Publications de Andreas Pautz | EPFL Graph Search

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Following the growing interest that we observe in the use of OpenFOAM for nuclear reactor analysis, the current work presents a point-kinetics model and associated open-source code that can easily be inte-grated in existing thermal-hydraulic solvers to str ...

The present article gives an overview of the first experimental campaigns carried out in the AKR-2 and CROCUS reactors within the framework of the Horizon 2020 European project CORTEX. CORTEX aims at developing innovative core monitoring techniques that al ...

Currently, the pin-by-pin homogenized solvers are a very active research field as they can, unlike the nodal codes, directly predict the local power, while requiring significantly less computational resources than the heterogeneous transport codes. This pa ...

The operation of a reactor on an open but self-sustainable cycle without actinide separation is known as breed-and-burn. It has mostly been envisioned for use in solid-fueled fast-spectrum reactors such as sodium-cooled fast reactors. In this paper the app ...

The advancement of high-fidelity nuclear reactor simulation methods is essential for reducing conservatism in safety margins and optimizing reactor design and operation. This work evaluates the predictive capabilities of the deterministic solvers GeN-Foam ...

In a collaboration between Ecole Polytechnique Fédérale de Lausanne (EPFL) and CEA, in the fall of 2020, the experimental Programme d’Étude en Transmission de l’Acier Lourd et ses Eléments (PETALE) was successfully carried out in the CROCUS reactor of EPFL ...

This paper summarizes the results of modeling methodologies developed for the zero-power (100 W) teaching and research reactor CROCUS located in the Laboratory for Reactor Physics and Systems Behavior (LRS) at the Swiss Federal Institute of Technology in L ...

This paper summarizes a methodology developed at Ecole Polytechnique Federale de Lausanne for the neutronic modeling of the CROCUS experimental reactor and proposes solutions to the challenges one may face while modeling a research reactor with a complex g ...

In this paper we followed the functional data analysis (FDA) paradigm to summarize the output of computer code simulating a reflood transient. The paradigm stands on the assumption that the data (code output) is generated by a smooth, albeit complex, proce ...

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 ...

Decay heat calculations of spent nuclear fuel (SNF) using Polaris and ORIGEN codes in the SCALE code sys-tem, and CASMO5 code, are validated using measurements from the Clab and GE-Morris facilities. Multiple hypothesis testing, relying on permutations and ...

In the pebble-bed high-temperature reactor under construction in China, called the HTR-PM, the spherical fuel elements continuously flow downward in the cylindrical core. The burnup of each pebble is checked at the core outlet and, according to the achieve ...

Nuclear data uncertainty propagation based on stochastic sampling ( SS) is becoming more attractive while leveraging modern computer power. Two variants of the SS approach are compared in this paper. The Total Monte Carlo (TMC) method by the Nuclear Resear ...

The PETALE experimental programme in the CROCUS reactor intends to provide integral measurements to constrain stainless steel nuclear data. This article presents the tools and the methodology developed to design and optimize the experiments, and its operat ...

The present article describes the preliminary validation study of simulated in-core and reflector n eutron spectra in preparation of oncoming experimental programs in the zero-power reactor CROCUS at EPFL. For this purpose, a set of activation foils were i ...

The paper demonstrates the influence of uncertainties in microscopic nuclear data on the results of reactor simulations, both for stationary and transient states. It gives an overview on the methods in use for uncertainty and sensitivity analyses with resp ...

CROCUS is a zero power (100 W) reactor of the Laboratory for Reactor Physics and Systems Behavior (LRS) at the Swiss Federal Institute of Technology, Lausanne (EPFL). It is used for teaching and research purposes. Its modeling has relied so far on diffusio ...

Nuclear data, especially fission yields, create uncertainties in the predicted concentrations of fission products in spent fuel. Herein, we present a new framework that extends data assimilation methods to burnup simulations by using data from post-irradia ...

Recent numerical advancements at the Laboratory for Reactor Physics and Systems Behaviour (LRS) have led to the development of a novel two-phase porous medium-based thermal-hydraulic simulation tool, integrated in the broader multi-physics framework of the ...

In this study, an optimization of the canister loading for used nuclear fuels is performed, solely based on their decay heat values. By taking into account the individual irradiation history for each fuel assembly, as well as with a number of assumptions o ...

Andreas Pautz

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Development of a point-kinetics model in OpenFOAM, integration in GeN-Foam, and validation against FFTF experimental data

Neutron noise experiments in the AKR-2 and CROCUS reactors for the European project CORTEX

Development and verification of pin-by-pin homogenized simplified transport solver Tortin for PWR core analysis

Breed-and-burn fuel cycle in molten salt reactors

Comparison of high-fidelity transport code predictions against the NECTAR intra-pin reaction rate measurements

High-Resolution Modeling Without Computation Slowdown for PETALE in CROCUS

Full Core modeling techniques for research reactors with irregular geometries using Serpent and PARCS applied to the CROCUS reactor

Investigation and Validation of Unstructured Mesh Methodologies for Modeling Experimental Reactors

Exploring Variability in Reflood Simulation Results: an Application of Functional Data Analysis

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

Validation of spent nuclear fuel decay heat calculations using Polaris, ORIGEN and CASMO5

Statistical Burnup Distribution of Moving Pebbles in the High-Temperature Reactor HTR-PM

Comparison of Two Approaches for Nuclear Data Uncertainty Propagation in MCNPX for Selected Fast Spectrum Critical Benchmarks

Uncertainty propagation for the design study of the PETALE experimental programme in the CROCUS reactor

In-core dosimetry for the validation of neutron spectra in the CROCUS reactor

Reactor simulations with nuclear data uncertainties

Verification of a reactor physics calculation scheme for the CROCUS reactor

Data assimilation of post irradiation examination experiments to adjust fission yields

Development of a novel two-phase flow solver for nuclear reactor analysis: Validation against sodium boiling experiments

Loading optimization for Swiss used nuclear fuel assemblies into final disposal canisters