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## Publications

Publications (283)

Decay heat calculations of spent nuclear fuel (SNF) using Polaris and ORIGEN codes in the SCALE code system, and CASMO5 code, are validated using measurements from the Clab and GE-Morris facilities. Multiple hypothesis testing, relying on permutations and bootstrapping, is conducted to simultaneously analyze the significance of differences between...

Following the EUROfusion PPPT-programme action for an advanced modeling approach of deuteron-induced reaction cross sections, as well as specific data evaluations in addition of the TENDL files, an assessment of the details and corresponding outcome for the latter option of TALYS for the breakup model has been carried out. The breakup enhancement o...

In this paper, the impact of the thermal scattering data for H in H 2 0 is estimated on criticality benchmarks, based on the variations of the CAB model parameters. The Total Monte Carlo method for uncertainty propagation is applied for 63 k eff criticality cases, sensitive to H in H 2 0. It is found that their impact is of a few tenth of pcm, up t...

Most elements heavier than iron have been generated in the stellar media by means of neutron capture reactions, approximately half are produced by the slow neutron capture or s -process. Radiative neutron capture cross section measurements are of fundamental importance for the study of this mechanism. In this contribution we present a brief summary...

This paper presents an approach for the optimisation of geological disposal canister loadings, combining high resolution simulations of used nuclear fuel characteristics with an articial neural network and a genetic algorithm. The used nuclear fuels (produced in an open fuel cycle without reprocessing) considered in this work come from a Swiss Pres...

A stochastic technique in view of determining neutronic parameters is used along with the JEFF-3.3 nuclear data library, trying to adjust in conjunction with the Asymptotic Generalized Linear Least-Squares (AGLLS) assimilation methodology, a demonstrative limited number of basic ENDF parameters for a few nuclides, which are U-235, U-238 and hydroge...

There is a long-standing controversy on nuclear data uncertainty assessment for general purpose nuclear data libraries. On the one hand, nuclear data users would like the libraries to predict uncertainties for selected integral quantities consistent with the integral experimental uncertainties, while on the other hand, doing so could make evaluatio...

The aim of this paper is to assess the reliability and accuracy of the PSI standard method, used in many previous works, for the quantification of ND uncertainties in the SPERT-III RIA transient, by quantifying the discrepancy between the actual inserted reactivity and the original static reactivity worth and their associated uncertainties. The ass...

The present study provides a detailed analysis of the calculations of 50 isotopic compositions for the Post Irradiation Examination sample GU1 from the ARIANE program. Two types of approach are performed: a lattice (2 dimensional) and a full core (3 dimensional) calculations. In the case of lattice calculations, the effect oflocation is also studie...

The current research summarizes recent developments performed within the PSI BEPU methodology and the corresponding results on the SPERT-III RIA experiments. In this regard, three main studies have been carried out. First, an uncertainty breakdown study has been performed in order to identify the dominant nuclear data and to quantify their respecti...

Presently, a criticality safety evaluation methodology for the final geological disposal of Swiss spent nuclear fuel is under development at the Paul Scherrer Institute in collaboration with the Swiss National Technical Competence Centre in the field of deep geological disposal of radioactive waste. This method in essence pursues a best estimate pl...

An accurate measurement of the 140Ce(n,γ) energy-dependent cross-section was performed at the n_TOF facility at CERN. This cross-section is of great importance because it represents a bottleneck for the s-process nucleosynthesis and determines to a large extent the cerium abundance in stars. The measurement was motivated by the significant differen...

i-TED is an innovative detection system which exploits Compton imaging techniques to achieve a superior signal-to-background ratio in (\(n,\gamma \)) cross-section measurements using time-of-flight technique. This work presents the first experimental validation of the i-TED apparatus for high-resolution time-of-flight experiments and demonstrates f...

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 decay heat residuals to their uncertainties, focusing on four PWRs and four BWRs. Uncertainties i...

We discuss the design and software implementation of a nuclear data evaluation pipeline applied for a fully reproducible evaluation of neutron-induced cross sections of ⁵⁶Fe above the resolved resonance region using the nuclear model code TALYS combined with relevant experimental data. The emphasis of this paper is on the mathematical and technical...

