Oxidation of aluminum alloy cladding for research and test reactor fuel

Kim, Yeon Soo; Hofman, G.L.; Robinson, Adam; Snelgrove, J.L.; Hanan, N. A.

doi:10.1016/j.jnucmat.2008.06.032

Cited by 62 publications

(49 citation statements)

References 6 publications

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“…The matrix was either pure Al or Al with a Si addition (Al-Si). The first goal of the IRIS4 experiment was to evaluate the interest of this ''oxide'' solution, and the second was to check whether this potentially beneficial effect could be cumulated with the well-established influence of Si addition to the Al matrix [14][15][16][17][18][19].…”

Section: Introductionmentioning

confidence: 99%

Annealing tests of in-pile irradiated oxide coated U–Mo/Al–Si dispersed nuclear fuel

Zweifel¹,

Valot²,

Pontillon³

et al. 2014

Journal of Nuclear Materials

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is an open access repository that collects the work of Arts et Métiers ParisTech researchers and makes it freely available over the web where possible. a b s t r a c t U-Mo/Al based nuclear fuels have been worldwide considered as a promising high density fuel for the conversion of high flux research reactors from highly enriched uranium to lower enrichment. In this paper, we present the annealing test up to 1800 °C of in-pile irradiated U-Mo/Al-Si fuel plate samples. More than 70% of the fission gases (FGs) are released during two major FG release peaks around 500 °C and 670 °C. Additional characterisations of the samples by XRD, EPMA and SEM suggest that up to 500 °C FGs are released from IDL/matrix interfaces. The second peak at 670 °C representing the main release of FGs originates from the interaction between U-Mo and matrix in the vicinity of the cladding.

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Section: Introductionmentioning

confidence: 99%

Annealing tests of in-pile irradiated oxide coated U–Mo/Al–Si dispersed nuclear fuel

Zweifel¹,

Valot²,

Pontillon³

et al. 2014

Journal of Nuclear Materials

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“…This oxidation layer can pose a considerable performance issue for the dispersion fuel owing to its low thermal conductivity (~ 10-fold lower than that of Al). As determined from the previous RERTR test data, the oxide growth is a function of irradiation time, coolant temperature, fission rate, and pH value [27]. The growth of oxide layers was not substantial in the ATR as compared to other reactors with similar power, which is probably a result of the relatively low pH of the primary coolant water.…”

Section: Resultsmentioning

confidence: 73%

A Description of the Mechanistic Dart - Thermal Dispersion Fuel Performance Code and Application to Irradiation Behavior Analysis of U-Mo/Al

Ye¹,

Rest²,

Kim³

2013

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“…2 In general, the parameter A is a function of alloy composition, cover gas, water vapor composition, and relative humidity (RH). The activation energy Q is a function of alloy composition and n is a function of alloy composition, and water vapor composition.…”

Section: Sealed Dry Storagementioning

confidence: 99%

“…Aluminium alloys show very different oxidation behaviour from fuels produced with stainless steel and Zircaloy claddings. For example, during reactor operation, aluminium claddings can develop relatively thick oxide layers [2] that are subject to spallation [3]. Also, aluminum fuel is particularly susceptible to corrosion degradation in water, particularly if the quality of the water is poor [4].…”

Section: Introductionmentioning

confidence: 99%

OVERVIEW OF CRITERIA FOR INTERIM WET & DRY STORAGE OF RESEARCH REACTOR SPENT NUCLEAR FUEL

Sindelar¹,

Vinson²,

Iyer³

et al. 2010

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Following discharge from research reactors, spent nuclear fuel may be stored "wet" in water pools or basins, or it may be stored "dry" in various configurations including non-sealed or sealed containers until retrieved for ultimate disposition. Interim safe storage practices are based on avoiding degradation to the fuel that would impact functions related to safety. Recommended practices including environmental controls with technical bases, are outlined for wet storage and dry storage of aluminum-clad, aluminum-based research reactor fuel.For wet storage, water quality must be maintained to minimize corrosion degradation of aluminum fuel. For dry storage, vented canister storage of aluminum fuel readily provides a safe storage configuration. For sealed dry storage, drying must be performed so as to minimize water that would cause additional corrosion and hydrogen generation. Consideration must also be given to the potential for radiolytically-generated hydrogen from the bound water in the attendant oxyhydroxides on aluminum fuel from reactor operation for dry storage systems.

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Oxidation of aluminum alloy cladding for research and test reactor fuel

Cited by 62 publications

References 6 publications

Annealing tests of in-pile irradiated oxide coated U–Mo/Al–Si dispersed nuclear fuel

Annealing tests of in-pile irradiated oxide coated U–Mo/Al–Si dispersed nuclear fuel

A Description of the Mechanistic Dart - Thermal Dispersion Fuel Performance Code and Application to Irradiation Behavior Analysis of U-Mo/Al

OVERVIEW OF CRITERIA FOR INTERIM WET & DRY STORAGE OF RESEARCH REACTOR SPENT NUCLEAR FUEL

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