Neutronic and burn‐up calculations of the ( ThO2‐UO2 ) pin cell benchmark using DRAGON5 and MCNP6 .2 codes with ENDF / B‐VIII .0 nuclear data library

Improvement of CINDER library for PWR burnup calculations

(2020) Ehsan Tavakkoli et al.   ANNALS OF NUCLEAR ENERGY

Development and validation of depletion code system IMPC-Burnup for ADS

(2019) Zhao-Qing Liu et al.   Nuclear Science and Techniques

Coupling of Neutronics and Thermal-Hydraulic Codes for Simulation of the MNSR Reactor

(2019) Jamal Al Zain et al.   NUCLEAR SCIENCE AND ENGINEERING

Feasibility study of thorium-fueled molten salt reactor with application in radioisotope production

(2019) Gangyang Zheng et al.   ANNALS OF NUCLEAR ENERGY

Thorium-based mixed oxide fuel in a pressurized water reactor: A burnup analysis with MCNP

(2018) Lucas P. Tucker et al.   ANNALS OF NUCLEAR ENERGY

ENDF/B-VIII.0: The 8 th Major Release of the Nuclear Reaction Data Library with CIELO-project Cross Sections, New Standards and Thermal Scattering Data

(2018) D.A. Brown et al.   NUCLEAR DATA SHEETS

Development of a MCNP5 and ORIGEN2 based burnup code for molten salt reactor

(2016) Guo-Min Sun et al.   Nuclear Science and Techniques

Nuclear reactions and physical models for neutron activation analysis

(2015) Andrej Trkov et al.   JOURNAL OF RADIOANALYTICAL AND NUCLEAR CHEMISTRY

Testing actinide fission yield treatment in CINDER90 for use in MCNP6 burnup calculations

(2015) Michael Lorne Fensin et al.   PROGRESS IN NUCLEAR ENERGY

Utilizing the burnup capability in MCNPX to perform depletion analysis of an MNSR fuel

(2014) E.K. Boafo et al.   ANNALS OF NUCLEAR ENERGY

Comparison of thermal scattering processing options for S(α,β) cards in MCNP

(2013) Štefan Čerba et al.   ANNALS OF NUCLEAR ENERGY

Neutronics and material attractiveness for PWR thorium systems using monte carlo techniques

(2011) Holly R. Trellue et al.   PROGRESS IN NUCLEAR ENERGY

Validation of IRBURN calculation code system through burnup benchmark analysis

(2010) S. Jafarikia et al.   ANNALS OF NUCLEAR ENERGY

Methods for Processing ENDF/B-VII with NJOY

(2010) R.E. MacFarlane et al.   NUCLEAR DATA SHEETS

Calculation of fuel burn up and radioactive inventory in a CANDU reactor using WIMS-D4 code

(2009) Zafar Yasin   ANNALS OF NUCLEAR ENERGY

Performance of advanced self-shielding models in DRAGON Version4 on analysis of a high conversion light water reactor lattice

(2007) Ramamoorthy Karthikeyan et al.   ANNALS OF NUCLEAR ENERGY