Assessing neutron emission of LWR spent fuel is necessary for the back-end of the fuel cycle, such as the dimensioning of transport and storage casks of spent fuel. Although core and depletion codes can calculate the isotopic composition of the discharged fuel and therefore infer its neutron source, accurate measured neutron emission values remain...

The CASMO/SIMULATE/MCNP/FISPACT-II calculation route has been established at the Paul Scherrer Institute (PSI) for reactor dosimetry and activation studies. Furthermore, the in-house tool NUSS is in use at PSI for nuclear data (ND) related uncertainties quantifications with Monte Carlo neutron transport calculations. The use of randomly sampled ACE...

This paper describes the effect of input uncertainties on a set of integral parameters (kinf, nuclide compositions) associated with the validation of CASMO-5 against PIE data. The nuclear data under consideration are the cross-sections, fission spectrum and neutron multiplicities and fission yields. Various sources of covariance information are con...

The main goal of this work is to perform pin-by-pin calculations of Swiss LWR fuel assemblies with neutron transport deterministic methods. At Paul Scherrer Institut (PSI), LWR calculations are performed with the core management system CMSYS, which is based on the Studsvik suite of codes. CMSYS includes models for all the Swiss reactors validated a...

Cross sections and fission yields can be correlated, depending on the selection of integral experimental data. To support this statement, this work presents the use of experimental isotopic compositions (both for actinides and fission products) from a sample irradiated in a reactor, to construct correlations between various cross sections and fissi...

This study presents an analysis of the ARIANE GU3 sample, in terms of nuclide inventory, as well as sample rod and assembly decay heat. The validation of a number of CASMO5 and library versions are performed with regards to the measured nuclide inventory, taking into account two dimensional lattice simulations. Uncertainties due to various sources...

The Mixed Oxide samples (MOX) ARIANE Post Irradiation Examination samples BM1 and BM3 have been analyzed in this work, based on various two- and three-dimensional models. Calculated and measured nuclide inventories are compared based on CASMO5, SIMULATE and SNF simulations, and calculated values for the decay heat of the assembly containing the sam...

The impact of the H in H 2 O thermal scattering data are calculated for burnup quantities, considering models of a UO 2 pincell with DRAGON and SERPENT. The Total Monte Carlo method is applied, where the CAB model parameters are randomly varied to produce sampled (random) LEAPR input files for NJOY. A large number of burnup calculations is then per...

i-TED is an innovative detection system which exploits Compton imaging techniques to achieve a superior signal-to-background ratio in ($n,\gamma$) cross-section measurements using time-of-flight technique. This work presents the first experimental validation of the i-TED apparatus for high-resolution time-of-flight experiments and demonstrates for...

In this paper the isotopic content of the U1 sample is re-analyzed taking into account the full irradiation environment with a 3-dimensional core simulator. It is found that by accounting for the neighboring assemblies as well as the various axial segments, the agreement between the measured and calculated isotopic concentrations is greatly improve...

In this paper, the possibilities and advantages of loading mixed used nuclear fuel (UNF) assemblies into canisters for final repository are presented from a criticality aspect within the burnup credit approach (BUC). UNF coming from a Swiss reactor are taken into account at the pin-by-pin level and two canister loading patterns are studied: mixed a...

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 on the number of used fuels to be disposed, this study indicates that a canister filling fracti...

The activities of the EUROfusion consortiums on the development of high quality nuclear data for fusion applications are resented.The activities, implemented in the Power Plant Physics and Technology (PPPT) programme of EUROfusion, include nuclear data evaluations for neutron and deuteron induced reactions and thep roduction of related data librari...

The neutron capture cross sections of several unstable nuclides acting as branching points in the s-process are crucial for stellar nucleosynthesis studies. The unstable 171Tm (t1/2=1.92yr) is part of the branching around mass A∼170 but its neutron capture cross section as a function of the neutron energy is not known to date. In this work, followi...

