Gonzalo Jimenez

• PhD
• Professor (Associate) at Universidad Politécnica de Madrid

Los códigos Computational Fluid Dynamics (CFD) son una herramienta ampliamente empleada para analizar sistemas de almacenamiento en seco de combustible nuclear gastado. Una simplificación típicamente usada en muchos de los modelos de combustible nuclear gastado disponible en el estado del arte es no modelar (o modelar de forma no explicita) los cab...

This document explores the role of CFD in nuclear safety studies, with a particular focus on advancements in Spain and Portugal. The field of nuclear safety has increasingly included CFD models to address complex safety–critical phenomena, given their ability to capture three-dimensional behavior with high resolution. While traditional one-dimensio...

Best-Estimate Plus Uncertainty (BEPU) nuclear safety analysis is an approach widely used in the field of nuclear engineering and reactor safety assessment. It aims to provide a more realistic assessment of reactor safety by considering a combination of best-estimate models and the quantification of the uncertainty of the input and code data. The Sp...

The Spanish R&D efforts in dry interim storage of spent nuclear fuel (SNF) are mainly focused on supporting safety under storage and transportation, including intermediate operations. Major actors involved are research entities, academia, the waste management organization and industry. Experimental and modelling activities are carried out to unders...

The dry storage of nuclear spent fuel is a common solution chosen by many nuclear power plants to store their fuel after depleting the nuclear spent fuel pool. The process of designing and licensing a cask is long and complicated. It is not uncommon that a cask design suffers modifications after being licensed, which must be approved by the regulat...

Computational Fluid Dynamic (CFD) codes are a widely used tool in spent nuclear fuel dry storage systems applications. A recurrent simplification used in most spent nuclear fuel CFD models found in the literature is the complete removal (or a non-explicit modelling) of the fuel assemblies flow nozzles and the spacer grids. In this paper, a full sco...

The interest in medium to long-term nuclear spent fuel storage is growing worldwide. After being cooled during several years in the spent fuel pool, the fuel can be stored in dry casks in a second phase, being moved to a temporary repository for the mid-term and/or into a deep storage facility for the long-term. To ensure the fuel integrity, the pe...

The evaluation of the containment spray, a safety system used in most nuclear power plants operating worldwide, is the main research line of GO-MERES, a collaboration agreement between UPM and CSN. In a companion paper, we have presented a validation of GOTHIC8.3(Q.A.) using experiments on the spray system performed in the experimental facility PAN...

Computational Fluid Dynamic (CFD) codes have been used widely in the last years to study the thermo-hydraulics of nuclear spent fuel dry storage systems. In this study, a new model has been developed with the CFD code ANSYS CFX, which allows to assess, simultaneously, the inner details of the cask and the cooling effects of the environment on it, w...

Resumen-En la industria nuclear existe un creciente interés por contar con información detallada sobre el comportamiento de los contendedores de combustible gastado ante diversos escenarios y en el medio y largo plazo. La utilización de códigos de dinámica de fluidos computacional (CFD) para realizar cálculos best-estimate es un método de simulació...

En la industria nuclear existe un creciente interés por contar con información detallada sobre el comportamiento de los contendedores de combustible gastado ante diversos escenarios y en el medio y largo plazo. La utilización de códigos de dinámica de fluidos computacional (CFD) para realizar cálculos best-estimate es un método de simulación útil y...

As the pools in nuclear power plants are reaching their full capacity, the use of dry cask systems to store the spent fuel is growing. This leads to an increasing need of a better understanding of nuclear spent fuel behavior during its storage phase in dry casks, ultimately, ensuring that the safety limits are not compromised. Previous computationa...

AP1000® Generation III+ reactor bases its safety concept on passive systems, differently from the previous Generation II reactors. This fact has led the approximations and methodologies previously used for modeling active safety systems to be reviewed and adapted to simulate the physics of passive systems. Diverse studies about the AP1000 containme...

