Reactor Physics Methods and Analysis Capabilities in SCALE

DeHart, Mark D.; Bowman, Stephen M.

doi:10.13182/nt174-196

Cited by 74 publications

(28 citation statements)

References 8 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…The neutronic behavior is performed using TRITON and KENO-VI modules from SCALE 6.1 [16][17][18]. SCALE provides a "plug-and-play" framework with 89 computational modules, including three Monte Carlo and three deterministic radiation transport solvers.…”

Section: Methodology and Input Parametersmentioning

confidence: 99%

Neutronic Analysis on Potential Accident Tolerant Fuel-Cladding Combination U₃Si₂-FeCrAl

Chen

Yuan

2017

Science and Technology of Nuclear Installations

View full text Add to dashboard Cite

Neutronic performance is investigated for a potential accident tolerant fuel (ATF), which consists of U 3 Si 2 fuel and FeCrAl cladding. In comparison with current UO 2 -Zr system, FeCrAl has a better oxidation resistance but a larger thermal neutron absorption cross section. U 3 Si 2 has a higher thermal conductivity and a higher uranium density, which can compensate the reactivity suppressed by FeCrAl. Based on neutronic investigations, a possible U 3 Si 2 -FeCrAl fuel-cladding system is taken into consideration. Fundamental properties of the suggested fuel-cladding combination are investigated in a fuel assembly. These properties include moderator and fuel temperature coefficients, control rods worth, radial power distribution (in a fuel rod), and different void reactivity coefficients. The present work proves that the new combination has less reactivity variation during its service lifetime. Although, compared with the current system, it has a little larger deviation on power distribution and a little less negative temperature coefficient and void reactivity coefficient and its control rods worth is less important, variations of these parameters are less important during the service lifetime of fuel. Hence, U 3 Si 2 -FeCrAl system is a potential ATF candidate from a neutronic view.

show abstract

Section: Methodology and Input Parametersmentioning

confidence: 99%

Neutronic Analysis on Potential Accident Tolerant Fuel-Cladding Combination U₃Si₂-FeCrAl

Chen

Yuan

2017

Science and Technology of Nuclear Installations

View full text Add to dashboard Cite

show abstract

“…A set of simplified reactivity calculations is performed using SCALE/TRITON from the SCALE 6.1 software system package to model a single fuel rod [6][7][8]. The CENTRM module from SCALE was used to determine cross-section approximations in one dimension; the continuous energy spectrum was flux weighted in order to produce resonance-shielded multi-group cross-section data.…”

Section: Methodology and Input Parametersmentioning

confidence: 99%

Neutronic analysis of candidate accident-tolerant iron alloy cladding concepts

George¹,

Terrani²,

Powers³

2013

View full text Add to dashboard Cite

“…Monte Carlo depletion SCALE 6.1 provided MG Monte Carlo depletion that coupled SCALE MG cross-section processing capabilities with KENO and ORIGEN (Oak Ridge Isotope GENeration) via the TRITON sequence (DeHart and Bowman, 2011). In addition to the existing MG capability, a new CE-based KENO/ORIGEN Monte Carlo depletion capability has been developed for SCALE 6.2 and can be utilized by simply changing the input library specification (Goluoglu et al, 2012).…”

Section: Samplermentioning

confidence: 99%

Monte Carlo capabilities of the SCALE code system

Rearden

Petrie

Peplow

et al. 2015

Annals of Nuclear Energy

View full text Add to dashboard Cite

a b s t r a c t SCALE is a widely used suite of tools for nuclear systems modeling and simulation that provides comprehensive, verified and validated, user-friendly capabilities for criticality safety, reactor physics, radiation shielding, and sensitivity and uncertainty analysis. For more than 30 years, regulators, licensees, and research institutions around the world have used SCALE for nuclear safety analysis and design. SCALE provides a ''plug-and-play'' framework that includes three deterministic and three Monte Carlo radiation transport solvers that can be selected based on the desired solution, including hybrid deterministic/ Monte Carlo simulations. SCALE includes the latest nuclear data libraries for continuous-energy and multigroup radiation transport as well as activation, depletion, and decay calculations. SCALE's graphical user interfaces assist with accurate system modeling, visualization, and convenient access to desired results. SCALE 6.2 will provide several new capabilities and significant improvements in many existing features, especially with expanded continuous-energy Monte Carlo capabilities for criticality safety, shielding, depletion, and sensitivity and uncertainty analysis. An overview of the Monte Carlo capabilities of SCALE is provided here, with emphasis on new features for SCALE 6.2.

show abstract

Reactor Physics Methods and Analysis Capabilities in SCALE

Cited by 74 publications

References 8 publications

Neutronic Analysis on Potential Accident Tolerant Fuel-Cladding Combination U₃Si₂-FeCrAl

Neutronic Analysis on Potential Accident Tolerant Fuel-Cladding Combination U₃Si₂-FeCrAl

Neutronic analysis of candidate accident-tolerant iron alloy cladding concepts

Monte Carlo capabilities of the SCALE code system

Contact Info

Product

Resources

About

Reactor Physics Methods and Analysis Capabilities in SCALE

Cited by 74 publications

References 8 publications

Neutronic Analysis on Potential Accident Tolerant Fuel-Cladding Combination U3Si2-FeCrAl

Neutronic Analysis on Potential Accident Tolerant Fuel-Cladding Combination U3Si2-FeCrAl

Neutronic analysis of candidate accident-tolerant iron alloy cladding concepts

Monte Carlo capabilities of the SCALE code system

Contact Info

Product

Resources

About

Neutronic Analysis on Potential Accident Tolerant Fuel-Cladding Combination U₃Si₂-FeCrAl

Neutronic Analysis on Potential Accident Tolerant Fuel-Cladding Combination U₃Si₂-FeCrAl