César Queral

• PhD
• Professor at Universidad Politécnica de Madrid

Since 2012, many NEA member countries have implemented deterministic safety analyses for operating nuclear power plants under design extension conditions without significant fuel degradation or core melt (DEC-A). However, variations persist among these countries in defining DEC-A scenarios and acceptance criteria, validation and application of comp...

Research on small modular reactors (SMRs) is gaining importance since they are key for addressing energy challenges in various sectors. These types of reactors integrate novel technologies that rely heavily on passive safety systems. Among the most developed light water reactors, SMR designs are SMART and NuScale. This work analyzes the academic bo...

This paper presents a summary of theoretical and applied achievements in the field of integrated methods of nuclear safety analysis produced by Spain's nuclear safety community. These integrated methods combine deterministic as well as probabilistic tools and include the approaches termed "Extended Best-Estimate Plus Uncertainty" (E-BEPU), dynamic...

The development of Probabilistic Safety Analysis, which began in Spain in 1986, is based on the document "Integrated Program on Development and Use of Probabilistic Safety Analysis in Spain" adopted by the Spanish regulatory body. A second edition of this document in 1998 considered the activities needed to apply its principles in different fields....

After the severe accident at Fukushima, the importance of BWR design and related structures and their contribution to the severe accident progression has increased. Fuel channel boxes, absorber crosses, water rods, and smaller primary containment design of the BWR have been considered in the ASTEC code to increase the knowledge of BWR design and as...

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...

This work has been performed within the framework of the 'High-Performance Advanced Methods and Experimental Investigations for the Safety Evaluation of Generic Small Modular Reactors' McSAFER H2020 European project. In this project, the main aim is to advance the research in safety analysis methods for water cooled Small Modular Reactors (SMRs). H...

In the present work, one of the studies performed by the Universidad Politécnica de Madrid research group within McSAFER H2020 European project is presented. This project is pursuing the aim of improving the research in the safety simulations by means of the application of conventional and advanced modeling tools to several SMRs designs. In particu...

VVER are one of the most common reactor types in the world. Moreover, a significant percentage of the Gen-III/Gen III+ reactors that are currently being built or have recently come into operation are VVER. Therefore, there is a growing interest in studying their behavior under both anticipated and accidental transients. A joint effort between the U...

One of the key lessons learned from the Fukushima Dai-ichi accident was the importance of the challenge presented by the loss of safety-related systems after a Beyond Design-Basis External Event (BDBEE). In particular, Extended Loss of AC Power (ELAP) or Loss of Ultimate Heat Sink (LUHS) can severely compromise key safety functions associated with...

Research on Small Modular Reactors (SMR) has increased over the last years due to their potential for meeting energy demands in various sectors. In these designs, novel technologies and passive safety systems are being integrated. Two of the most developed SMR designs are SMART and NuScale. This work analyzes the KSMR, a boron free core concept bas...

A review of the current thermomechanical needs along with a recommendation of the most urgent ones to be implemented in system codes are presented in this paper as part of the OECD/NEA/CSNI Specialists Meeting on Transient Thermal-hydraulics in Water Cooled Nuclear Reactors (SM-TH). In that sense, it has been widely demonstrated that considering th...

During the cooling of the reactor core, as a result of a severe accident, a strongly exothermic oxidation reaction of the zirconium cladding occurs and leads to hydrogen production, this has been observed in Fukushima, Chernobyl and Three Mile Island accidents. Several experiments have been carried out for the understanding of the phenomena that ar...

The specific configuration of safety systems of VVER-1000/V320 reactors allows a comprehensive study of the Loss of Coolant Accident (LOCA). In the present paper, a verification of the success criteria of the event trees headers for the Medium and Large Break LOCA sequences is conducted. A detailed TRACEV5P5 thermal-hydraulic model of the reactor h...

The classical Probabilistic Safety Analysis (PSA) does not include any time dependence explicitly. However, the success criteria (SC) could evolve during the cycle for some initiating events. In that sense, there is a type of sequence in which this time-dependency is quite important, the family of Anticipated Transient without Scram (ATWS) sequence...

