Publications related to Nuclear renaissance

Design of an Open-Loop Pile-Oscillation Program in the CROCUS Reactor

Andreas Pautz, Vincent Pierre Lamirand, Thomas Jean-François Ligonnet, Axel Guy Marie Laureau

As a follow-up to the CEA-EPFL PETALE experimental program on stainless steel nuclear data, the EPFL initiated an open-loop pile-oscillation experimental program in the CROCUS reactor: BLOOM. A reproduction of the critical experiments of PETALE, the progra ...

2024

Comparison of reduced model predictions for divertor detachment onset and reattachment timescales in ASDEX Upgrade and JET experiments

Holger Reimerdes, Haomin Sun

Building on prior analysis of ASDEX Upgrade (AUG) experiments (Henderson et al 2023 Nucl. Fusion 63 086024), this study compares simple analytical formula predictions for divertor detachment onset and reattachment timescales in JET experiments. Detachment ...

Iop Publishing Ltd2024

Filiform Variable Stiffness Technologies for Medical Robotics

Yegor Piskarev

In recent years, soft robotics has surged in applications like wearables, drones, smart fabrics, and medical instruments. Due to their compliance, these devices excel in tasks demanding dexterity and adaptability, such as manipulation, locomotion, crash re ...

EPFL2024

Analysis of the preliminary campaign for the PETALE program

Andreas Pautz, Vincent Pierre Lamirand, Thomas Jean-François Ligonnet, Axel Guy Marie Laureau

The PETALE program aims to provide new experimental data to constrain the stainless steel nuclear data. In this frame, a preliminary measurement campaign has been performed to characterize the neutron flux in key positions of the CROCUS reactor and to deve ...

EDP SCIENCES S A2023

Multi-physics modeling of VVERs with high fidelity high-resolution codes

Marianna Papadionysiou

The calculations performed for the design and operation of a Nuclear Power Plant (NPP) are a key factor for their safety analyses. The standard for the computational analysis of NPPs is the so called conventional approach, which relies on coarse mesh diffu ...

EPFL2022

Offshore Utility Systems for FPSOs: A Techno-Environomic Assessment Considering the Uncertainty About the Natural Gas Price

Due to restricted weight and space budget on floating production, storage and offloading units (FPSO), the offshore utility systems have been limited to low-efficiency energy technologies. Moreover, owing to time-varying energy demands of the FPSOs, the ex ...

2022

Uncertainty analyses of spent nuclear fuel decay heat calculations using SCALE modules

Andreas Pautz, Dimitri Rochman, Stefano Caruso

Decay heat residuals of spent nuclear fuel (SNF), i.e., the differences between calculations and measurements, were obtained previously for various spent fuel assemblies (SFA) using the Polaris module of the SCALE code system. In this paper, we compare dec ...

KOREAN NUCLEAR SOC2021

A Hybrid Approach to Neutron Transport with Thermal-hydraulic Feedback for Reactor Transient Analysis

Alexander Aures

Understanding the time-dependent behaviour of a nuclear reactor following an intended or unintended change of the reactor conditions is of crucial importance to the safe operation of nuclear reactors. Safety evaluations of nuclear reactors involve the anal ...

EPFL2021

Offshore utility systems for FPSOs: A techno-environomic assessment considering the uncertainty about the natural gas price

Daniel Alexander Florez Orrego

Due to restricted weight and space budget on floating production, storage and offloading units (FPSO), the offshore utility systems have been limited to low-efficiency energy technologies. Moreover, owing to time-varying energy demands of the FPSOs, the ex ...

2021

Uncertainty and sensitivity analysis of PWR mini-core transients in the presence of nuclear data uncertainty using non-parametric tolerance limits

Andreas Pautz, Winfried Zwermann, Alexander Aures

The impact of nuclear data uncertainties is studied for the reactor power and the reactivity during control rod withdrawal transients with reactivity insertions of 0.5

and 0.97

, respectively, for a PWR mini-core model. Multi-group cross sections, the mul ...

PERGAMON-ELSEVIER SCIENCE LTD2020

Nuclear renaissance

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Design of an Open-Loop Pile-Oscillation Program in the CROCUS Reactor

Comparison of reduced model predictions for divertor detachment onset and reattachment timescales in ASDEX Upgrade and JET experiments

Filiform Variable Stiffness Technologies for Medical Robotics

Analysis of the preliminary campaign for the PETALE program

Multi-physics modeling of VVERs with high fidelity high-resolution codes

Offshore Utility Systems for FPSOs: A Techno-Environomic Assessment Considering the Uncertainty About the Natural Gas Price

Uncertainty analyses of spent nuclear fuel decay heat calculations using SCALE modules

A Hybrid Approach to Neutron Transport with Thermal-hydraulic Feedback for Reactor Transient Analysis

Offshore utility systems for FPSOs: A techno-environomic assessment considering the uncertainty about the natural gas price

Uncertainty and sensitivity analysis of PWR mini-core transients in the presence of nuclear data uncertainty using non-parametric tolerance limits

Design of an Open-Loop Pile-Oscillation Program in the CROCUS Reactor

Offshore utility systems for FPSOs: A techno-environomic assessment considering the uncertainty about the natural gas price

Multi-physics modeling of VVERs with high fidelity high-resolution codes

Filiform Variable Stiffness Technologies for Medical Robotics

Comparison of reduced model predictions for divertor detachment onset and reattachment timescales in ASDEX Upgrade and JET experiments

Offshore Utility Systems for FPSOs: A Techno-Environomic Assessment Considering the Uncertainty About the Natural Gas Price

Analysis of the preliminary campaign for the PETALE program

Uncertainty and sensitivity analysis of PWR mini-core transients in the presence of nuclear data uncertainty using non-parametric tolerance limits

Uncertainty analyses of spent nuclear fuel decay heat calculations using SCALE modules

A Hybrid Approach to Neutron Transport with Thermal-hydraulic Feedback for Reactor Transient Analysis