Abstract

Understanding fuel system behavior during postulated loss-of-coolant accidents is pertinent for continued safe and efficient operation of light water reactors, particularly as higher burnups are being pursued and safety margins re-evaluated. Conventional mechanical models for the incumbent Zr alloys typically rely on the assumption that steady-state creep is the dominant fuel cladding response during transient accident conditions. To investigate this assumption, simulated accident burst testing was performed on Zircaloy-4 claddings with balloon behavior measured in-situ. Here, two distinct loading conditions were utilized during burst testing: (1) constant-gas-inventory where pressure was allowed to increase with temperature and (2) constant pressure. In-situ strains and strain rates were measured via 2-dimensional digital image correlation techniques and synchronized with temperature to determine deformation dependencies. The temperature dependence of strain rate was characterized by a two segment Arrhenius relationship, with a distinct transition between the high and low temperature/strain regimes. The average activation energy of the lower temperature/strain regime was 328 ± 25 kJ/mol, in agreement with the ~320 kJ/mol used for conventional LOCA models. However, the higher temperature/strain segment, which encompassed most of ballooning, showed increased activation energies as well as a dependence on whether the burst region was in view. For tests thatmore » burst away from the camera view, the average high temperature/strain segment activation energy was 635 ± 150 kJ/mol. For samples where the rupture opening formed in view, the average activation energy was 1015 ± 179 kJ/mol. This observed shift in temperature dependence indicates a transition in deformation mechanism at the end of life, possibly to time independent failure mechanisms, which has not yet been visualized in the literature for Zr alloys. Parameters at the transition points were analyzed to determine thresholds for this change in behavior, which occurred at an average hoop strain of 6.9 ± 2.1 %.« less

Authors:

Bell, Samuel B.  ^[1];

• Search DOE PAGES for author "Bell, Samuel B."
• Search DOE PAGES for ORCID "0000-0002-5190-5657"

Kane, Kenneth A.  ^[2];

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• Search DOE PAGES for ORCID "0000-0003-3647-3419"

Ridley, Mackenzie J.  ^[1];

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• Search DOE PAGES for ORCID "0000-0002-4043-2209"

Garrison, Ben E. ^[1]; Johnston, Brandon S. ^[1]; Capps, Nathan A.  ^[1]

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• Search DOE PAGES for ORCID "0000-0003-0046-4289"

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+ Show Author Affiliations

1. Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)
2. Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Johns Hopkins Univ., Baltimore, MD (United States)

Publication Date:

Thu Jan 11 00:00:00 EST 2024

Research Org.:

Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)

Sponsoring Org.:

USDOE Office of Nuclear Energy (NE), Nuclear Fuel Cycle and Supply Chain. Advanced Fuel Campaign

OSTI Identifier:

2333851

Alternate Identifier(s):

OSTI ID: 2281443

Grant/Contract Number:  

AC05-00OR22725

Resource Type:

Accepted Manuscript

Journal Name:

Journal of Nuclear Materials

Additional Journal Information:

Journal Volume: 591; Journal Issue: 1; Journal ID: ISSN 0022-3115

Publisher:

Elsevier

Country of Publication:

United States

Language:

English

Subject:

36 MATERIALS SCIENCE; LOCA; fuel cladding; in-situ strain; DIC; design basis; zircaloy; fuel safety