After the Fukushima accident, the interest on hydrogen combustion hazard management has increased considerably. Many European nuclear plants reinforced their strategies for severe accident management with the installation of Passive Autocatalytic Recombiners (PARs) and Filtered Containment Venting Systems (FCVS), among others. In this study, the hy...

The AP1000 Passive Residual Heat Removal (PRHR) system plays a significant role as it helps to remove the core decay heat, using the In-containment Refueling Water Storage Tank (IRWST) as a heat sink. The IRWST is located above the core, promoting natural circulation and allowing to be the mid-term heat-sink for the reactor core. The thermo-hydraul...

To understand the behavior of commercial nuclear fuel during its storage phase in spent fuel dry storage casks, simulations are necessary and to validate the simulation models, full-scale experiments are required. Therefore, an experimental facility called the dry cask simulator (DCS) was built in Sandia National Laboratories with the purpose of pr...

En la industria nuclear existe un creciente interés por contar con información detallada sobre el comportamiento de los contendedores de combustible gastado ante diversos escenarios. Debido al elevado coste de la realización de experimentos, una alternativa viable es la utilización de códigos de dinámica de fluidos computacional (CFD) para realizar...

Peak Cladding Temperature (PCT) becomes the reference parameter when assessing the safety during the spent fuel drystorage phase and is dependent on the overall thermal performance. Many variables have been studied in the literature, including the wind speed, air humidity, or heat load. Nevertheless, very limited attention has been given to the imp...

The management of the spent fuel is one of the main challenges of nuclear power. The most popular option for mid-term storage is the use of dry casks. The dry cask storage system is safe and thermally efficient, as it doesn´t need any additional cooling system than natural convection and radiation. In some cases, it could be needed to unload the dr...

Desde la industria nuclear hay un creciente interés por contar con información detallada best estimate sobre el comportamiento de los contendedores de combustible gastado ante diversos escenarios. Debido al elevado coste de la realización de experimentos, una alternativa viable es la utilización de códigos de dinámica de fluidos computacional (CFD)...

Boiling water reactors use the Pressure Suppression Pool (PSP) to relieve the containment pressure in case of an accident. During the event of a Loss of Coolant Accident (LOCA), drywell air and steam are injected into the PSP through blowdown pipes. This may lead to thermal stratification, which is a relevant safety issue as it leads to higher wate...

Severe accidents in nuclear power plants are still an active field in containment safety research because of their complex phenomenology, their wide range of consequences and the uncertainty involved in the computational modeling. The main contributors to this difficulty are the potential harsh conditions in the containment due to hydrogen combusti...

Containment safety analysis is a field in expansion given the complex phenomenology developed during an accident, and the recent code capabilities improvement. In this paper, a modeling strategy to create complex containments with porous CFDs in a single control volume is presented. This methodology involves using solely Computer Aided Design (CAD)...

Licensing process of the nuclear power plants are supported by computational analysis. The large and complex geometry of the containment building, along with a wide spectrum of phenomena , create a demanding scenario for thermal-hydraulics calculation. Over the last decades , this complexity has been addressed with the Lumped Parameters (LP) approa...

The response of the nuclear industry to decrease the potential risk in case of unavailability of AC power was the inclusion of passive mechanisms for the safety systems to the Generation III+ reactors. Regarding to the AP1000 reactor, it is equipped with a Passive Containment cooling System (PCS), relying on natural phenomena such as gravity and de...

The authors implemented two integer programming models to monitor student achievements at the Universidad Politécnica de Madrid by periodically assessing (i.e., testing) these students.

Ante el creciente interés en la industria nuclear por el comportamiento de los contenedores de combustible gastado a medio plazo, es necesario el desarrollo de modelos para el análisis best estimate de los mismos. La dinámica de fluidos computacional (CFD) es una de las herramientas habitualmente empleadas a nivel internacional para estos estudios....

The continuous development of evaluation models in the nuclear industry makes mandatory to keep the highest safety standards. The model updating performed for the Almaraz NPP containment has been focused on: the addition of the newest phenomenological models available on the latest version of GOTHIC, the use of new data that increase the detail in...