Neutron capture cross sections are one of the fundamental nuclear data in the study of the s (slow) process of nucleosynthesis. More interestingly, the competition between the capture and the decay rates in some unstable nuclei determines the local isotopic abundance pattern. Since decay rates are often sensible to temperature and electron density,...

Radiative neutron capture cross section measurements are of fundamental importance for the study of the slow neutron capture (s-) process of nucleosynthesis. This mechanism is responsible for the formation of most elements heavier than iron in the Universe. Particularly relevant are branching nuclei along the s-process path, which are sensitive to...

The idea of slow-neutron capture nucleosynthesis formulated in 1957 triggered a tremendous experimental effort in different laboratories worldwide to measure the relevant nuclear physics input quantities, namely (n, γ) cross sections over the stellar temperature range (from few eV up to several hundred keV) for most of the isotopes involved from Fe...

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 used for code and nuclear data validation and for the evaluation of the reflecting properties o...

In this work, an overview on the relevance of the nuclear data (ND) uncertainties with respect to the Light Water Reactors (LWR) neutron dosimetry is presented. The paper summarizes results of several studies realized at the LRT laboratory of the Paul Scherrer Institute over the past decade. The studies were done using the base LRT calculation meth...

New measurements of the ⁷ Be(n, α ) ⁴ He and ⁷ Be(n,p) ⁷ Li reaction cross sections from thermal to keV neutron energies have been recently performed at CERN/n_TOF. Based on the new experimental results, astrophysical reaction rates have been derived for both reactions, including a proper evaluation of their uncertainties in the thermal energy rang...

The (n, γ ) cross sections of the gadolinium isotopes play an important role in the study of the stellar nucleosynthesis. In particular, among the isotopes heavier than Fe, ¹⁵⁴ Gd together with ¹⁵² Gd have the peculiarity to be mainly produced by the slow capture process, the so-called s-process, since they are shielded against the β -decay chains...

Accurate neutron capture cross section data for minor actinides (MAs) are required to estimate the production and transmutation rates of MAs in light water reactors with a high burnup, critical fast reactors like Gen-IV systems and other innovative reactor systems such as accelerator driven systems (ADS). Capture reactions of ²⁴⁴ Cm open the path f...

The neutron induced fission of ²³⁵ U is extensively used as a reference for neutron fluence measurements in various applications, ranging from the investigation of the biological effectiveness of high energy neutrons, to the measurement of high energy neutron cross sections of relevance for accelerator driven nuclear systems. Despite its widespread...

Neutron capture on ²⁴¹ Am plays an important role in the nuclear energy production and also provides valuable information for the improvement of nuclear models and the statistical interpretation of the nuclear properties. A new experiment to measure the ²⁴¹ Am(n, γ ) cross section in the thermal region and the first few resonances below 10 eV has b...

The study of neutron-induced reactions on actinides is of considerable importance for the design of advanced nuclear systems and alternative fuel cycles. Specifically, ²³⁰ Th is produced from the α-decay of ²³⁴ U as a byproduct of the ²³² Th/ ²³³ U fuel cycle, thus the accurate knowledge of its fission cross section is strongly required. However, f...

The neutron-induced fission cross section of ²³⁵ U, a standard at thermal energy and between 0.15 MeV and 200 MeV, plays a crucial role in nuclear technology applications. The long-standing need of improving cross section data above 20 MeV and the lack of experimental data above 200 MeV motivated a new experimental campaign at the n_TOF facility at...

Since the start of its operation in 2001, based on an idea of Prof. Carlo Rubbia [1], the neutron time of-flight facility of CERN, n_TOF, has become one of the most forefront neutron facilities in the world for wide-energy spectrum neutron cross section measurements. Thanks to the combination of excellent neutron energy resolution and high instanta...

Feasibility, design and sensitivity studies on innovative nuclear reactors that could address the issue of nuclear waste transmutation using fuels enriched in minor actinides, require high accuracy cross section data for a variety of neutron-induced reactions from thermal energies to several tens of MeV. The isotope ²⁴¹ Am (T 1/2 = 433 years) is pr...

