Publications de Dimitri Rochman

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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 ...

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 ...

Nuclear data, especially fission yields, create uncertainties in the predicted concentrations of fission products in spent fuel which can exceed engineering target accuracies. Herein, we present a new framework that extends data assimilation methods to bur ...

Developments in data assimilation theory allow to adjust integral parameters and cross sections with stochastic sampling. This work investigates how two stochastic methods, MOCABA and BMC, perform relative to a sensitivity-based methodology called GLLS. St ...

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 ...

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 ...

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 ...

The international experimental program PETALE will be carried out at the CROCUS research reactor of EPFL. The program aims at measuring neutron penetration in slabs made of materials composing typical LWR reactor pressure vessel. The measurements will be u ...

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 ...

A correlated sampling technique has been implemented to estimate the impact of cross section modifications on the neutron transport and in Monte Carlo simulations in one single calculation. This implementation has been coupled to a Total Monte Carlo approa ...

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 at EPFL intends to contribute to the validation and improvement of neutron nuclear data in the MeV energy range for stainless steel, particularly in the prospect of heavy reflector elements of pressur ...

This article presents the methodology developed to generate and use dosimeter covariances and to estimate nuisance parameters for the PETALE experimental programme. In anticipation of the final experimental results, this work investigates the consideration ...

The Bayesian Monte Carlo technics requires individual evaluations of random cross section files based on a Total Monte Carlo propagation. This article discusses the use of a Correlated Sampling acceleration applied to TMC calculations for experiments where ...

Dimitri Rochman

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

Nuclear data uncertainties for Swiss BWR spent nuclear fuel characteristics

Data assimilation of post-irradiation examination data for fission yields from GEF

Stochastic vs. sensitivity-based integral parameter and nuclear data adjustments

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

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

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

Assessment of representativity of the PETALE experiments for validation of Swiss LWRs ex-core dosimetry calculations

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

Uncertainty propagation based on correlated sampling technique for nuclear data applications

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

An Experimental Programme optimized with Uncertainty Propagation: PETALE in the CROCUS Reactor

Bayesian Monte Carlo assimilation for the PETALE experimental programme using inter-dosimeter correlation

Total Monte Carlo acceleration for the PETALE experimental programme in the CROCUS reactor

Loading optimization for Swiss used nuclear fuel assemblies into final disposal canisters (vol 370, 110897, 2020)