Journal of Nuclear Science, Engineering and Technology (JONSAT)
In cooperation with the Iranian Nuclear Society

Investigating new cladding material for NuScale reactor based on neutronic and thermo-mechanical properties using M5, E110, Zircaloy 4 and FeCrAl claddings

Document Type : Research Paper

Authors

H. Zayermohammadi Rishehri
M. Zaidabadi Nejad

Nuclear Engineering Department, Faculty of Sciences and Modern Technologies, Graduate University of Advanced Technology, P.O.Box: 7631885356, Kerman - Iran

https://doi.org/10.24200/nst.2023.1246.1811
Abstract
Nuclear fuel clads are among the most important components of a nuclear reactor. These clads are typically made of zirconium alloy; However these alloys do not possess the best thermo-mechanical properties. Therefore, in this study we attempted to achieve a clad with more suitable neutronic and thermo-mechanical properties by changing the clad materials used in the NuScale reactor. The changes in the material used in the reactor core cause a change in the effective multiplication factor as well as the length of the refueling interval. These effective multiplication factors and refueling period were obtained as the most important neutronic results within this research through simulating the NuScale reactor that uses the M5 cladding as a reference reactor and furthermore, the simulation of this reactor accommodating a clad change with three other types of common alloys, two of which are zirconium alloy and one of which is FeCrAl alloy. The results reveal that zirconium alloys in tandem with different thermo-mechanical properties are not much different in the neutronic matter of the core. In addition, FeCrAl alloy however endowed with excellent thermo-mechanical properties has a negative compact on the effective multiplication factor and the duration of the fueling period. Ultimately, in order to compensate for the negative effect of FeCrAl alloy on the neutronic properties, two solutions were used; The solution of using FeCrAl as a coating on a zirconium clad which presented outstanding neutronic results.

Highlights

  1. International Atomic Energy Agency (IAEA), Advances in Small Modular Reactor Technology Developments, IAEA (2014).

 

  1. Ingersoll D.T, Carelli M.D. Handbook of small modular nuclear reactors. Elsevier. (2014).

 

  1. Zinkle Steven J, Kurt A. Terrani, Jess C. Gehin, Larry J. Ott, Lance Lewis Snead. Accident tolerant fuels for LWRs: A perspective. Journal of Nuclear Materials. 2014;448(1-3).

 

  1. Pint B.A, Terrani K.A, Brady M.P, Cheng T, Keiser J.R, High temperature oxidation of fuel cladding candidate materials in steam–hydrogen environments. J. Nucl. Mater. 2013;440.

 

  1. Pan D, Zhang R, Wang H, Lu C, Liu Y. Formation and stability of oxide layer in FeCrAl fuel cladding material under high-temperature steam. Journal of Alloys and Compounds. 2016;684.

 

  1. Chen P, Qiu B, Wu J, Zheng M, Gao S, Zhang K, Zhou Y, Wu Y. A comparative study of in-pile behaviors of FeCrAl cladding under normal and accident conditions with updated FROBA-ATF code, Nuclear Engineering and Design. 2021;371.

 

  1. Quadakkers W.J, Elschner A, Speier W, Nickel H. Composition and growth mechanisms of alumina scales on FeCrAl-based alloys determined by SNMS. Appl. Surf. Sci. 1991;52.

 

  1. Badini C, Laurella F. Oxidation of fecral alloy: influence of temperature and atmosphere on scale growth rate and mechanism. Surf. Coat. Technol. 2001;135.

 

  1. Nychka J.A, Clarke D.R. Quantification of aluminum outward diffusion during oxidation of FeCrAl alloys. Oxid. Met. 2005;63.

 

  1. NuScale Power LLC, NuScale Standard Plant Design Certification Application. US Nuclear Regulatory Commission (NRC). (2016).

 

  1. Sadegh-Noedoost A, Faghihi F, Fakhraei A, Amin-Mozafari, M. Investigations of the fresh-core cycle-length and the average fuel depletion analysis of the NuScale core. Annals of Nuclear Energy. 2020;136.

 

  1. Ott L.J, Robb K.R, Wang D. Preliminary assessment of accident-tolerant fuels on LWR performance during normal operation and under DB and BDB accident conditions. Journal of Nuclear Materials. 2014;448(1-3).

 

  1. Wu X, Kozlowski T, Hales J.D. Neutronics and fuel performance evaluation of accident tolerant FeCrAl cladding under normal operation conditions. Annals of Nuclear Energy. 2015;85.

