Desarrollo de la competencia transversal de “Análisis y Resolución de Problemas” en la asignatura Centrales Nucleares Avanzadas

[EN] The course "Advanced Nuclear Power Plants" is taught in the 4th year of the Grade of Energy Engineering, in the Universitat Politècnica de València. This is the first year this course have been taught and has a reduced number of students enrolled, as it is offered as an optional subject. All the students have a prior training in Nuclear Energy through the subject "Nuclear Technology". In this course, design and safety analysis of new nuclear fission reactors is studied. During this course it has begun work on the transversal skill "Problem Analysis and Resolution” in both theory and practice sessions. Aware of the importance of the development of this skill, different sessions are designed to solve a complex problem (not exercises) in which students should be involved in decision-making processes. For this purpose, the calculation of different design parameters of an advanced nuclear reactor using a Monte Carlo code is proposed. The problem to solve is open and not completely defined, and the class sessions are designed to provide guidance and help, but trying to encourage self-work and research. The evaluation was consistent with the methodology used and it has taken into account the scope of the analysis, giving priority to the metholology used even more than the results themselves.[ES] La asignatura “Centrales Nucleares Avanzadas” se imparte en 4º curso del Grado de Ingeniero de la Energía de la Universitat Politècnica de València. Se trata del primer año de docencia y actualmente cuenta con un número reducido de alumnos matriculados, con formación previa en Energía Nuclear gracias a la asignatura troncal “Tecnología Nuclear”. En la asignatura se aborda el diseño y el análisis de seguridad de reactores nucleares de fisión que se prevé que entren en funcionamiento comercial en un futuro próximo. Durante este curso se ha comenzado a trabajar en la competencia transversal de Análisis y Resolución de Problemas en las sesiones de prácticas y en algunas sesiones de teoría. Conscientes de la importancia del desarrollo de esta competencia, se han diseñado diferentes sesiones para resolver un problema (no ejercicio) complejo, en el que los alumnos deben involucrarse en la toma de decisiones. En este marco, se ha propuesto como problema el cálculo de una serie de parámetros de diseño de un reactor nuclear avanzado con un código de Monte Carlo. Se ha utilizado una sesión para resolver ejercicios como entrenamiento para enfrentarse al problema. Se ha propuesto un problema abierto y no completamente definido. Se han diseñado las sesiones proporcionando guías y ayuda, pero intentando fomentar el trabajo autónomo y la investigación. La evaluación ha sido coherente con la metodología empleada y se ha tenido muy en cuenta el alcance del análisis realizado, priorizándolo incluso más que los propios resultados.Gallardo Bermell, S.; Carlos Alberola, S. (2015). Desarrollo de la competencia transversal de “Análisis y Resolución de Problemas” en la asignatura Centrales Nucleares Avanzadas. En In-Red 2015 - CONGRESO NACIONAL DE INNOVACIÓN EDUCATIVA Y DE DOCENCIA EN RED. Editorial Universitat Politècnica de València. https://doi.org/10.4995/INRED2015.2015.1561OC

Simulation studies on natural circulation phenomena during an SBO accident

[EN] Natural circulation flow capability for removing decay core power has been demonstrated, and several studies have focused on taking advantage of this fact. This work studies the sequence of events that occur during a station blackout accident, in which natural circulation is the dominant flow pattern in the primary system. To this end, the Test A1.1 carried out in ATLAS facility is analyzed and a TRACE5 model is developed paying special attention on the modeling of heat losses. This phenomenon is very influential in the flow capacity and this is demonstrated through the correlation G~(Q-qloss )m between the net power Q-qloss and the mass flow G, that has been established from simulations under steady-state conditions. The Test A1.1 reproduction shows the TRACE5 code adequacy to investigate natural circulation phenomena, which are difficult to control in a facility. In addition, the heat loss modeling technique using constant heat transfer coefficients is substantiated.The authors are grateful to the Management Board of the OECD-ATLAS Project for their consent to this publication, and thank the Spanish Nuclear Regulatory Body (CSN) for the technical and financial support under the agreement STN/4524/2015/640 and the Spanish Ministerio de Economia, Industria y Competitividad under the agreement ENE2017-89029-P. They also thank Ronald Harrington, from USNRC, for sharing the preliminary TRACE model used in this work.Lorduy, M.; Gallardo Bermell, S.; Verdú Martín, GJ. (2018). Simulation studies on natural circulation phenomena during an SBO accident. Applied Thermal Engineering. 139:514-523. https://doi.org/10.1016/j.applthermaleng.2018.04.130S51452313