Although the ¹² C(n,p) ¹² B and ¹² C(n,d) ¹¹ B reactions are of interest in several fields of basic and applied Nuclear Physics the present knowledge of these two cross-sections is far from being accurate and reliable, with both evaluations and data showing sizable discrepancies. As part of the challenging n_TOF program on (n,cp) nuclear reactions...

We have measured the γ -rays following neutron capture on ²⁴⁰ Pu and ²⁴⁴ Cm at the n_TOF facility at CERN with the Total Absorption Calorimeter (TAC) and with C6D6 organic scintillators. The TAC is made of 40 BaF2 crystals operating in coincidence and covering almost the entire solid angle. This allows to obtain information concerning the energy sp...

The main goal of this work is to perform pin-by-pin calculations of Swiss LWR fuel assemblies with neutron transport deterministic methods. At Paul Scherrer Institut (PSI), LWR calculations are performed with the core management system CMSYS, which is based on the Studsvik suite of codes. CMSYS includes models for all the Swiss reactors validated a...

In order to reduce methodological effects, a stochastic technique is proposed for adjusting basic ENDF parameters instead of preprocessed multi-group data as it is normally the case; the required sensitivity coefficients of the integral parameters to these parameters are recomputed by means of NJOY/Serpent within an iterative scheme based upon the...

We discuss the design and software implementation of a nuclear data evaluation pipeline applied for a fully reproducible evaluation of neutron-induced cross sections of 56Fe above the resolved resonance region using the nuclear model code TALYS combined with relevant experimental data. The emphasis is on the mathematical and technical aspects of th...

The effects of the nuclear data uncertainties are quantified for specific core parameters from a real Boiling Water Reactor operated over 25 consecutive cycles. Nominal calculations are performed with the CASMO5 and SIMULATE-3 codes and the ENDF/B-VII.0 nuclear data libraries. The uncertainties are calculated using a Monte Carlo propagation method,...

The joint evaluated fission and fusion nuclear data library 3.3 is described. New evaluations for neutron-induced interactions with the major actinides \(^{235}\hbox {U}\), \(^{238}\hbox {U}\) and \(^{239}\hbox {Pu}\), on \(^{241}\hbox {Am}\) and \(^{23}\hbox {Na}\), \(^{59}\hbox {Ni}\), Cr, Cu, Zr, Cd, Hf, W, Au, Pb and Bi are presented. It includ...

The CASMO/SIMULATE/MCNP/FISPACT-II calculation route has been established at the Paul Scherrer Institute (PSI) for reactor dosimetry and activation studies. Furthermore, the in-house tool NUSS is in use at PSI for nuclear data (ND) related uncertainties quantifications with Monte Carlo neutron transport calculations. The use of randomly sampled ACE...

In this work, a method is proposed for combining differential and integral benchmark experimental data within a Bayesian framework for nuclear data adjustments and multi-level uncertainty propagation, using the Total Monte Carlo method. First, input parameters to basic nuclear physics models implemented within the TALYS code, were randomly varied t...

The neutron capture cross section of ¹⁵⁴Gd was measured from 1 eV to 300 keV in the experimental area located 185 m from the CERN n_TOF neutron spallation source, using a metallic sample of gadolinium, enriched to 67% in ¹⁵⁴Gd. The capture measurement, performed with four C6D6 scintillation detectors, has been complemented by a transmission measure...

In this work, we explore the use of an iterative Bayesian Monte Carlo (IBM) procedure for nuclear data evaluation within a Talys Evaluated Nuclear data Library (TENDL) framework. In order to identify the model and parameter combinations that reproduce selected experimental data, different physical models implemented within the TALYS code, were samp...

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 and SNF codes and the ENDF/B-VII.0 nuclear data library. The uncertainties are calculated with...

In this paper, a statistical analysis of resonance evaluations from the most recent nuclear data libraries is performed with the code TARES. A description of the TARES framework is provided, but also for its use in the production of resonance parameters for the entire TENDL library. Various observables are calculated (e.g., thermal cross sections,...