In order to enhance Generation II and III reactors safety, Generation III+ reactors have included passive mechanisms for their safety systems that do not need alternating current supply to work correctly. Specifically, the AP1000® reactor uses these mechanisms in both its safety systems including the Passive Containment Cooling System (PCS), which...

The AP1000® advanced reactor passive safety systems are based on natural phenomena to ensure the containment integrity during an accident. One of the most important passive systems is the Passive Core Cooling System (PXS) which includes the In-containment Refueling Water Storage Tank (IRWST), a pool that serves as a heat sink for the Passive Residu...

In order to enhance Generation II reactors safety, Generation III+ reactors have adopted passive mechanisms for their safety systems. In particular, the AP1000® reactor uses these mechanisms to evacuate heat from the containment by means of the Passive Containment Cooling System (PCS). The PCS uses the environment atmosphere as the ultimate heat si...

Containment safety analyses are still performed using conservative assumptions due to the difficulties in modeling the phenomena associated, and/or to satisfy the regulatory bodies' requirements. Nevertheless, in order to obtain accurate predictions, realistic analyses have to be performed. In 1989, the U.S. NRC modified the licensing requirements...

The operation of recently implanted low-leakage seals after Fukushima has altered the analysis of classical PWR Station Blackout (SBO) sequences , as Seal Loss of Coolant Accident (SLOCA) is no longer one of the dominant factors in the accident progression . An analysis of different management strategies in non-SLOCA sequences has been performed by...

Hydrogen management is still one of the main nuclear safety topics because of its violent reaction with oxygen. During a severe accident, hydrogen can be generated and it can be released into the containment atmosphere. To deal with this threat, the severe accident management guidelines must be used. These guidelines include several actions to coup...

Containment safety analyses of Design Basis Accidents (DBAs) rely on the use of Lumped Parameters (LP) codes, and therefore, on pressure and temperature averaged values. In those analyses the full containment building is usually modeled with a single or few computational cells. During the latest years, many efforts have been done to develop 3D cont...

The evaluation of Passive Autocatalytic Recombiners (PARs) performance has been foreseen from the EU stress tests in the framework of a complementary and comprehensive review of the safety of the Nuclear Power Plants (NPPs). The study presented in this work analyses the size, location and number of the PARs to minimise the risk arising from a hydro...

During a severe accident in a nuclear power plant, there are several challenges that need to be faced. To coup with a containment overpressure, the venting action will lower the pressure but it will release radioactivity to the environment. In order to reduce the radioactivity released, a filtered containment venting system can be used to retain io...

The AP1000® advanced reactor passive safety systems are based on natural phenomena to ensure the containment integrity during an accident. One of the most important passive systems is the Passive Core Cooling System (PXS) which includes the In-containment Refueling Water Storage Tank (IRWST), a pool that serves as a heat sink for the Passive Residu...

During a nuclear accident, several drop-formation processes occur, such as flashing in the break, the use of containment sprays, or film condensation. Each phenomenon has a characteristic drop size which it is important in containment thermal-hydraulics. In numerical simulations of containments, it is common to model the drop phase with a unique si...

A VTT Fukushima Daiichi Unit 3 (1F3) method for estimation of liquids and consequences of releases (MELCOR) model was modified to simulate the Fukushima Daiichi Unit 2 (1F2) accident. Five simulations were performed using different modeling approaches. The model 1F2 v1 includes only the basic modifications to reproduce the 1F2 accident. The model 1...

The passive safety systems of the AP1000 ® nuclear reactor are based on natural phenomena to ensure containment integrity during an accident. One of the most important passive systems is the Passive Containment Cooling System (PCS), responsible for cooling the containment in operation and during an accident, guaranteeing the core residual heat evac...

AP1000® is a Generation III+ reactor in which all safety systems relay on passive elements. AP1000 containment includes an In-containment Refueling Water Storage Tank (IRWST), a stainless steel liner that acts as a heat sink, and a Passive Residual Heat Removal System (PRHRS). In a containment Design Basis Accident (DBA) all these components are in...