Análisis mediante el método de Monte Carlo de las dosis alrededor de la piscina de almacenamiento de las barras de control irradiadas en un reactor BWR

[ES] Las barras de control de un reactor de agua en ebullición (BWR) están sometidas a un flujo neutrónico y por tanto, resultan activadas durante su permanencia en el núcleo del reactor. La activación se produce especialmente en los componentes del acero inoxidable y en las impurezas. La actividad generada da lugar a una dosis alrededor de la barra, sin importancia mientras está en el reactor, pero que debe tenerse en cuenta cuando se extrae del mismo. Las barras extraídas se almacenan en colgadores situados en las piscinas de almacenamiento del combustible irradiado de la central. Cada colgador aloja 12 barras de control y se disponen de modo que haya al menos tres metros de agua por encima de los cabezales de las barras de control. La dosis potencialmente recibida por los trabajadores profesionalmente expuestos que se encuentren en las inmediaciones de la piscina de almacenamiento debe calcularse para asegurar la adecuada protección de los mismos. Esta dosis puede disminuirse de modo importante si se cambia la disposición de las barras en los colgadores.Ródenas Diago, J.; Gallardo Bermell, S. (2011). Análisis mediante el método de Monte Carlo de las dosis alrededor de la piscina de almacenamiento de las barras de control irradiadas en un reactor BWR. Revista de Física Médica. 12(Supl):508-508. http://hdl.handle.net/10251/99443S50850812Sup

Simulation of a SBLOCA in a hot leg. Scaling considerations andapplication to a nuclear power plant

The main goal of this work is to study the physical phenomena observed during a Small Break Loss-Of-Coolant Accident transient performed in a small-scale Integral Test Facility and to determine the capabilityof the thermal hydraulic code TRACE5 to reproduce these phenomena in a scale-up model. The accidentscenario analyzed is based on Test 1.2 in the frame of the OECD/NEA ROSA Project, which simulatesa 1% hot leg Small Break Loss-Of-Coolant Accident in the Large Scale Test Facility of the Japan AtomicEnergy Agency. During this test, natural circulation in primary loops occurs, cooling the core during someminutes. This is an important phenomenon, which needs to be checked by means of different TRACE5models. With this aim, Test 1 2 has been simulated using a TRACE5 model reproducing the geometricaland thermal hydraulic features of Large Scale Test Facility. In order to determine if this phenomenon canbe reliably extrapolated to a scale-up plant, a new TRACE5 model has been developed. The geometricalfeatures of this scale-up model are determined using a fixed scaling ratio respect to the original LSTFfeatures. On the other hand, 4 and 3-loop standard Westinghouse PWR models are used in order tosimulate the same transient and compare the behaviour of the main thermal hydraulic variables withthose obtained in the Large Scale Test Facility model and in the Large Scale Test Facility scale-up model.Results show that both Large Scale Test Facility and the scale-up models present the same behaviourduring the whole transient. Important discrepancies are found in the results corresponding to 4 and3-loop PWR TRACE5 models. In both models, natural circulation is not properly reproduced. Trying toimprove the simulation results, the nodalizations of U-tubes and pressure vessel were tested. Resultsstate that the nodalization of U-tubes clearly affects the natural circulation simulation. However, thevessel nodalization effect is not as important.This work contains findings produced within the OECD-NEA ROSA Project. This work is partially supported by the Grant-in-Aid for Scientific Research of the Spanish Ministerio de Educacion (Grant number: AP2009-2600), the Spanish Ministerio de Ciencia e Innovacion under Projects ENE2011-22823 and ENE2012-34585 and the Generalitat Valenciana under Projects PROMETEO/2010/039 and ACOMP/2013/237.Querol Vives, A.; Gallardo Bermell, S.; Verdú Martín, GJ. (2015). Simulation of a SBLOCA in a hot leg. Scaling considerations andapplication to a nuclear power plant. Nuclear Engineering and Design. 283:81-99. https://doi.org/10.1016/j.nucengdes.2014.10.006S819928

