Consejo de Seguridad Nuclear

Madrid, Spain

Consejo de Seguridad Nuclear

Madrid, Spain
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Munoz-Cobo J.L.,Polytechnic University of Valencia | Escriva A.,Polytechnic University of Valencia | Mendizabal R.,Consejo de Seguridad Nuclear | Pelayo F.,Consejo de Seguridad Nuclear | Melara J.,Iberdrola
Nuclear Engineering and Design | Year: 2014

This paper shows an application of the CSAU methodology to an ATWS in a BWR reactor, when the temperature of the suppression pool is taken as the critical safety parameter. The method combines CSAU methodology with recent techniques of information theory. In this paper we use auxiliary tools to help in the evaluation and improvement of the parameters distribution that enter in the elements II and III of CSAU based methodologies. These tools have been implemented in two FORTRAN programs: GEDIPA (Generation of the Parameter Distribution) and UNTHERCO (Uncertainty in Thermal Hydraulic Codes). The first one analyzes the information data available on a given parameter or parameters with the goal to know all the information about the probability distribution function of these parameters. The second apply information theory methods, as the maximum entropy principle (MEP) and the maximum relative entropy Principle (MREP), in order to build conservative distribution functions for the parameters from the available data. Also, the distribution function of a given parameter can be updated using the MREP principle when new information is provided. UNTHERCO performs the MONTECARLO sampling for a given set of parameters when the distribution function of these parameters is previously known. If the distribution of a parameter is unknown, then, the MEP is applied to deduce the distribution function for this parameter. © 2014 Elsevier B.V. All rights reserved.

Tasset D.,Institute for Radiological Protection and Nuclear Safety | Frischknecht A.,Swiss Federal Nuclear Safety Inspectorate ENSI | Lamarre G.,Oecd Nuclear Energy Agency | Gil-Montes B.,Consejo de Seguridad Nuclear
8th International Topical Meeting on Nuclear Plant Instrumentation, Control, and Human-Machine Interface Technologies 2012, NPIC and HMIT 2012: Enabling the Future of Nuclear Energy | Year: 2012

The crucial role that human and organizational factors (HOF) plays in the safety performance of high hazard industries, including nuclear has been proven time and again based on the lessons coming out of such high profile events and accidents such as Texas City, Deepwater Horizon, Challenger/Columbia shuttle accidents as well as the Chernobyl and other well known nuclear accidents. Further, nuclear industry accidents and events such as Fukushima Daiichi NPP, Sellafield MOx, Tokai Mura and Davis-Besse have increased awareness of the contribution to nuclear safety performance that is made by a licensee's leadership and the way in which it manages for safety. Within this context, the NEA/CSNI Working Group on Human and Organisational Factors (WGHOF) brings HOF experts together, representing regulators, technical safety organizations, research institutions and industry to discuss and develop common positions on some of the key challenges facing the safety of the nuclear industry in these specialized fields. Realizing that the strength of its influence and ability to advance the fields of knowledge of HOF (which includes safety culture) lies within its ability to be able to bring together the key international voices in the field of HOF, WGHOF has consistently reached out to other organizations such as the IAEA, EC-JRC, WANO in its work. The products of WGHOF from safety reports including best practice guides to technical opinion papers have had a significant impact on both the operational and regulatory approaches in the areas of human and organizational factors within the participating OECD-NEA member and associated member countries. Spanning the topical areas of HOF in new technology, safety culture, leadership and managing for safety, human reliability analysis and organizational performance and capabilities, to name a few, WGHOF has had a very positive influence on harmonizing the thinking and approaches in these technical areas internationally. Consequently, these international collaborative efforts have translated into some important safety improvements in the participating countries. Looking to the future, WGHOF recognizes that with the most recent Japanese accident at Fukushima Daiichi, the international nuclear safety community is once again looking to it to address some of the key human and organizational factors issues (for example, human performance under severe accident conditions over an extended time period, training and validation of training procedures for severe accident situations, and many others) coming out of the lessons learnt from this tragedy. WGHOF is well positioned and well structured to continue to meet these challenges and deliver key products in a timely fashion into the hands of the regulators, technical safety organizations and industry participants.

Gil J.,Indizen Technologies S.L. | Fernandez I.,Indizen Technologies S.L. | Murcia S.,Indizen Technologies S.L. | Gomez J.,Indizen Technologies S.L. | And 9 more authors.
Nuclear Engineering and Design | Year: 2011

Over the past years, many Nuclear Power Plant organizations have performed Probabilistic Safety Assessments to identify and understand key plant vulnerabilities. As part of enhancing the PSA quality, the Human Reliability Analysis is essential to make a realistic evaluation of safety and about the potential facility's weaknesses. Moreover, it has to be noted that HRA continues to be a large source of uncertainty in the PSAs. Within their current joint collaborative activities, Indizen, Universidad Politécnica de Madrid and Consejo de Seguridad Nuclear have developed the so-called SIMulator of PROCedures (SIMPROC), a tool aiming at simulate events related with human actions and able to interact with a plant simulation model. The tool helps the analyst to quantify the importance of human actions in the final plant state. Among others, the main goal of SIMPROC is to check the Emergency Operating Procedures being used by operating crew in order to lead the plant to a safe shutdown plant state. Currently SIMPROC is coupled with the SCAIS software package (Izquierdo et al.; 2008), but the tool is flexible enough to be linked to other plant simulation codes. SIMPROC-SCAIS applications are shown in the present article to illustrate the tool performance. The applications were developed in the framework of the Nuclear Energy Agency project on Safety Margin Assessment and Applications (SM2A). First an introductory example was performed to obtain the damage domain boundary of a selected sequence from a SBLOCA. Secondly, the damage domain area of a selected sequence from a loss of Component Cooling Water with a subsequent seal LOCA was calculated. SIMPROC simulates the corresponding human actions in both cases. The results achieved shown how the system can be adapted to a wide range of purposes such as Dynamic Event Tree delineation, Emergency Operating Procedures and damage domain search. © 2010 Elsevier B.V.