After the nuclear accident at the Fukushima Daiichi nuclear power plant, hydrogen generation has resurfaced as a key issue, with respect to its quantification in the simulation of severe accidents, because of the potential risk of deflagration or detonation, which would compromise the integrity of the plant facilities. Severe accident simulation co...

The Instrumentation performance during a Severe Accident (SA) is currently one of the key identified gaps in Nuclear Safety. The instrumentation is licensed using the pressure and temperature reached during a design basis accident, but SA phenomena are not considered on its design limits. During a SA, it is likely that the pressure or temperature r...

This paper describes the main objectives, technical content, and status of the H2020 project entitled “High-performance advanced methods and experimental investigations for the safety evaluation of generic Small Modular Reactors (McSAFER)”. The main pillars of this project are the combination of safety-relevant thermal hydraulic experiments and num...

Modern Russian Generation-III/III+ Nuclear Power Plants (NPP) have experienced an outstanding expansion outside the Russian Federation in countries such as China, Turkey, India, Bangladesh, Egypt, Belarus or Iran including ongoing discussions to build VVER-1200 NPPs in Hungary and Finland. In this article, it is described how the safety features of...

The AZTLAN Platform project is a Mexican national initiative which aims to have a platform for analysis and design of nuclear reactors. In order to enhance the AZTLAN Platform with Uncertainty and Sensitivity (U&S) capabilities, the AZTUSIA (AZtlan Tool for Uncertainty and SensItivity Analysis) code has been developed to perform U&S analysis. The m...

After the Fukushima-Daiichi accident, different methods, strategies and guidelines were developed to improve the Severe Accident Management Guidelines (SAMG). One of these additions, named the FLEX strategies, relies on portable equipment to obtain power or water to restore key safety functions in the reactor. In the present study, the FLEX strate...

During the last decades, the safety analysis of nuclear power plants, have been shifting from conservative models and hypotheses to best estimate. Within this trend, the different safety analyses can be categorized depending on their associated conservatisms. The approach that includes expanded event trees together with best estimate model and cond...

Uncertainty and sensitivity analyses are a necessary step for best estimate calculations for licensing. In this sense, a complete statistical characterization of a sequence would be a useful tool for these assessments as it can detect behaviors and correlation of the Figures of Merit (FoM) and the input space. For this reason, the present paper pre...

Anticipated Transient Without Scram (ATWS) sequences belong to the beyond design basis events. Within ATWS sequences, the most limiting for PWRs is the Loss Of Normal Feedwater (LONF). This sequence produces a pressure peak in the reactor cooling system that can compromise its integrity. To cope with this sequence, systems like the AMSAC were insta...

Anyone clicking on the link before October 16, 2019 will be taken directly to the final version of the article on ScienceDirect, no fees are required. https://authors.elsevier.com/a/1ZdqA3OQ~fPN4h The Best Estimate Plus Uncertainty (BEPU) approach is being used worldwide for nuclear power plants licensing. This method relies on the use of best est...

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...

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...

After the accident at Fukushima Dai-ichi, considerable efforts were put on enhancing the capability of the Nuclear Power Plants to cope with conditions resulting from the loss of plant safety-related systems. The most widespread solution adopted worldwide has been to define and implement new procedures and emergency actuation plans, the so called F...

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...

In this paper, the severe accident code ASTECV2.0 is used for the verification and improvement of in-vessel Severe Accident Management (SAM) strategies in a German Konvoi PWR considering the lessons learnt from Fukushima. The scenario selected for the analysis is the total Station Blackout (SBO), which is the most risk-relevant scenario for the ref...

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...

The severe accidents at Fukushima have shown that a further development of Severe Accident Management Guidelines (SAMGs) is necessary. Within this work, the severe accident code ASTEC V2.0 is used to assess the impact of selected SAM measures on the in-vessel progression of Small Break (SBLOCA) scenarios in a generic German Konvoi PWR. The progress...

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 integrated safety assessment (ISA) methodology, developed by the Spanish nuclear safety council (CSN), has been applied to the analysis of full spectrum loss of coolant accident (FSLOCA) sequences in a 3-loop pressurized water reactor (PWR). The ISA methodology proposal starts from the unfolding of the dynamic event tree (DET), focusing on the...

