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Title: Irradiation performance of fast reactor MOX fuel pins with ferritic/martensitic cladding irradiated to high burnups

Abstract

The ACO-3 irradiation test, which attained extremely high burnups of about 232 GWd/t and resisted a high neutron fluence (E > 0.1 MeV) of about 39 × 1026 n/m2 as one of the lead tests of the Core Demonstration Experiment in the Fast Flux Test Facility, demonstrated that the fuel pin cladding made of ferritic/martensitic HT-9 alloy had superior void swelling resistance. The measured diameter profiles of the irradiated ACO-3 fuel pins showed axially extensive incremental strain in the MOX fuel column region and localized incremental strain near the interfaces between the MOX fuel and upper blanket columns. These incremental strains were as low as 1.5% despite the extremely high level of the fast neutron fluence. Evaluation of the pin diametral strain indicated that the incremental strain in the MOX fuel column region was substantially due to cladding void swelling and irradiation creep caused by internal fission gas pressure, while the localized strain near the MOX fuel/upper blanket interface was likely the result of the pellet/cladding mechanical interaction (PCMI) caused by cesium/fuel reactions. The evaluation also suggested that the PCMI was effectively mitigated by a large gap size between the cladding and blanket column.

Authors:
; ; ; ; ;
Publication Date:
Research Org.:
Idaho National Laboratory (INL)
Sponsoring Org.:
USDOE
OSTI Identifier:
1022744
Report Number(s):
INL/JOU-11-22959
Journal ID: ISSN 0022-3115; TRN: US1104354
DOE Contract Number:  
DE-AC07-05ID14517
Resource Type:
Journal Article
Journal Name:
Journal of Nuclear Materials
Additional Journal Information:
Journal Volume: 412; Journal Issue: 3; Journal ID: ISSN 0022-3115
Country of Publication:
United States
Language:
English
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; ALLOYS; CREEP; EVALUATION; FAST NEUTRONS; FAST REACTORS; FFTF REACTOR; FISSION; FUEL PINS; IRRADIATION; NEUTRON FLUENCE; PERFORMANCE; STRAINS; SWELLING

Citation Formats

Uwaba, Tomoyuki, Ito, Masahiro, Katsuyama, Kozo, Makenas, Bruce J, Wootan, David W, and Carmack, Jon. Irradiation performance of fast reactor MOX fuel pins with ferritic/martensitic cladding irradiated to high burnups. United States: N. p., 2011. Web. doi:10.1016/j.jnucmat.2011.02.052.
Uwaba, Tomoyuki, Ito, Masahiro, Katsuyama, Kozo, Makenas, Bruce J, Wootan, David W, & Carmack, Jon. Irradiation performance of fast reactor MOX fuel pins with ferritic/martensitic cladding irradiated to high burnups. United States. doi:10.1016/j.jnucmat.2011.02.052.
Uwaba, Tomoyuki, Ito, Masahiro, Katsuyama, Kozo, Makenas, Bruce J, Wootan, David W, and Carmack, Jon. Sun . "Irradiation performance of fast reactor MOX fuel pins with ferritic/martensitic cladding irradiated to high burnups". United States. doi:10.1016/j.jnucmat.2011.02.052.
@article{osti_1022744,
title = {Irradiation performance of fast reactor MOX fuel pins with ferritic/martensitic cladding irradiated to high burnups},
author = {Uwaba, Tomoyuki and Ito, Masahiro and Katsuyama, Kozo and Makenas, Bruce J and Wootan, David W and Carmack, Jon},
abstractNote = {The ACO-3 irradiation test, which attained extremely high burnups of about 232 GWd/t and resisted a high neutron fluence (E > 0.1 MeV) of about 39 × 1026 n/m2 as one of the lead tests of the Core Demonstration Experiment in the Fast Flux Test Facility, demonstrated that the fuel pin cladding made of ferritic/martensitic HT-9 alloy had superior void swelling resistance. The measured diameter profiles of the irradiated ACO-3 fuel pins showed axially extensive incremental strain in the MOX fuel column region and localized incremental strain near the interfaces between the MOX fuel and upper blanket columns. These incremental strains were as low as 1.5% despite the extremely high level of the fast neutron fluence. Evaluation of the pin diametral strain indicated that the incremental strain in the MOX fuel column region was substantially due to cladding void swelling and irradiation creep caused by internal fission gas pressure, while the localized strain near the MOX fuel/upper blanket interface was likely the result of the pellet/cladding mechanical interaction (PCMI) caused by cesium/fuel reactions. The evaluation also suggested that the PCMI was effectively mitigated by a large gap size between the cladding and blanket column.},
doi = {10.1016/j.jnucmat.2011.02.052},
journal = {Journal of Nuclear Materials},
issn = {0022-3115},
number = 3,
volume = 412,
place = {United States},
year = {2011},
month = {5}
}