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In the field of reactor physics analysis, the coupling of multiple physics phenomena plays a crucial role in enhancing the accuracy of predictions and understanding complex systems. Multiphysics refers to the study and analysis of physical phenomena that involve multiple interconnected physical processes occurring simulta- neously in a nuclear system to ensure essential aspects of reactor design, safety analysis, and optimization, among others. Multiphysics encompasses various domains of physics, such as Thermal-Hydraulics, Neutronics, Structural Mechanics, Radiation Transport, Chemistry, and Materials Science. On the other hand, the Multiscale approach involves combining high-fidelity models with lower-fidelity ones to accurately capture the relevant physical phenomena in a Nuclear Power Plant (NPP). Multiscale involves three main scales: system level, core level, and detailed analysis scale. This paper presents an overview of the main research performed by Spanish groups in the field of multiphysics and multiscale calculations, particularly on thermal–hydraulic analysis. This revision is focused mainly on two key aspects: thermal–hydraulic multiscale coupling and thermal–hydraulic / neutronic coupling. The paper also includes the expected future trends in this field., The authors are grateful to the Spanish Nuclear Safety Council (CSN) for the support in many of the activities presented in this article., Article signat per 16 autors/es: Sergio Gallardo, Francisco Álvarez-Velarde, Teresa Barrachina, Óscar Cabellos, Emilio Castro, Max Casamor, Diana Cuervo, Alberto Escrivá, Jordi Freixa, Nuria García-Herranz, Victor Martinez-Quiroga, Rafael Miró, César Queral, Yago Rivera, Jorge Sánchez-Torrijos, Amparo Soler., Postprint (published version)

Entre los reactores avanzados identificados como más prometedores se encuentran los reactores rápidos refrigerados por sodio. Aunque se dispone de cierta experiencia operacional, aún quedan retos por superar antes de su implantación industrial. Idealmente, se debería disponer de instalaciones experimentales que puedan apoyar su diseño y seguridad, pero teniendo en cuenta su coste económico, estas actividades se basan cada vez más en simulaciones computacionales. Por ello, la simulación precisa de estos reactores es crítica para su diseño, optimización y evaluación de los márgenes de seguridad. Las simulaciones neutrónicas dependen de los datos nucleares, almacenados en las librerías evaluadas. Actualmente, para algunos isótopos y algunos rangos de energía, los datos disponibles difieren entre sí, y respecto a los valores experimentales. Esas diferencias hacen que las simulaciones computacionales aporten resultados diferentes en función de la librería utilizada. El objetivo de este Trabajo Fin de Grado es estudiar qué datos nucleares son los más adecuados para la simulación de reactores rápidos de sodio, comparando los resultados con los valores experimentales disponibles para el reactor japonés JOYO. Para ello, tras la correspondiente contextualización de la situación actual de los reactores rápidos de sodio, y un estudio del reactor JOYO, se ha realizado un modelo del núcleo de este reactor para el código de transporte neutrónico de Monte Carlo MCNP. Se han comparado las predicciones con distintas librerías, en particular la librería europea JEFF, con las medidas experimentales para detectar la librería idónea. Un análisis de sensibilidad y perturbación ha permitido concluir qué secciones eficaces son responsables de las desviaciones entre librerías y cuya revisión contribuiría a mejorar la librería JEFF.

The ESFR-SMART project is the latest iteration of research into the behaviour of a commercial-size SFR core throughout its lifetime. As part of this project the ESFR core has been modelled by a range of different reactor physics simulation codes at its end of cycle state, and the important safety relevant parameters evaluated. These parameters are found to agree well between the different codes, giving good confidence in the results. A detailed mapping of the local sodium void worth is also performed due to the problems associated with the positive void coefficient seen in large SFR designs. The local void worth maps show that the use of zone-wise coefficients replicates the important reactivity feedbacks to a high degree, indicating their suitability for use in SFR simulations.

Monte Carlo neutron transport codes can be used for high-fidelity predictions of the performance of nuclear systems. However, validation against experiments is required in order to establish the credibility in the results and identify the inaccuracies due to the used calculation scheme and associated databases. The International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP) contains criticality safety benchmarks derived from experiments that have been performed at various nuclear critical facilities around the world and are very valuable for validation purposes. The main objective of this work is the identification and modelling of experimental benchmarks included at ICSBEP in support of the validation of Monte Carlo neutron transport calculations when applied to fast systems, and in particular, KENO-VI and associated AMPX-formatted continuous-energy libraries from SCALE package. In such systems, the predicted k-eff values can be very sensitive to the treatment of nuclear data in the Unresolved Resonance Region (URR). Consequently, benchmarks with intermediate and fast spectra are identified and modelled with KENO-VI. Then, calculated results with and without probability tables in the URR are compared with each other in order to identify the most sensitive configurations to the URR. As a result of the proposed study, recommendations are given about the benchmarks that should be modelled and analysed to qualify the processed continuous-energy libraries before their use in Monte Carlo transport codes for practical fast reactor applications.

