Cladding burst behavior of Fe-based alloys under LOCA

Terrani, Kurt A.; Dryepondt, Sebastien N.; Pint, Bruce A.; Massey, Caleb P.

doi:10.1016/j.jnucmat.2015.12.018

Title: Cladding burst behavior of Fe-based alloys under LOCA

Full Record
Other Related Research

Abstract

Burst behavior of austenitic and ferritic Fe-based alloy tubes has been examined under a simulated large break loss of coolant accident. Specifically, type 304 stainless steel (304SS) and oxidation resistant FeCrAl tubes were studied alongside Zircaloy-2 and Zircaloy-4 that are considered reference fuel cladding materials. Following the burst test, characterization of the cladding materials was carried out to gain insights regarding the integral burst behavior. Given the widespread availability of a comprehensive set of thermo-mechanical data at elevated temperatures for 304SS, a modeling framework was implemented to simulate the various processes that affect burst behavior in this Fe-based alloy. The most important conclusion is that cladding ballooning due to creep is negligible for Fe-based alloys. Thus, unlike Zr-based alloys, cladding cross-sectional area remains largely unchanged up to the point of burst. Furthermore, for a given rod internal pressure, the temperature onset of burst in Fe-based alloys appears to be simply a function of the alloy's ultimate tensile strength, particularly at high rod internal pressures.

Authors:

Terrani, Kurt A. ^[1]; Dryepondt, Sebastien N. ^[1]; Pint, Bruce A. ^[1]; Massey, Caleb P. ^[2]

Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States)
Oak Ridge National Laboratory (ORNL), Oak Ridge, TN (United States); Virginia Commonwealth Univ., Richmond, VA (United States)

Publication Date:: 2015-12-17

Research Org.:: Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

Sponsoring Org.:: USDOE Office of Nuclear Energy (NE)

OSTI Identifier:: 1244189

Alternate Identifier(s):: OSTI ID: 1396643

Grant/Contract Number:: AC05-00OR22725

Resource Type:: Accepted Manuscript

Journal Name:: Journal of Nuclear Materials

Additional Journal Information:: Journal Volume: 470; Journal Issue: C; Journal ID: ISSN 0022-3115

Publisher:: Elsevier

Country of Publication:: United States

Language:: English

Subject:: 42 ENGINEERING

Citation Formats


                    Terrani, Kurt A., Dryepondt, Sebastien N., Pint, Bruce A., and Massey, Caleb P.. Cladding burst behavior of Fe-based alloys under LOCA.  United States: N. p., 2015. 
Web.  doi:10.1016/j.jnucmat.2015.12.018.

Copy to clipboard


                    Terrani, Kurt A., Dryepondt, Sebastien N., Pint, Bruce A., & Massey, Caleb P.. Cladding burst behavior of Fe-based alloys under LOCA.  United States.  https://doi.org/10.1016/j.jnucmat.2015.12.018

Copy to clipboard


                    Terrani, Kurt A., Dryepondt, Sebastien N., Pint, Bruce A., and Massey, Caleb P.. Thu .  
"Cladding burst behavior of Fe-based alloys under LOCA".  United States.  https://doi.org/10.1016/j.jnucmat.2015.12.018.  https://www.osti.gov/servlets/purl/1244189.

Copy to clipboard


                    
@article{osti_1244189,

  title        = {Cladding burst behavior of Fe-based alloys under LOCA},

  author       = {Terrani, Kurt A. and Dryepondt, Sebastien N. and Pint, Bruce A. and Massey, Caleb P.},

  abstractNote = {Burst behavior of austenitic and ferritic Fe-based alloy tubes has been examined under a simulated large break loss of coolant accident. Specifically, type 304 stainless steel (304SS) and oxidation resistant FeCrAl tubes were studied alongside Zircaloy-2 and Zircaloy-4 that are considered reference fuel cladding materials. Following the burst test, characterization of the cladding materials was carried out to gain insights regarding the integral burst behavior. Given the widespread availability of a comprehensive set of thermo-mechanical data at elevated temperatures for 304SS, a modeling framework was implemented to simulate the various processes that affect burst behavior in this Fe-based alloy. The most important conclusion is that cladding ballooning due to creep is negligible for Fe-based alloys. Thus, unlike Zr-based alloys, cladding cross-sectional area remains largely unchanged up to the point of burst. Furthermore, for a given rod internal pressure, the temperature onset of burst in Fe-based alloys appears to be simply a function of the alloy's ultimate tensile strength, particularly at high rod internal pressures.},

  doi          = {10.1016/j.jnucmat.2015.12.018},

  journal      = {Journal of Nuclear Materials},

  number       = C,

  volume       = 470,

  place        = {United States},

  year         = {2015},

  month        = {12}

}

Copy to clipboard

Journal Article:

Free Publicly Available Full Text

Accepted Manuscript (Publisher)

Accepted Manuscript (DOE)