Unfolding X-ray spectra using a flat panel detector. Determination of the accuracy of the method with the Monte Carlo method

[EN] The primary X-ray spectrum depends on different parameters such as high voltage, filament current, high voltage ripple, anode angle and thickness of filter material. The objective of this work is to determine whether the unfolding technique based on the Tikhonov regularization method is accurate enough to estimate the X-ray spectrum when slight changes in the operation variables are considered. In this frame, several X-ray spectra are considered (extracted from the IPEM78 Catalogue Report) varying the main operation variables of the X-ray tube (high voltage, voltage ripple, filter thickness and filter material). With those spectra, the corresponding absorbed dose curves are obtained by simulation with a MCNP5 model reproducing a flat panel detector and a PMMA wedge. Once the absorbed dose curves are simulated and applying the unfolding Tikhonov regularization method, the unfolded spectrum is obtained, which is finally compared with the theoretical one (IPEM78 Catalogue Report). Discrepancies between unfolded and primary X-ray spectra can be attributed to the fact that this is an ill-posed problem, and the unfolding of the spectrum is strongly affected by the method used to improve the conditioning of the response function (response matrix).Gallardo Bermell, S.; Ródenas Diago, J.; Verdú Martín, GJ. (2019). Unfolding X-ray spectra using a flat panel detector. Determination of the accuracy of the method with the Monte Carlo method. Radiation Physics and Chemistry. 155:233-238. https://doi.org/10.1016/j.radphyschem.2018.09.014S23323815

Application of dosimetry measurements to analyze the neutron activation of a stainless steel sample in a training nuclear reactor

All materials present in the core of a nuclear reactor are activated by neutron irradiation. The activity so generated produces a dose around the material. This dose is a potential risk for workers in the surrounding area when materials are withdrawn from the reactor. Therefore, it is necessary to assess the activity generated and the dose produced. In previous works, neutron activation of control rods and doses around the storage pool where they are placed have been calculated for a Boiling Water Reactor using the MCNP5 code based on the Monte Carlo method. Most of the activation is produced indeed in stainless steel components of the nuclear reactor core not only control rods. In this work, a stainless steel sample is irradiated in the Training Reactor AKR-2 of the Technical University Dresden. Dose measurements around the sample have been performed for different times after the irradiation. Experimental dosimetric values are compared with results of Monte Carlo simulation of the irradiation. Comparison shows a good agreement. Hence, the activation Monte Carlo model can be considered as validated.Ródenas Diago, J.; Gallardo Bermell, S.; Weirich, F.; Hansen, W. (2014). Application of dosimetry measurements to analyze the neutron activation of a stainless steel sample in a training nuclear reactor. Radiation Physics and Chemistry. 104:368-371. doi:10.1016/j.radphyschem.2014.05.013S36837110

Break location influence in pressure vessel SBLOCA scenarios

[EN] The inspections performed in Davis Besse and the South Texas Project Unit-I reactors pointed out safety issues regarding the structural integrity of the Pressure Vessel (PV). In these inspections, two anomalies were found: a wall thinning and degradation in the PV upper head of the Davis Besse reactor and a small amount of residue around two instrument-tube penetration nozzles located in the PV lower plenum of the South Texas Project Unit-I reactor. The evolution of these defects could have resulted in Small Break Loss-Of-Coolant Accidents (SBLOCA) if they had not been detected in time. In this frame, the OECD/NEA considered the necessity to simulate these accidental sequences in the OECD/NEA ROSA Project using the Large Scale Test Facility (LSTF). This work is focused on simulating different hypothetical accidental scenarios in the PV using the thermalhydraulic code TRACE5. These simulations allow studying the break localization influence in the transient and the effectiveness of the accident management (AM) actions considered mitigating the consequences of these hypothetical accidental scenarios.This work contains findings that were produced within the OECD-NEA ROSA Project. The authors are grateful to the Management Board of the ROSA Project for their consent to this publication and thank the Spanish Nuclear Regulatory Body (CSN) for the technical and financial support under the agreement STN/1388/05/748Lorduy, M.; Querol, A.; Gallardo Bermell, S.; Verdú Martín, GJ. (2021). Break location influence in pressure vessel SBLOCA scenarios. Brazilian Journal of Radiation Sciences. 8(3B):1-17. http://hdl.handle.net/10251/182307S11783

