International audienceIn the framework of the WPNCS Expert Group on Uncertainty Analysis for Criticality Safety Assessment,we propose a blind numerical benchmark. Since the WPNCS Expert Group on Experimental Needs hadhighlighted the poor database of representative experiments of low-moderated MOX powders, this blindbenchmark is focused on MOX wet powders Three PuO$_2$ contents are proposed : 100%, 30% and 12.5%.Two Pu isotopic vectors are considered : a conservative one (0%, 71%, 17%, 11%, 1% for $^{238}$Pu up to$^{242}$Pu) and a realistic one obtained from reprocessed LWR 30 GWd/T fuels. For these 6 cases the powdermoisture rate is 3%. The MOX fissile medium is a sphere surrounded by 20 cm water reflector.The required results are the calcu...
International audienceAn overestimation of the $k_{eff}$ values for MOX fuels was identified with Mo...
The paper presents the results of a computational analysis of the OECD/NEA benchmark conducted to es...
AbstractIn the development of JENDL-4.0, the thermal and resonance capture cross section of 241Am wa...
International audienceIn the framework of the WPNCS Expert Group on Uncertainty Analysis for Critica...
International audienceThe use of mixed oxide (MOx) fuels in commercial Light Water Reactors is affec...
Abstract – Sensitivity and uncertainty analysis methodologies under development at Oak Ridge Nationa...
A series of experiments referred to as BFS/MOX was conducted in the BFS-1 experimental facility at I...
The paper presents calculation results for experiments performed to address needs of the MOX fuel ma...
The OECD Nuclear Energy Agency (NEA) Working Party on Nuclear Criticality Safety (WPNCS) identified ...
Fast reactor driver fuels are based on MOX with a high Pu content. They can be prepared by classical...
The international Expert Group on Reactor-Based Plutonium Disposition (TFRPD) has been established a...
The isotopic composition of mixed-oxide (MOX) fuel, fabricated with both uranium and plutonium, afte...
Several studies based on the JEFF-3.1.1 nuclear data library show a systematic overestimation of the...
One of the proposed ways to dispose of surplus weapons-grade plutonium (Pu) is to irradiate the high...
Simulations were performed using ORIGEN-ARP to investigate ¹³⁷Cs and ¹⁰⁶Ru-¹⁰⁶Rh as suitable fission...
International audienceAn overestimation of the $k_{eff}$ values for MOX fuels was identified with Mo...
The paper presents the results of a computational analysis of the OECD/NEA benchmark conducted to es...
AbstractIn the development of JENDL-4.0, the thermal and resonance capture cross section of 241Am wa...
International audienceIn the framework of the WPNCS Expert Group on Uncertainty Analysis for Critica...
International audienceThe use of mixed oxide (MOx) fuels in commercial Light Water Reactors is affec...
Abstract – Sensitivity and uncertainty analysis methodologies under development at Oak Ridge Nationa...
A series of experiments referred to as BFS/MOX was conducted in the BFS-1 experimental facility at I...
The paper presents calculation results for experiments performed to address needs of the MOX fuel ma...
The OECD Nuclear Energy Agency (NEA) Working Party on Nuclear Criticality Safety (WPNCS) identified ...
Fast reactor driver fuels are based on MOX with a high Pu content. They can be prepared by classical...
The international Expert Group on Reactor-Based Plutonium Disposition (TFRPD) has been established a...
The isotopic composition of mixed-oxide (MOX) fuel, fabricated with both uranium and plutonium, afte...
Several studies based on the JEFF-3.1.1 nuclear data library show a systematic overestimation of the...
One of the proposed ways to dispose of surplus weapons-grade plutonium (Pu) is to irradiate the high...
Simulations were performed using ORIGEN-ARP to investigate ¹³⁷Cs and ¹⁰⁶Ru-¹⁰⁶Rh as suitable fission...
International audienceAn overestimation of the $k_{eff}$ values for MOX fuels was identified with Mo...
The paper presents the results of a computational analysis of the OECD/NEA benchmark conducted to es...
AbstractIn the development of JENDL-4.0, the thermal and resonance capture cross section of 241Am wa...