Dentro de los reactores nucleares de la Generación IV, los reactores rápidos de sodio (SFR) están considerados como una de las opciones más prometedoras en el futuro de la energía nuclear dadas sus características técnicas y la experiencia operativa que se tiene de ellos. En el ámbito europeo, el proyecto ESFR-SMART pretende demostrar una mayor viabilidad y fiabilidad de los reactores rápidos de sodio con el diseño del reactor ESFR (European Sodium Fast Reactor). El objetivo principal de dicho proyecto es mejorar las medidas de seguridad del reactor, así como llevar a cabo la evaluación de estas. Dentro del marco del proyecto ESFR-SMART, el presente Trabajo de Fin de Grado tiene como objetivo la evaluación del funcionamiento del reactor ESFR al utilizar un combustible nuclear alternativo basado en torio. Para ello, se llevarán a cabo las simulaciones de los ciclos del combustible utilizando el código SCALE, que acopla los cálculos de transporte neutrónico y quemado del combustible, realizados por los módulos KENO-VI y ORIGEN respectivamente. ABSTRACT Within the IV Generation nuclear reactors, sodium fast reactors (SFR) are considered to be one of the most promising options in the future of nuclear energy given their technical characteristics and operational experience. At the European level, the ESFR-SMART project aims to demonstrate the increased feasibility and reliability of sodium fast reactors with the ESFR (European Sodium Fast Reactor) design. The main goal of this project is to improve the safety measures of the reactor, as well as to carry out the evaluation of these measures. Within the framework of the ESFR-SMART project, the scope of the present dissertation is to evaluate the operation of the ESFR reactor when using an alternative thorium-based nuclear fuel. To this end, fuel cycles simulations will be carried out using the SCALE code, which couples the neutron transport and fuel burnout calculations, performed by the KENO-VI and ORIGEN’s modules respec

The safety level of Sodium Fast Reactors is directly related with the sodium void reactivity. A low-void effect design has been proposed within the Horizon2020 ESFR-SMART project thanks to the introduction of a sodium plenum above the active core. In order to assess the impact of this core conception on transient analysis, a map with the spatial distribution of sodium void worth can be computed and fed into a point-kinetics-based transient code. Due to the spatial correlations between neighboring zones, the global effect of voiding two different axial or radial regions is not necessarily the sum of both individual contributions. Neglecting those correlations in the void worth map and consequently in the transient analysis may lead to an unrealistic prediction of the transient sequences. In this work, a method based on sensitivity analysis and similarity assessment is proposed for predicting those correlations. The method proved to be able to establish correlations between axial slices of a sub-assembly and was checked against realistic sodium void propagation patterns.

La irradiación de alimentos como método de conservación consiste en la aplicación de radiación ionizante a alimentos (humanos o animales) para lograr una serie de objetivos como alargar su tiempo de vida, evitar enfermedades por la presencia de patógenos en ellos, retrasar la germinación de brotes o esterilizar los alimentos. En función de la dosis aplicada se consiguen unos u otros objetivos: las dosis más bajas (menores de 1 kGy) se emplean para el retraso en la germinación, y las más altas (hasta 45 kGy) para esterilizar completamente. Sin embargo, esta técnica también presenta algunos riesgos, como la posible pérdida de nutrientes o la modificación de algunas propiedades organolépticas. Es muy importante ser consciente de que irradiar un alimento con las energías empleadas para ello no lo hace radiactivo. La radiación empleada en esta técnica puede ser de 3 tipos: rayos gamma () procedentes de fuentes radiactivas encapsuladas (Co60 o Cs137); rayos X generados a partir de un acelerador de electrones con energía igual o inferior a 5 MeV; o electrones generados por aparatos que los aceleran hasta una energía igual o inferior a 10 MeV. La energía de los rayos de las fuentes radiactivas empleadas, así como de los rayos X generados, es lo suficientemente baja como para no activar los alimentos, es decir, no los hace radiactivos. Existen multitud de métodos para detectar si un alimento ha sido irradiado o no; el objetivo de este Trabajo es profundizar en el método de termoluminiscencia (TL). Cuando un alimento es irradiado con radiación ionizante, algunos de los minerales presentes en ellos, fundamentalmente los silicatos, absorben dicha radiación debido a su estructura de trampas, que básicamente son niveles discretos de energía que existen en la región entre la banda de valencia y la de conducción. Al irradiar el material, se transfiere energía a los electrones de la banda de valencia. Algunos de estos electrones llegan a la banda de conducción y se recombinan de for