Publisher's Version of Record

https://doi.org/10.1016/j.jnucmat.2015.12.018

Other availability

Search WorldCat to find libraries that may hold this journal

Citation Metrics:

Cited by: 48 works

Citation information provided by
Web of Science

Save / Share:

Export Metadata

Save to My Library

Similar Records in DOE PAGES and OSTI.GOV collections:

Strength and rupture geometry of un-irradiated C26M FeCrAl under LOCA burst testing conditions

Journal Article Bell, Sam B. ; Kane, Kenneth A. ; Massey, Caleb P. ; ... - Journal of Nuclear Materials

Ferritic iron-chromium-aluminum (FeCrAl) alloys are an accident tolerant fuel candidate to replace the incumbent Zr-based claddings. Nuclear grade FeCrAl alloys are marked by superior high temperature mechanical behavior and exceptional steam oxidation resistance, both of which increase safety margins during accident scenarios. In the present study, the loss of coolant accident (LOCA) burst behaviors of three un-irradiated commercially fabricated cladding materials in a simulated LOCA environment were compared: (1) T35Y2, a 1^st generation nuclear grade FeCrAl, (2) C26M, a 2^nd generation nuclear grade FeCrAl, and (3) Zircaloy-2. Both FeCrAl alloys showed improved mechanical strength and steam oxidation resistance compared tomore »« less
https://doi.org/10.1016/j.jnucmat.2021.153242

Full Text Available
FUEL PERFORMANCE MODELING OF FECRAL CLADDING FAILURE DURING LOSS OF COOLANT CONDITIONS

Conference Gamble, K. A. ; Hales, J. D. ; Pastore, G. ; ...

Accident tolerant materials are defined as those that provide significantly increased response time in the event of an accident while providing similar or improved performance as the conventional UO2/Zircaloy-4 fuel rods during normal operation. One avenue for materials to demonstrate enhanced accident tolerance is to have improved reaction (e.g., oxidation) kinetics with steam, resulting in slower hydrogen (or other combustible gases) generation rate, while maintaining acceptable thermo-mechanical properties, fuel-clad interactions, and fission product behavior. Due to their improved reaction kinetics with steam compared to Zircaloy-4, iron-chromium-aluminum (FeCrAl) alloys are being considered as an accident tolerant cladding material. In order tomore »« less
Full Text Available
Fuel performance modeling of FeCrAl cladding failure during loss of coolant conditions

Conference Gamble, K. A. ; Hales, J. D. ; Pastore, G. ; ...

Accident tolerant materials are defined as those that provide significantly increased response time in the event of an accident while providing similar or improved performance as the conventional UO2/Zircaloy-4 fuel rods during normal operation. One avenue for materials to demonstrate enhanced accident tolerance is to have improved reaction (e.g., oxidation) kinetics with steam, resulting in slower hydrogen (or other combustible gases) generation rate, while maintaining acceptable thermo-mechanical properties, fuel-clad interactions, and fission product behavior. Due to their improved reaction kinetics with steam compared to Zircaloy-4, iron-chromium-aluminum (FeCrAl) alloys are being considered as an accident tolerant cladding material. In order tomore »« less
Full Text Available
Acid Dissolution Behavior of Ferritic FeCrAl Tubes Candidates for Nuclear Fuel Cladding

Journal Article Rebak, Raul B. ; Yin, Liang ; Jurewicz, Timothy B. ; ... - Corrosion

The international materials community is engaged in finding safer alternatives to zirconium alloys for the cladding of fuel in light water reactors. One solution is to replace the zirconium cladding using ferritic iron-chromium-aluminum (FeCrAl) alloys, which offer extraordinary resistance to high-temperature reaction with air or steam due to the formation of a protective alumina layer on the external surface. It is important to characterize the behavior of FeCrAl not only during accident conditions but in the entire fuel cycle, which may include reprocessing of the used fuel after it is removed from the power reactors. The reprocessing may involve themore »« less
https://doi.org/10.5006/3965

Full Text Available
Neutronic analysis of candidate accident-tolerant cladding concepts in pressurized water reactors

Journal Article George, Nathan Michael ; Terrani, Kurt A. ; Powers, Jeffrey J. ; ... - Annals of Nuclear Energy

A study analyzed the neutronics of alternate cladding materials in a pressurized water reactor (PWR) environment. Austenitic type 310 (310SS) and 304 stainless steels, ferritic Fe-20Cr-5Al (FeCrAl) and APMT™ alloys, and silicon carbide (SiC)-based materials were considered and compared with Zircaloy-4. SCALE 6.1 was used to analyze the associated neutronics penalty/advantage, changes in reactivity coefficients, and spectral variations once a transition in the cladding was made. In the cases examined, materials containing higher absorbing isotopes invoked a reduction in reactivity due to an increase in neutron absorption in the cladding. Higher absorbing materials produced a harder neutron spectrum in themore »« less
Cited by 128
https://doi.org/10.1016/j.anucene.2014.09.005

Full Text Available

Similar Records