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...

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...

The Westinghouse AP1000® reactor is an advanced design whose safety systems are mainly passive safety systems. Due to the passive nature of the safety related systems and its dependency on small changes on certain variables (e.g. pressure, friction coefficients) together with the use of a simplified code like MAAP in Probabilistic Risk Assessment (...

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...

This paper reviews current status of the unified approach known as integrated safety assessment (ISA), as well as the associated SCAIS (simulation codes system for ISA) computer platform. These constitute a proposal, which is the result of collaborative action among the Nuclear Safety Council (CSN), University of Madrid (UPM), and NFQ Solutions S.L...

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...

As part of AP1000® design adaptation to European Utility Requirements, Westinghouse has carried out several modifications of the standard AP1000 design. One of these modifications has been the physical separation of the normal residual heat removal system (RNS) into two independent trains which improves the plant defense in depth. The AP1000 Probab...

Este curso de una semana de duración organizado por la Cátedra Juan Manuel Kindelán pretende dar una visión global sobre distintos aspectos del COMBUSTIBLE NUCLEAR tales como su diseño, fabricación, gestión, evolución y comportamiento a lo largo de la primera y segunda fase del ciclo de combustible así como la investigación y el desarrollo actuales...

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...

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 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 report reviews several numerical aspects of boron dilution sequences in pressurized water reactors and also loss of RHRS sequences. The report is written in Spanish. Date: 2004.

This report describes the phenomenology of boron dilution sequences in pressurized water reactors. It is written in Spanish. Date: 2003.

This report describes the numerical methods implemented in the RELAP5, TRAC-BF1, TRAC-PC1 and CATHARE-2 codes. Date: 1998.

The Integrated Safety Assessment (ISA) methodology, developed by the Consejo de Seguridad Nuclear, has been applied to an analysis of Zion NPP for sequences with Loss of the Component Cooling Water System (CCWS). The ISA methodology proposal starts from the unfolding of the Dynamic Event Tree (DET). Results from this first step allow assessing the...

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...

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 performing of such systems must be evaluated through the performance of experiments and simulations with a variety of thermal–hydraulic codes. This paper presents the re...

Abstract Current Westinghouse Emergency Operating Procedures (EOPs) indicate initially that the operator must keep the reactor coolant pumps (RCPs) running during a Small Break Loss of Coolant Accident (SBLOCA) if there is unavailability of high pressure safety injection (HPSI) system in order to cool the core by forced convection. However, the cre...

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 integrated safety assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council (CSN), has been applied to a thermal–hydraulic analysis of cold leg LOCA sequences with unavailable High Pressure Injection System in a Westinghouse 3-loop PWR. This analysis has been performed with TRACE 5.0 patch 1 code. ISA methodology allows obta...

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...

An analysis of LSTF Surge Line Break experiment (OECD/NEA ROSA-2 test 1) has been performed with TRACE code. This test, included within the OECD/NEA ROSA-2 project and performed in Large Scale Test Facility (LSTF), it is attempted to analyze the phenomenology and different accident management actions after the occurrence of a surge line (SL) break...

The AP1000® PRA thermal hydraulic simulations were performed with MAAP code, which allows simulating sequences with low computational efforts. On the other hand, the use of best estimate codes allows verifying PRA results as well as obtaining a greater knowledge of the phenomenology of such sequences. The initiating event with the greatest contribu...

Emergency Operating Procedures (EOP) of US PWRs establish that reactor coolant pumps (RCP) should be tripped during a Small-Break Loss Of Coolant Accident (SBLOCA) by the operating crew, provided that the subcooling margin has been lost at the core outlet and the High-Pressure Safety Injection (HPSI) is available. On the other hand, if HPSI is unav...

The dynamic response of several RTDs located at the cold leg of a PWR has been studied. A theoretical model for the heat transfer between the RTDs and the surrounding fluid is derived. It proposes a two real poles transfer function. By means of noise analysis techniques in the time domain (autoregressive models) and the Dynamic Data System methodol...