANIZATIONS; ZERO POWER REACTORS; 220600* - Nuclear Reactor Technology- Research, Test & Experimental Reactors; 360106 - Metals & Alloys- Radiation Effects; 360102 - Metals & Alloys- Structure & Phase Studies; 440200 - Radiation Effects on Instrument Components, Instruments, or Electronic Systems

Citation Formats

Pahl, R G, Porter, D L, Lahm, C E, and Hofman, G L. Experimental studies of U-Pu-Zr fast reactor fuel pins in the Experimental Breeder Reactor 2. United States: N. p., 1990. Web. doi:10.1007/BF02647233.
Pahl, R G, Porter, D L, Lahm, C E, & Hofman, G L. Experimental studies of U-Pu-Zr fast reactor fuel pins in the Experimental Breeder Reactor 2. United States. https://doi.org/10.1007/BF02647233
Pahl, R G, Porter, D L, Lahm, C E, and Hofman, G L. 1990. "Experimental studies of U-Pu-Zr fast reactor fuel pins in the Experimental Breeder Reactor 2". United States. https://doi.org/10.1007/BF02647233.
@article{osti_5906011,
title = {Experimental studies of U-Pu-Zr fast reactor fuel pins in the Experimental Breeder Reactor 2},
author = {Pahl, R G and Porter, D L and Lahm, C E and Hofman, G L},
abstractNote = {Argonne National Laboratory's Integral Fast Reactor (IFR) concept has been under demonstration in the Experimental Breeder Reactor II (EBR-II) since February 1985. Irradiation tests of U-Zr and U-Pu-Zr fuel pins to {gt}15 at. pct burnup have demonstrated their viability as driver fuel prototypes in innovative design liquid metal reactors. A number of technically challenging irradiation effects have been observed and are now under study. Microstructural changes in the fuel are dominated early in exposure by grain boundary cavitation and fission gas bubble growth, producing large amounts of swelling. Irradiation creep and swelling of the austenitic (D9) and martensitic (HT-9) candidate cladding alloys have been measured and correlate well with property modeling efforts. Chemical interaction between the fuel and cladding alloys has been characterized to assess the magnitude of cladding wastage during steady-state irradiation. Significant interdiffusion of the uranium and zirconium occurs producing metallurgically distinct zones in the fuel.},
doi = {10.1007/BF02647233},
url = {https://www.osti.gov/biblio/5906011}, journal = {Metallurgical Transactions, A (Physical Metallurgy and Materials Science); (USA)},
issn = {0360-2133},
number = ,
volume = 21:7,
place = {United States},
year = {Sun Jul 01 00:00:00 EDT 1990},
month = {Sun Jul 01 00:00:00 EDT 1990}
}