 

  1. Lemaignan C. 2.07-Zirconium alloys: properties and characteristics. Comprehensive Nuclear Materials. 2012.

 

  1. George N.M, Terrani K, Powers J, Worrall A, Maldonado I. Neutronic analysis of candidate accident-tolerant cladding concepts in pressurized water reactors. Annals of Nuclear Energy. 2015;75.

 

  1. Okumura K, Oka Y, Ishiwatari Y. Nuclear reactor design. Springer Japan 2014.

 

  1. Donnelly J.V. WIMS-CRNL: A user's manual for the Chalk River version of WIMS. No. AECL--8955. Atomic Energy of Canada Ltd. 1986.

 

  1. Fowler T.B, Vondy D.R. Nuclear reactor core analysis code: CITATION. No. ORNL-TM-2496. Oak Ridge National Lab. Tenn. 1969.

 

  1. Koo Y.H, Yang J.H, Park J.Y, Kim K.S, Kim H.G, Kim D.J, Jung Y.I, Song K.W. KAERI’s development of LWR accident-tolerant fuel. Nuclear Technology. 2014;186(2).

 

  1. Kim H.G, Kim I.H, Jung Y.I, Park D.J, Park J.Y, Koo Y.H. High-temperature oxidation behavior of Cr-coated zirconium alloy. In Proceeding of LWR Fuel Performance Meeting/TopFuel. Charlotte. USA. 2013;842:846.

 

  1. Yacout A.M, Pellin M, Billone M.C. Development and Testing of Nanolaminate Coatings for Conventional LWR Cladding. Transaction of TopFuel. 2013.

 

  1. Park J.Y, Kim I.H, Jung Y.I, Kim H.G, Park D.J, Choi B.K. High temperature steam oxidation of Al3Ti-based alloys for the oxidation-resistant surface layer on Zr fuel claddings. Journal of Nuclear Materials. 2013;437(1-3).

 

  1. Wang Y, Zhou W, Wen Q, Ruan X, Luo F, Bai G, Qing Y. Zhu D, Huang Z, Zhang Y, Liu T, Li R. Behavior of plasma sprayed Cr coatings and FeCrAl coatings on Zr fuel cladding under loss-of-coolant accident conditions. Surface and Coatings Technology. 2018;344.

 

  1. Park D.J, Kim H.G, Jung Y.I, Park J.H, Yang J.H, Koo Y.H. Microstructure and mechanical behavior of Zr substrates coated with FeCrAl and Mo by cold-spraying. Journal of Nuclear Materials. 2018;504.

 

  1. Kim I.H, Jung Y.I, Kim H.G, Jang J.I. Oxidation-resistant coating of FeCrAl on Zr-alloy tubes using 3D printing direct energy deposition. Surface and Coatings Technology. 2021;411.

 

  1. Okamoto H. Fe-Zr (iron-zirconium). Journal of Phase Equilibria. 1993;14(5).

 

  1. Dabney T, Johnson G, Yeom H, Maier B, Walters J, Sridharan K. Experimental evaluation of cold spray FeCrAl alloys coated zirconium-alloy for potential accident tolerant fuel cladding. Nuclear Materials and Energy. 2019;21.

Keywords

Zirconium alloys
FeCrAl alloy
Fuel cladding
NuScale reactor
  1. International Atomic Energy Agency (IAEA), Advances in Small Modular Reactor Technology Developments, IAEA (2014).

 

  1. Ingersoll D.T, Carelli M.D. Handbook of small modular nuclear reactors. Elsevier. (2014).

 

  1. Zinkle Steven J, Kurt A. Terrani, Jess C. Gehin, Larry J. Ott, Lance Lewis Snead. Accident tolerant fuels for LWRs: A perspective. Journal of Nuclear Materials. 2014;448(1-3).

 

  1. Pint B.A, Terrani K.A, Brady M.P, Cheng T, Keiser J.R, High temperature oxidation of fuel cladding candidate materials in steam–hydrogen environments. J. Nucl. Mater. 2013;440.

 

  1. Pan D, Zhang R, Wang H, Lu C, Liu Y. Formation and stability of oxide layer in FeCrAl fuel cladding material under high-temperature steam. Journal of Alloys and Compounds. 2016;684.

 

  1. Chen P, Qiu B, Wu J, Zheng M, Gao S, Zhang K, Zhou Y, Wu Y. A comparative study of in-pile behaviors of FeCrAl cladding under normal and accident conditions with updated FROBA-ATF code, Nuclear Engineering and Design. 2021;371.