Hasta comienzos de siglo, China no era un país con un peso específico en la industria nuclear mundial. Sin embargo, desde que el Gobierno chino apostó en 2005 por el desarrollo de la energía nuclear, se ha convertido en el país con más reactores en construcción y con mayor abanico de tecnologías, propias e importadas. En este artículo se expone una...

The Integrated Safety Assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council, has been applied to PWR Station Blackout (SBO) sequences in the context of the IDPSA (Integrated Deterministic-Probabilistic Safety Assessment) network. The ISA methodology allows obtaining the Damage Domain (DD), the region of the uncertain paramet...

Containment Design Basis Accident analysis are usually performed using lumped-parameters models as this kind of models do not require high computational resources. Moreover, they offer overall good results of average pressure and temperature evolution. However, in order to perform a detailed local analysis, a thermal-hydraulic behavior study of eve...

A Fukushima Daiichi Unit 3 MELCOR model from Technical Research Centre of Finland (VTT) was modified to simulate the Fukushima Daiichi Unit 2 accident. Several simulations were performed using three different modeling approaches. The base model (Case 0) includes only the basic modifications to reproduce the accident. The intermediate model (Case 1)...

A Fukushima Daiichi Unit 3 MELCOR model from Technical Research Centre of Finland (VTT) was modified to simulate the Fukushima Daiichi Unit 2 accident. Several simulations were performed using three different modeling approaches. The base model (1F2 v1) includes only the basic modifications to reproduce the accident. The intermediate model (1F2 v2)...

Efficient Communication is usually pointed out as a valuable skill for engineering companies. However, teaching efficient communication skills is normally not a first order priority in engineering schools in Spain. There are optional courses to learn how to speak in public or to write correctly, but in general these haven't been considered as core...

Creativity is usually defined as the capability of generating ideas or products that are new, useful and that have an impact. This looks like a very important aspect in an engineer's life. Some authors relate creativity and intelligence till the point of discussing if they are in fact the same thing. On the other hand, creativity has never been a...

El accidente de rotura de tubos de un generador de vapor (Steam Generator Tube Rupture, SGTR) en los reactores de agua a presión es uno de los transitorios más exigentes desde el punto de vista de operación. Los transitorios de SGTR son especiales, ya que podría dar lugar a emisiones radiológicas al exterior sin necesidad de daño en el núcleo previ...

The Stress Tests accomplished by the European nuclear plants assume to be a complementary and comprehensive review of the safety of nuclear facilities, taking into account the events occurred in Fukushima Daiichi. The analysis of Passive Autocatalytic Recombiners (PARs) installation in Cofrentes NPP (BWR Mark III, 1092 MWe) was a requirement emerge...

The current main figure of merit for risk based decision making process based on Probabilistic Safety Assessment level 1 is usually related with the fuel failure (i.e., Peak Cladding Temperature (PCT)>1477.15 K). In this approach, the core damage is the first and necessary step in a potential radiological release, being the containment failure the...

During a severe accident in light water reactor (LWR), hydrogen concentration can overpass the flammability limits locally, so the correct simulation of its behavior during a release is critical. The capability assessment of computational fluid dynamics tools to calculate the hydrogen distribution under different conditions has been the focus of in...

The confinement of radioactive material in a nuclear power plant, including the discharge control and the release minimization, is a fundamental safety function to be ensured in a design basis accident (DBA). For plant licensing analysis, the containment is usually modeled with a lumped parameter approach. Inherent to the lumped parameter approach...

La creatividad suele ser definida comúnmente como “la habilidad de generar ideas o productos que sean novedosos, apropiados y que tengan impacto” . La característica apropiados se refiere a que deben ser ideas o productos útiles, que en caso del arte, se podría intercambiar utilidad por belleza. Por lo tanto, productos novedosos pero no útiles o vi...

The AP1000® is an advanced Pressurized Water Reactor (PWR) design developed by Westinghouse which implements passive safety systems to provide core cooling in case of accident. The development of best-estimate codes produced the evolution of conservative safety analysis towards the so-called best-estimate plus uncertainty (BEPU) analysis in order t...

