SiC and FeCrAl as Potential Cladding Materials for APR-1400 Neutronic Analysis

The aim of this study is to investigate the potential improvement of accident-tolerant fuels in pressurized water reactors for replacing existing reference zircaloy (Zr) fuel-cladding systems. Three main strategies for improving accident-tolerant fuels are investigated: enhancement of the present state-of-the-art zirconium fuel-cladding system to improve oxidation resistance, replacement of the current referenced fuel-cladding system material with an alternative high-performance oxidation-resistant cladding, and replacement of the current fuel with alternative fuel forms. This study focuses on a preliminary analysis of the neutronic behavior and properties of silicon carbide (SiC)-fuel and FeCrAl cladding systems, which provide a better safety margin as accident-tolerant fuel systems for pressurized water reactors. The typical physical behavior of both cladding systems is investigated to determine their general neutronic performance. The multiplication factor, thermal neutron flux spectrum, 239Pu inventory, pin power distribution, and radial power are analyzed and compared with those of a reference Zr fuel-cladding system. Furthermore, the effects of a burnable poison rod (Gd2O3) in different fuel assemblies are investigated. SiC cladding assemblies present a softer neutron spectrum and a lower linear power distribution compared with the conventional Zr-fuel-cladding system. Additionally, the SiC fuel-cladding system exhibits behaviors that are consistent with the neutronic behavior of conventional Zr fuel-cladding systems, thereby affording greater economic and safety improvements.

[1]  C. Jang,et al.  Effects of ATF cladding properties on PWR responses to an SBO accident: A sensitivity analysis , 2022, Annals of Nuclear Energy.

[2]  S. Alameri,et al.  Preliminary neutronic analysis of alternative cladding materials for APR-1400 fuel assembly , 2021, Nuclear Engineering and Design.

[3]  S. Alameri,et al.  Preliminary three-dimensional neutronic analysis of IFBA coated TRISO fuel particles in prismatic-core advanced high temperature reactor , 2021 .

[4]  M. Kurata,et al.  Steam Oxidation of Silicon Carbide at High Temperatures for the Application as Accident Tolerant Fuel Cladding, an Overview , 2021, Thermo.

[5]  Yonghee Kim,et al.  Impacts of an ATF cladding on neutronic performances of the soluble‐boron‐free ATOM core , 2020, International Journal of Energy Research.

[6]  S. Alameri,et al.  Preliminary Study of a Prismatic-Core Advanced High-Temperature Reactor Fuel Using Homogenization Double-Heterogeneous Method , 2020, Nuclear Science and Engineering.

[7]  American Nuclear Society (ANS) , 2018, The Grants Register 2022.

[8]  State-of-the-Art Report on Light Water Reactor Accident-Tolerant Fuels , 2018, Nuclear Science.

[9]  Mohammad Alrwashdeh,et al.  239Pu evaluation comparison study , 2018, Annals of Nuclear Energy.

[10]  Jaakko Leppänen,et al.  On the use of delta-tracking and the collision flux estimator in the Serpent 2 Monte Carlo particle transport code , 2017 .

[11]  Wang Kan,et al.  U233 Data Evaluation for Criticality Study , 2016 .

[12]  M. Fratoni,et al.  Neutronic evaluation of coating and cladding materials for accident tolerant fuels , 2016 .

[13]  Chaohui He,et al.  Re-evaluation of neutron displacement cross sections for silicon carbide by a Monte Carlo approach , 2016 .

[14]  Tomasz Kozlowski,et al.  Neutronics and fuel performance evaluation of accident tolerant FeCrAl cladding under normal operation conditions , 2015 .

[15]  Nathan M George,et al.  Neutronic analysis of candidate accident-tolerant cladding concepts in pressurized water reactors , 2015 .

[16]  R. S. Sen,et al.  Impact of Accident-Tolerant Fuels and Claddings on the Overall Fuel Cycle: A Preliminary Systems Analysis , 2014 .

[17]  Shannon M. Bragg-Sitton,et al.  Evaluation Metrics Applied to Accident Tolerant Fuels , 2014 .

[18]  J. Leppänen,et al.  Internal Neutronics-Temperature Coupling in Serpent 2 , 2014 .

[19]  Tuomas Viitanen,et al.  The Serpent Monte Carlo Code: Status, Development and Applications in 2013 , 2014, ICS 2014.

[20]  Steven J. Zinkle,et al.  Accident tolerant fuels for LWRs: A perspective , 2014 .

[21]  S. Zinkle,et al.  Advanced Oxidation Resistant Iron-Based Alloys for LWR Fuel Cladding , 2014 .

[22]  Kevin R Robb,et al.  Preliminary assessment of accident-tolerant fuels on LWR performance during normal operation and under DB and BDB accident conditions , 2014 .

[23]  W. Lee,et al.  Fully Ceramic Microencapsulated (FCM) Replacement Fuel for LWRs , 2013 .

[24]  Kurt A. Terrani,et al.  Protection of Zirconium by Alumina- and Chromia-Forming Iron Alloys under High-Temperature Steam Exposure , 2013 .

[25]  B. Pritychenko,et al.  Neutron Thermal Cross Sections, Westcott Factors, Resonance Integrals, Maxwellian Averaged Cross Sections and Astrophysical Reaction Rates Calculated from the ENDF/B-VII.1, JEFF-3.1.2, JENDL-4.0, ROSFOND-2010, CENDL-3.1 and EAF-2010 Evaluated Data Libraries , 2012, 1208.2879.

[26]  Van Der Marck,et al.  Benchmarking ENDF/B-VII.1, JENDL-4.0 and JEFF-3.1 , 2012 .

[27]  E. Fridman,et al.  On the use of the Serpent Monte Carlo code for few-group cross section generation , 2011 .

[28]  Jeffrey J. Powers,et al.  A review of TRISO fuel performance models , 2010 .

[29]  Y. Katoh,et al.  Handbook of SiC properties for fuel performance modeling , 2007 .

[30]  S. Mughabghab.  Thermal neutron capture cross sections resonance integrals and g-factors , 2003 .

[31]  Takuya Ito,et al.  Benchmark Problem Suite for Reactor Physics Study of LWR Next Generation Fuels , 2002 .

[32]  Dong-Su Kim,et al.  APR1400 - Development Status and Design Features , 2002 .

[33]  W. Lippmann,et al.  Investigation of the use of ceramic materials in innovative light water reactor : fuel rod concepts , 2001 .

[34]  G. M. Kim,et al.  Multi-layer coating of silicon carbide and pyrolytic carbon on UO , 2000 .

[35]  V. Krstić,et al.  Silicon carbide as an inert-matrix for a thermal reactor fuel , 1999 .

[36]  Guan Jie,et al.  Improvement in oxidation resistance of the nuclear graphite by reaction-coated SiC coating , 1995 .

[37]  R. Schulten,et al.  The HTR with SiC-technology , 1993 .

[38]  R. J. Price,et al.  Properties of silicon carbide for nuclear fuel particle coatings , 1977 .

[39]  P. Allen,et al.  Nuclear Fuel Coated Particle Development in the Reactor Fuel Element Laboratories of the U.K. Atomic Energy Authority , 1977 .

[40]  Roko Bujas,et al.  The nuclear fuel element , 1972 .