 

  1. Quadakkers W.J, Elschner A, Speier W, Nickel H. Composition and growth mechanisms of alumina scales on FeCrAl-based alloys determined by SNMS. Appl. Surf. Sci. 1991;52.

 

  1. Badini C, Laurella F. Oxidation of fecral alloy: influence of temperature and atmosphere on scale growth rate and mechanism. Surf. Coat. Technol. 2001;135.

 

  1. Nychka J.A, Clarke D.R. Quantification of aluminum outward diffusion during oxidation of FeCrAl alloys. Oxid. Met. 2005;63.

 

  1. NuScale Power LLC, NuScale Standard Plant Design Certification Application. US Nuclear Regulatory Commission (NRC). (2016).

 

  1. Sadegh-Noedoost A, Faghihi F, Fakhraei A, Amin-Mozafari, M. Investigations of the fresh-core cycle-length and the average fuel depletion analysis of the NuScale core. Annals of Nuclear Energy. 2020;136.

 

  1. Ott L.J, Robb K.R, Wang D. Preliminary assessment of accident-tolerant fuels on LWR performance during normal operation and under DB and BDB accident conditions. Journal of Nuclear Materials. 2014;448(1-3).

 

  1. Wu X, Kozlowski T, Hales J.D. Neutronics and fuel performance evaluation of accident tolerant FeCrAl cladding under normal operation conditions. Annals of Nuclear Energy. 2015;85.

 

  1. Lemaignan C. 2.07-Zirconium alloys: properties and characteristics. Comprehensive Nuclear Materials. 2012.

 

  1. George N.M, Terrani K, Powers J, Worrall A, Maldonado I. Neutronic analysis of candidate accident-tolerant cladding concepts in pressurized water reactors. Annals of Nuclear Energy. 2015;75.

 

  1. Okumura K, Oka Y, Ishiwatari Y. Nuclear reactor design. Springer Japan 2014.

 

  1. Donnelly J.V. WIMS-CRNL: A user's manual for the Chalk River version of WIMS. No. AECL--8955. Atomic Energy of Canada Ltd. 1986.

 

  1. Fowler T.B, Vondy D.R. Nuclear reactor core analysis code: CITATION. No. ORNL-TM-2496. Oak Ridge National Lab. Tenn. 1969.

 

  1. Koo Y.H, Yang J.H, Park J.Y, Kim K.S, Kim H.G, Kim D.J, Jung Y.I, Song K.W. KAERI’s development of LWR accident-tolerant fuel. Nuclear Technology. 2014;186(2).

 

  1. Kim H.G, Kim I.H, Jung Y.I, Park D.J, Park J.Y, Koo Y.H. High-temperature oxidation behavior of Cr-coated zirconium alloy. In Proceeding of LWR Fuel Performance Meeting/TopFuel. Charlotte. USA. 2013;842:846.

 

  1. Yacout A.M, Pellin M, Billone M.C. Development and Testing of Nanolaminate Coatings for Conventional LWR Cladding. Transaction of TopFuel. 2013.

 

  1. Park J.Y, Kim I.H, Jung Y.I, Kim H.G, Park D.J, Choi B.K. High temperature steam oxidation of Al3Ti-based alloys for the oxidation-resistant surface layer on Zr fuel claddings. Journal of Nuclear Materials. 2013;437(1-3).

 

  1. Wang Y, Zhou W, Wen Q, Ruan X, Luo F, Bai G, Qing Y. Zhu D, Huang Z, Zhang Y, Liu T, Li R. Behavior of plasma sprayed Cr coatings and FeCrAl coatings on Zr fuel cladding under loss-of-coolant accident conditions. Surface and Coatings Technology. 2018;344.

 

  1. Park D.J, Kim H.G, Jung Y.I, Park J.H, Yang J.H, Koo Y.H. Microstructure and mechanical behavior of Zr substrates coated with FeCrAl and Mo by cold-spraying. Journal of Nuclear Materials. 2018;504.

 

  1. Kim I.H, Jung Y.I, Kim H.G, Jang J.I. Oxidation-resistant coating of FeCrAl on Zr-alloy tubes using 3D printing direct energy deposition. Surface and Coatings Technology. 2021;411.

 

  1. Okamoto H. Fe-Zr (iron-zirconium). Journal of Phase Equilibria. 1993;14(5).

 

  1. Dabney T, Johnson G, Yeom H, Maier B, Walters J, Sridharan K. Experimental evaluation of cold spray FeCrAl alloys coated zirconium-alloy for potential accident tolerant fuel cladding. Nuclear Materials and Energy. 2019;21.

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