Within the NURESAFE project, a main steam line break benchmark has been defined and solved by codes integrated into the European code platform NURESIM. The paper describes the results of the calculations for this benchmark. Six different solutions using different codes and code systems are provided for the comparison. The quantitative differences in the results are dominated by the differences in the secondary system parameters during the depressurization. The source of these differences comes mainly from the application of different models for the two-phase leak flow available in the system codes. The use of two different thermal hydraulic system codes influences the results more than expected when the benchmark was created. The codes integrated into the NURESIM platform showed their applicability to a challenging transient like a main steam line break.

A sensitivity and uncertainty analysis was carried out to estimate the uncertainty in the neutron multiplication factor keff and to identify the most important nuclear data for neutron induced reactions for criticality calculations of the latest MYRRHA designs. Sensitivity profiles, i.e. sensitivity to the nuclear data as a function of incoming neutron energy, were derived for both a critical and sub-critical core. They were calculated using codes that are based on different methodologies including stochastic and deterministic calculations (i.e. SCALE, MCNP and XSUN). The neutron induced nuclear data sensitivity analysis outlined the following quantities to be of special importance for the MYRRHA reactor concept: 239Pu(n,c) both in resonance and fast energy region, (n,f) fast, v and m fast; 238U(n,n0 ) fast, (n,c) resonance and fast, (n,n) resonance and fast; 240Pu m fast; 238Pu(n,f) both resonance and fast; 56Fe(n,c) both resonance and fast. Differences of less than 4% between codes were obtained for these quantities, with few exceptions ( 238Pu(n,f), 238U(n,n) and 56Fe(n,c) reactions). Nuclear data covariance matrices of different libraries (SCALE-6, COMMARA-2 and JENDL-4.0m) were used to derive the uncertainty in keff based on the calculated sensitivities. This study reveals that the largest contributions to keff uncertainty result from the uncertainty in the average prompt neutron fission multiplicity of 239Pu, in the 238U inelastic scattering cross section and 239Pu fission cross section, using the covariances from SCALE-6, COMMARA-2 and JENDL-4.0m, respectively.

The MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) concept is a flexible experimental lead-bismuth cooled and mixed-oxide (MOX) fueled fast spectrum facility designed to operate both in sub-critical (accelerator driven) and critical modes. One of the key issues for the safe operation of the reactor is the uncertainty assessment during the design works. The main objective of the European project CHANDA (solving CHAllenges in Nuclear DAta) Work Package 10 is to improve MYRRHA relevant nuclear data in order to reduce the reactor parameter uncertainties derived from them. In order to achieve this goal, several tasks have been undertaken. First, a sensitivity study of MYRRHA integral parameters, such as energy dependent cross sections, fission spectra and neutron multiplicities, to nuclear data has been conducted resulting in a list of MYRRHA relevant quantities (nuclides and reactions). On the second task, an analysis of the existing experimental data and evaluations for the quantities included in the list has been carried out. In this framework, the impact on the multiplication factor of quantities from different nuclear data libraries for different nuclides, reactions and energy regions has been investigated on the MYRRHA MOX critical core model. As the next step, new experiments and evaluations will be performed in order to improve existing nuclear data libraries.

The impact of the current nuclear data library covariances such as in ENDF/B-VII.1, JEFF-3.2, JENDL-4.0, SCALE and TENDL, for relevant current reactors is presented in this work. The uncertainties due to nuclear data are calculated for existing PWR and BWR fuel assemblies (with burn-up up to 40 GWd/tHM, followed by 10 years of cooling time) and for a simplified PWR full core model (without burn-up) for quantities such as k∞ , macroscopic cross sections, pin power or isotope inventory. In this work, the method of propagation of uncertainties is based on random sampling of nuclear data, either from covariance files or directly from basic parameters. Additionally, possible biases on calculated quantities are investigated such as the self-shielding treatment. Different calculation schemes are used, based on CASMO, SCALE, DRAGON, MCNP or FISPACT-II, thus simulating real-life assignments for technical-support organizations. The outcome of such a study is a comparison of uncertainties with two consequences. One: although this study is not expected to lead to similar results between the involved calculation schemes, it provides an insight on what can happen when calculating uncertainties and allows to give some perspectives on the range of validity on these uncertainties. Two: it allows to dress a picture of the state of the knowledge as of today, using existing nuclear data library covariances and current methods.