The purpose of this study is to establish a detailed three-dimensional model of Cofrentes NPP BWR/6 Mark III containment building using the containment code GOTHIC 8.0. This paper presents the model construction, the phenomenology tests conducted and the selected transient for the model evaluation. In order to study the proper settings for the mod...

The detailed 3D calculation of a boron slug transient with neutronics/thermal-hydraulics coupled systems has been a highly demanding exercise because of the difficulty coupling with boron transport models. Within subproject 3 of the FP7 European Project NURISP, two neutron kinetics codes, COBAYA3 and DYN3D, coupled with the thermal-hydraulics code...

The use of suppression pools to limit the pressure increase in nuclear power plant containments has been a common strategy since the development of the first BWR designs more than 50 years ago. In the new Generation III+ PWR reactors design, suppression pools design has been also incorporated; AP1000 (Westinghouse), EPR (AREVA) and APR1400 (KEPCO)....

This project was started in December 2013, with the International Benchmark Exercise (IBE-3) performed in the PANDA facility at the PSI. The benchmark provides a good start to a deep study of the simulation of hydrogen distribution with the GOTHIC code. The Parameter Influence Chart was the final conclusion to this study and it could be a worldwide...

The Fukushima accident has drawn attention even more to the importance of external events and loss of energy supply on safety analysis. Since 2011, several Station Blackout (SBO) analyses have been done for all type of reactors. The most post-Fukushima studies analyze a pure and straight SBO transient, but the Fukushima accident was more complex th...

Transversal skills are known to be often forgotten in the engineering studies plans or relegated to the will of the professors. The nuclear engineering subjects at the Universidad Politécnica de Madrid are taught in the last courses of the degree or at master and doctorate levels. Therefore, the alumni are expected to be almost ready for the workin...

The new reactor concepts proposed in the Generation IV International Forum require the development and validation of computational tools able to assess their safety performance. In the first part of this paper the models of the ESFR design developed by several organisations in the framework of the CP-ESFR project were presented and their reliabilit...

From the 60s to the 90s, a great number of events related to the Emergency Core Cooling Systems Strainers have been happened in all kind of reactors all over the world. Thus, the Nuclear Regulatory Commission of the USA emitted some Bulletins to address the concerns about the adequacy of Emergency Core Cooling Systems (ECCS) strainer performance at...

The simulation of design basis accidents in a containment building is usually conducted with a lumped parameter model. The codes normally used by Westinghouse Electric Company (WEC) for that license analysis are WGOTHIC or COCO, which are suitable to provide an adequate estimation of the overall peak temperature and pressure of the containment. Ho...

The integrated Safety Assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council (CSN), has been applied to a thermal-hydraulic analysis of PWR Station Blackout (SBO) sequences in the context of the IDPSA (Integrated Deterministic-Probabilistic Safety Assessment) network objectives. The ISA methodology allows obtaining the damage...

The simulation of design basis accidents in a containment building is usually conducted with a lumped parameter model. The codes normally used by Westinghouse Electric Company (WEC) for that license analysis are WGOTHIC or COCO, which are suitable to provide an adequate estimation of the overall peak temperature and pressure of the containment. How...

The integrated Safety Assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council (CSN), has been applied to a thermal-hydraulic analysis of PWR Station Blackout (SBO) sequences in the context of the IDPSA (Integrated Deterministic-Probabilistic Safety Assessment) network objectives. The ISA methodology allows obtaining the damage...

The Japan Atomic Energy Agency (JAEA) began in 2009 the OECD/NEA ROSA-2 project, to realize relevant thermal-hydraulic analysis for safety in light water reactors. The operator management during a steam generator tube rupture (SGTR) accident in a PWR reactor was considered as one of the relevant issues to be studied. The SB -SG-15 experiment was ac...

The Westinghouse AP1000® reactor is an advanced design whose safety systems are based on natural mechanisms such as gravity or natural circulation, namely, they are passive safety systems. Because of the passive nature of the safety related systems and its dependency on small changes on certain variables (e.g. pressure), it is necessary to confirm...

