High Temperature Fuel Cladding Chemical Interactions Between TRIGA Fuels and 304 Stainless Steel

Technical Report · Mon Feb 01 00:00:00 EST 2016

DOI:https://doi.org/10.2172/1259949· OSTI ID:1259949

Perez, Emmanuel ^[1]; Keiser, Jr., Dennis D. ^[1]; Forsmann, Bryan ^[2]; Janney, Dawn E. ^[1]; Henley, Jody ^[1]; Woolstenhulme, Eric C. ^[1]

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1. Idaho National Lab. (INL), Idaho Falls, ID (United States)
2. Boise State Univ., ID (United States)

High-temperature fuel-cladding chemical interactions (FCCI) between TRIGA (Training, Research, Isotopes, General Atomics) fuel elements and the 304 stainless steel (304SS) are of interest to develop an understanding of the fuel behavior during transient reactor scenarios. TRIGA fuels are composed of uranium (U) particles dispersed in a zirconium-hydride (Zr-H) matrix. In reactor, the fuel is encased in 304-stainless-steel (304SS) or Incoloy 800 clad tubes. At high temperatures, the fuel can readily interact with the cladding, resulting in FCCI. A number of FCCI can take place in this system. Interactions can be expected between the cladding and the Zr-H matrix, and/or between the cladding and the U-particles. Other interactions may be expected between the Zr-H matrix and the U-particles. Furthermore, the fuel contains erbium-oxide (Er-O) additions. Interactions can also be expected between the Er-O, the cladding, the Zr-H and the U-particles. The overall result is that very complex interactions may take place as a result of fuel and cladding exposures to high temperatures. This report discusses the characterization of the baseline fuel microstructure in the as-received state (prior to exposure to high temperature), characterization of the fuel after annealing at 950C for 24 hours and the results from diffusion couple experiments carries out at 1000C for 5 and 24 hours. Characterization was carried out via scanning electron microscopy (SEM) and transmission electron microscopy (TEM) with sample preparation via focused ion beam in situ-liftout-technique.

• Research Organization: Idaho National Lab. (INL), Idaho Falls, ID (United States)
• Sponsoring Organization: USDOE National Nuclear Security Administration (NNSA)
• DOE Contract Number: AC07-05ID14517
• OSTI ID: 1259949
• Report Number(s): INL/EXT-15-36708; TRN: US1601536
• Country of Publication: United States
• Language: English

Related Subjects

36 MATERIALS SCIENCE
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS
22 GENERAL STUDIES OF NUCLEAR REACTORS
FUEL-CLADDING INTERACTIONS
STAINLESS STEEL-304
TRIGA TYPE REACTORS
TEMPERATURE RANGE 1000-4000 K
NUCLEAR FUELS
INCOLOY 800
CHEMICAL REACTIONS
URANIUM
ERBIUM OXIDES
ZIRCONIUM HYDRIDES
MATRIX MATERIALS
FUEL ELEMENTS
ANNEALING
FUEL CANS
DIFFUSION
MICROSTRUCTURE
TRANSIENTS
304 Stainless Steel
304SS
FCCI
Fuel Cladding Chemical Interactions
Interactions
TRIGA