This paper summarizes the nodal level results from the VVER MSLB core simulation in the NURESAFE EU project. The main objective is to implement and verify new developments in the models and couplings of 3D core simulators for cores with hexagonal fuel assemblies. Recent versions of the COBAYA and DYN3D core physics codes, and the FLICA4 and CTF thermal-hydraulic codes were tested standalone and coupled through standardized coupling functions in the Salome platform. The MSLB core transient was analyzed in coupled code simulation of a core boundary condition problem derived from the OECD VVER MSLB benchmark. The impact of node sub-division and different core mixing models, as well as the effects of CFD computed core inlet thermal-hydraulic boundary conditions on the core dynamics were explored.

At UPM, in-depth modifications of the core simulator COBAYA, able to perform neutronics diffusion calculations at both nodal and pin-by-pin levels, have been accomplished during the 7th Framework EURATOM NURESAFE project. The main goal was to upgrade its integration in the European Platform for Nuclear Reactor Safety Simulation in order to facilitate the coupling with any other code of the platform for multi-physics analysis, focusing also on the code legibility and maintainability. An external and flexible coupling with the thermal-hydraulics code COBRA-TF was designed. As a result, COBAYA4/COBRA-TF allows multiscale coupled calculations, enabling both nodal and pin-by-pin neutronics resolutions using both assembly-based channels and pin-based subchannels at the thermal-hydraulics domain. Flexible mapping schemes also in axial direction can be defined. The coupled system was applied to a Main Steam Line Break transient benchmark. Pin-by-pin 3D simulations using one thermal-hydraulic channel per assembly were carried out in a reasonable computing time, and results compared to nodal solutions demonstrating the multiscale coupling capability for full core transients. Pin-by-pin calculations using thermal-hydraulics subchannels will be performed in a near future to assess the role that a very detailed mapping can play to predict realistic local parameters. While in asymmetric transients the effect can be important, it is expected that in symmetric transients assembly-based thermal-hydraulics channels can provide accurate pin-by-pin solutions in execution times suitable for routine analysis. The performed work will bring the ability to explore in an easy way multiscale effects on safety transient evaluations and give recommendations for the neutronics/thermal-hydraulics mapping depending on the application.

Diseño de un reactor, en este caso el prototipo de reactor rápido refrigerado por plomo ALFRED, requiere llevar a cabo simulaciones computacionales de múltiples escenarios para evaluar su seguridad. El punto de partida de dichas simulaciones son los datos nucleares. Estos datos tienen incertidumbres, por lo que resulta fundamental evaluar el impacto de dichas incertidumbres sobre parámetros característicos del reactor como su reactividad. En este Trabajo Fin de Grado se ha llevado a cabo un estudio del impacto de las incertidumbres de los datos nucleares en la reactividad (k-eff) del reactor ALFRED, y se valorado el efecto de utilizar distintas librerías de datos. Para ello se ha utilizado la herramienta computacional NDaST (Nuclear Data Sensitivity Tool), desarrollada por la OECD NEA (Nuclear Energy Agency), y se ha elaborado un manual de usuario. The design of a reactor, in this case the prototype ALFRED lead-cooled fast reactor, requires carrying out computational simulations of various scenarios to evaluate its safety. The nuclear data are the starting point of these simulations. These data have uncertainties, so it is essential to evaluate the impact of such uncertainties on representative parameters of the reactor as its reactivity. In this Project, a study of the impact of nuclear data uncertainties on the reactivity (k-eff) of the ALFRED reactor is carried out, and the effect of using different data libraries is evaluated. For this purpose, the NDaST (Nuclear Data Sensitivity Tool), developed by the OECD NEA (Nuclear Energy Agency), has been used and a user manual has been developed.

The Monte Carlo-based Bayesian inference model MOCABA is applied to the prediction of reactor operation parameters of a PWR nuclear power plant. In this non-perturbative framework, high-dimensional covariance information describing the uncertainty of microscopic nuclear data is combined with measured reactor operation data in order to provide statistically sound, well founded uncertainty estimates of integral parameters, such as the boron letdown curve and the burnup-dependent reactor power distribution. The performance of this methodology is assessed in a blind test approach, where we use measurements of a given reactor cycle to improve the prediction of the subsequent cycle. As it turns out, the resulting improvement of the prediction quality is impressive. In particular, the prediction uncertainty of the boron letdown curve, which is of utmost importance for the planning of the reactor cycle length, can be reduced by one order of magnitude by including the boron concentration measurement information of the previous cycle in the analysis. Additionally, we present first results of non-perturbative nuclear-data updating and show that predictions obtained with the updated libraries are consistent with those induced by Bayesian inference applied directly to the integral observables.