During the last ten years new trends in deterministic and probabilistic safety analysis have been developed. The main part of the new methodologies is motivated by the Generation IV reactors evaluation. In those reactors, the concepts of safety margins that apply to the light water reactors are not directly applicable, needing a different definitio...

One of the main objectives of the Master on Nuclear Science and Technology implemented in the Universidad Politecnica de Madrid, is the training for the development of methodologies of simulation and advanced analysis necessary in research and in professional work in the nuclear field, for Fission Reactors and Nuclear Fusion, including fuel cycle a...

This paper aims to study the feasibility of implementing a strategic plan for a gradual introduction of zero-emission vehicles in the city of Madrid during 2014–2024. The study estimate the amount of emissions saved if the electrical energy needed for the vehicles is generated with nuclear power plants. The use of zero-emission vehicles could play...

An analysis of LSTF Upper Head Break experiment (OECD/NEA ROSA test 6.1) has been performed with TRACE code. This test, included within the OECD/NEA ROSA project, attempts to analyze the phenomenology and different accident management actions after the occurrence of a Upper Head break with failure of High Pressure Safety Injection (HPSI). The compa...

The new reactor concepts proposed in the Generation IV International Forum (GIF) are conceived to improve the use of natural resources, reduce the amount of high-level radioactive waste and excel in their reliability and safe operation. Among these novel designs sodium fast reactors (SFRs) stand out due to their technological feasibility as demonst...

IDPSA (Integrated Deterministic-Probabilistic Safety Assessment) is a family of methods which use tightly coupled probabilistic and deterministic approaches to address respective sources of uncertainties, enabling Risk informed decision making in a consistent manner. The starting point of the IDPSA framework is that safety justification must be bas...

The advantages of fast-spectrum reactors consist not only of an efficient use of fuel through the breeding of fissile material and the use of natural or depleted uranium, but also of the potential reduction of the amount of actinides such as americium and neptunium contained in the irradiated fuel. The first aspect means a guaranteed future nuclear...

The Nuclear Safety research requires a wide international collaboration of several involved groups. In this sense this paper pretends to show several examples of the Nuclear Safety research under international frameworks that is being performed in different Universities and Research Institutions like CIEMAT, Universitat Politècnica de Catalunya (UP...

Steam Generator Tube Rupture (SGTR) sequences in Pressurized Water Reactors are known to be one of the most demanding transients for the operating crew. SGTR are a special kind of transient as they could lead to radiological releases without core damage or containment failure, as they can constitute a direct path from the reactor coolant system to...

The Integrated Safety Assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council (CSN), has been applied to a thermo-hydraulical analysis of a Westinghouse 3-loop PWR plant by means of the dynamic event trees (DET) for Steam Generator Tube Rupture (SGTR) sequences. The ISA methodology allows obtaining the Damage Domains of a SGTR...

The integrated Safety Assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council (CSN), has been applied to a thermal-hydraulic analysis of PWR Station Blackout (SBO) sequences in the context of the IDPSA (Integrated Deterministic-Probabilistic Safety Assessment) network objectives. The ISA methodology allows obtaining the damage...

The integrated Safety Assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council (CSN), has been applied to a thermal-hydraulic analysis of PWR Station Blackout (SBO) sequences in the context of the IDPSA (Integrated Deterministic-Probabilistic Safety Assessment) network objectives. The ISA methodology allows obtaining the damage...

The Integrated Safety Assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council (CSN), has been applied to a thermo-hydraulical analysis of a Westinghouse 3-loop PWR plant by means of the dynamic event trees (DET) for Steam Generator Tube Rupture (SGTR) sequences. The ISA methodology allows obtaining the SGTR Dynamic Event Tree...

Steam Generator Tube Rupture accidents in Pressurized Water Reactors are known to one of the most demanding transients for the operating crew. SGTR are special transient as they could lead to radiological releases without core damage or containment failure, as they can constitute a direct path to the environment. The SGTR is analysed from a Determi...