Publications de Mathieu Hursin

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Analyzing neutron noise in zero-power reactors provides information about integral parameters, such as the prompt decay constant

\alpha

alpha. This information can be used to validate codes that predict said parameters, or time-dependent behavior, part ...

Recent fuel anomalies at a Swiss power plant have motivated the development at PSI of the "Cycle Check-Up" (CHUP) methodology which consists in performing Monte Carlo transport calculations using nuclide compositions and thermal-hydraulic boundary conditio ...

In response to the need for validating high-fidelity deterministic neutronics solvers capable of providing pin-resolved neutron flux distributions, intra-pin reaction rates measurements were performed in the CROCUS experimental reactor. A specially designe ...

The design and safety analysis of advanced fuel materials and innovative Gen-IV reactor concepts, with their distinct physical and engineering requirements compared to conventional reactors, has often been the main drive behind the development of modern mu ...

In the CORTEX project, methods to simulate neutron flux oscillations were enhanced and machine-learning based tools to determine the causes of measured neutron flux oscillations were developed, using the results of simulations as training and validation da ...

Monte Carlo (MC) based neutron transport codes provide high-fidelity numerical solutions to problems beyond the modelling capabilities of conventional core simulation methods. However, performing core-follow calculations with MC codes remains challenging d ...

In this work, we explore the use of an iterative Bayesian Monte Carlo (iBMC) method for nuclear data evaluation within a TALYS Evaluated Nuclear Data Library (TENDL) framework. The goal is to probe the model and parameter space of the TALYS code system to ...

Sodium-cooled fast reactors (SFRs) offer a promising alternative to conventional light-water technologies but present unique safety and design challenges. Among others, reactivity feedback effects associated to core deformations are significantly larger th ...

The paper describes the source term estimation of CROCUS, the zero power research reactor of EPFL, to be used for dispersion analysis under accidental conditions. To fulfil regulatory requirements, the source term of the CROCUS fuel is estimated through Mo ...

The effect of nuclear data (fission yields, cross sections and emitted spectra) is quantified for spent nuclear fuel assemblies from a realistic boiling water reactor operated over 25 cycles. Nominal calculations are performed with the CASMO5, SIMULATE-3 a ...

Determination of the local void fraction in BWRs from in-core neutron noise measurements requires the knowledge of the axial velocity of the void. The purpose of this paper is to revisit the problem of determining the axial void velocity profile from the t ...

In response to the need for validating high-fidelity deterministic neutronics solvers capable of pin-resolved neutron flux distributions, an instrumented fuel rod was designed for the experimental reactor CROCUS, operated by EPFL’s Laboratory for Reactor P ...

This paper presents a validation exercise for a natural circulation cooled, light-water moderated and cooled pool-type reactor system. Measurements were performed in the 100 kW Training Reactor of the Budapest University of Technology and Economics, which ...

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 ...

CORTEX is a EU H2020 project (2017-2021) devoted to the analysis of 'reactor neutron noise' in nuclear reactors, i.e. the small fluctuations occurring around the stationary state due to external or internal disturbances in the core. One important aspect of ...

This paper presents nuclear reactor transient measurements with thermal feedback and corresponding coupled neutronic-thermohydraulic simulations performed for the natural circulation cooled Training Reactor of Budapest University of Technology and Economic ...

Fuel performance codes are essential for predicting fuel rod behavior in nuclear reactors. These codes comprise a complex system of models with empirical constants that require calibration against experimental data. To support this calibration, this study ...

This paper introduces a methodology for modeling Pressurized Water Reactors (PWRs) across multiple depletion cycles using the DRAGON and PARCS codes. The approach incorporates history effects to improve the accuracy of reactor simulations, specifically foc ...

Nodal diffusion codes have been successfully used for decades as a primary tool of commercial power reactor design, safety calculations and plant cycle simulations. The large-size, small-leakage property of these reactor cores and the appropriately generat ...

This study investigates the impact of 35Cl nuclear data uncertainties on the neutronics of Molten Chloride Fast Reactors (MCFR), specifically focusing on two models: MCFR-C and MCFR-D. Using the Monte Carlo code SERPENT2, a comprehensive sensitivity analys ...

Mathieu Hursin

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Measurement of the Prompt Neutron Decay Constant in a Zero-Power Reactor Using a Novel 3D Detector System

Monte Carlo analysis of BWR geometries using a cycle check-up methodology

Intra-pin Reaction Rate Measurements in the CROCUS Experimental Reactor

In-Memory, Fully-Parallel Coupling of GeN-Foam and OFFBEAT: Towards a New Milestone in High-Fidelity, Finite-Volume Nuclear Reactor Simulation

Sensitivity analysis in core diagnostics

Application of the PSI cycle check-up methodology for full-core Monte Carlo simulations verified using pin power estimates and validated for hot zero power conditions

Iterative Bayesian Monte Carlo for nuclear data evaluation

Modeling of Sodium Fast Reactors Thermal Expansion Reactivity Feedback using GeN-Foam and Serpent

Determination of the activity inventory and associated uncertainty quantification for the CROCUS zero power research reactor

Nuclear data uncertainties for Swiss BWR spent nuclear fuel characteristics

Development of a new method to determine the axial void velocity profile in BWRs from measurements of the in-core neutron noise

Design of an Instrumented Fuel Rod for Intra-pin Neutron Flux Measurements in the CROCUS Experimental Reactor

Thermal-hydraulic measurements and TRACE system code analysis performed on the natural circulation cooled BME Training Reactor

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

Modeling noise experiments performed at AKR-2 and CROCUS zero-power reactors

Analysis of transient measurements with thermal feedback and coupled TRACE/PARCS calculations performed on the BME Training Reactor

Preliminary Sensitivity Analysis of Time-Dependent Code Outputs in Fuel Performance Models Using OFFBEAT for Future Calibration

Modeling, verification and validation of multiple PWR depletion cycles with DRAGON and PARCS

Analysis of diffusion coefficient correction methods applied for small-core, high-leakage reactors

Uncertainty quantification of Cl-35 nuclear data on a Molten Chloride Fast Reactor