Using lattice tools and unfolding methods for hpge detector efficiency simulation with the Monte Carlo code MCNP5

In environmental radioactivity measurements, High Purity Germanium (HPGe) detectors are commonly used due to their excellent resolution. Efficiency calibration of detectors is essential to determine activity of radionuclides. The Monte Carlo method has been proved to be a powerful tool to complement efficiency calculations. In aged detectors, efficiency is partially deteriorated due to the dead layer increasing and consequently, the active volume decreasing. The characterization of the radiation transport in the dead layer is essential for a realistic HPGe simulation. In this work, the MCNP5 code is used to calculate the detector efficiency. The F4MESH tally is used to determine the photon and electron fluence in the dead layer and the active volume. The energy deposited in the Ge has been analyzed using the *F8 tally. The F8 tally is used to obtain spectra and to calculate the detector efficiency. When the photon fluence and the energy deposition in the crystal are known, some unfolding methods can be used to estimate the activity of a given source. In this way, the efficiency is obtained and serves to verify the value obtained by other methods.Querol Vives, A.; Gallardo Bermell, S.; Ródenas Diago, J.; Verdú Martín, GJ. (2015). Using lattice tools and unfolding methods for hpge detector efficiency simulation with the Monte Carlo code MCNP5. Radiation Physics and Chemistry. 116:219-225. doi:10.1016/j.radphyschem.2015.01.027S21922511

Coincidence summing correction factors for 238U and 232Th decay series using the Monte Carlo method

[EN] Environmental samples analyzed in gamma spectrometry laboratories usually contain natural radionuclides such as 238U and 232Th. Using gamma spectrometry techniques is possible to estimate the activity of these radionuclides by measuring the gamma emissions of radionuclides belonging to their decay chain. Nonetheless, some of these radionuclides emit photons in cascade presenting Coincidence Summing (CS), which if not corrected, may affect the final activity quantification. The aim of this work is to apply the Monte Carlo method to calculate the True Summing Correction Factors (TSCFs) for 238U and 232Th decay series for different sample configurations (geometry and matrix) using the GEANT4 toolkit. In order to validate the results provided by GEANT4 using the RDM, the software TRUECOINC has been applied to calculate also the TSCFs. In addition, the influence of the geometry/matrix on the TSCFs is analyzed.The authors gratefully acknowledge financial support from the Catedra CSN-UPV Vicente Serradell, Spain as well as the Laboratorio de Radiactividad Ambiental (Universitat Politecnica de Valencia), Spain for the dedicated funding and resources to this research work under Grant no. FPI-2015-S2-1576Ordóñez-Ródenas, J.; Gallardo Bermell, S.; Ortiz Moragón, J.; Martorell Alsina, SS. (2019). Coincidence summing correction factors for 238U and 232Th decay series using the Monte Carlo method. Radiation Physics and Chemistry. 155:244-247. https://doi.org/10.1016/j.radphyschem.2018.09.013S24424715

Dosimetric characterization of a brachytherapy applicator using MCNP5 modelisation and in-phantom measurements

[EN] A gynaecological applicator consisting of a metallic intra-uterine tube with a plastic vaginal applicator and an HDR Ir-192 source have been simulated with MCNP5 (Monte Carlo code). A solid phantom has been designed to perform measurements around the applicator with radiochromic films. The isodose curves obtained are compared with curves calculated with the F4MESH tally of MCNP5 with a good agreement. A pinpoint ionization chamber has been used to evaluate dose at some reference points. (c) 2009 Elsevier Ltd. All rights reserved.Gerardy, I.; Ródenas Diago, J.; Van Dycke, M.; Gallardo Bermell, S.; Ceccolini, E. (2010). Dosimetric characterization of a brachytherapy applicator using MCNP5 modelisation and in-phantom measurements. Applied Radiation and Isotopes. 68(4):735-737. doi:10.1016/j.apradiso.2009.10.018S73573768