Flores A.,European Commission | Izquierdo J.M.,Consejo de Seguridad Nuclear | Tucek K.,European Commission | Gallego E.,Technical University of Madrid
Annals of Nuclear Energy | Year: 2014

This paper presents an assessment analysis of damage domains of the 30 MWth prototype High-Temperature Engineering Test Reactor (HTTR) operated by the Japan Atomic Energy Agency (JAEA). For this purpose, an in-house deterministic risk assessment computational tool was developed based on the Theory of Stimulated Dynamics (TSD). To illustrate the methodology and applicability of the developed modelling approach, assessment results of a control rod (CR) withdrawal accident during subcritical conditions are presented and compared with those obtained by the JAEA. © 2014 Elsevier Ltd. All rights reserved.

Mendizabal R.,Consejo de Seguridad Nuclear | Pelayo F.,Consejo de Seguridad Nuclear
American Society of Mechanical Engineers, Fluids Engineering Division (Publication) FEDSM | Year: 2010

The Technical Specifications (TS) of a nuclear power plant define the conditions for a safe normal operation. With such an objective, the TS set limits on operational parameters of the plant and give surveillance requirements for the observation of such bounds. The values of TS limits are obtained from the safety analyses of the plant. In fact, the traditional conservative methodologies of deterministic safety analysis (DSA) have been profusely used in this task. Nevertheless, in recent years realistic (also termed BEPU) methodologies have started to replace the conservative ones. This new methodologies use realistic models and assumptions and implement techniques for performing uncertainty analysis of their results. Many of them are statistical, with a probabilistic representation of uncertainty, and based on the random sampling of uncertain inputs and uncertainty propagation to the outputs. In this paper the relation between BEPU safety analyses and TS is analyzed. The authors have a deep regulatory experience in the evaluation and licensing of DSA methodologies. Safety analyses are aimed at showing that the real operation of the plant is safe, but they have a stronger goal: to prove that the allowed operation of the plant is safe. BEPU methodologies are not fitted for the estimation of TS bounds. They rather are used to prove the coherence of the safety analysis with the preestablished TS. Procedures for proving such coherence, with different degree of strictness, are discussed in the case of Monte Carlo- based methodologies. Copyright © 2010 by ASME.

Queral C.,Technical University of Madrid | Mena-Rosell L.,Technical University of Madrid | Jimenez G.,Technical University of Madrid | Sanchez-Perea M.,Consejo de Seguridad Nuclear | And 2 more authors.
International Conference on Nuclear Engineering, Proceedings, ICONE | Year: 2014

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 domain (the region of the uncertain parameters space where the damage limit is exceeded) for each sequence of interest as a function of the operator actuations times. Given a particular safety limit or damage limit, several data of every sequence are necessary in order to obtain the exceedance frequency of that limit. In this application these data are obtained from the results of the simulations performed with MAAP code transients inside each damage domain and the time-density probability distributions of the manual actions. Damage limits that have been taken into account within this analysis are: local cladding damage (PCT>1477 K); local fuel melting (T >2499 K); fuel relocation in lower plenum and vessel failure. Therefore, to every one of these damage variables corresponds a different damage domain. The operation of the new passive thermal shutdown seals developed by several companies since Fukushima accident is considered in the paper. The results show the capability and necessity of the ISA methodology, or similar, in order to obtain accurate results that take into account time uncertainties. Copyright © 2014 by ASME.

Rebollo M.J.,Technical University of Madrid | Queral C.,Technical University of Madrid | Jimenez G.,Technical University of Madrid | Gomez-Magan J.,NFQ Solutions | And 2 more authors.
Reliability Engineering and System Safety | Year: 2016

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 second one. Nevertheless, SGTR sequences in PWR plants are able to release large quantities of radioactive products without previous core damage or containment failure. For that reason, it seems necessary to analyze which sequences exceed the allowed offsite dose criteria prior to the core damage criterion. The aim of this analysis has been to evaluate the risk contribution due to the offsite dose and the core damage in case of Steam Generator Tube Rupture (SGTR) sequences at full power in a 3-loop Pressurized Water Reactor (PWR) Westinghouse-design. The study has been performed with SCAIS/MAAP and RADTRAD codes. For that purpose, this analysis unfolds the SGTR Dynamic Event Tree for both the core damage and the offsite dose risk metrics. The results indicate that dose criteria complement the PCT criterion and allow quantifying both risk contributions in SGTR sequences. © 2015 Elsevier Ltd.

Asensio E.M.,Consejo de Seguridad Nuclear | Santos R.H.,Consejo de Seguridad Nuclear
International Topical Meeting on Probabilistic Safety Assessment and Analysis, PSA 2015 | Year: 2015

The Spanish Nuclear regulatory body (CSN) has implemented a regulatory system inspired by the USNRC Reactor Oversight Process (ROP). As such, the system makes use of PSA results and insights in several tasks. Inspection results are assessed by PSA quantification, determining the consequences of licensee performance deficiencies by mapping onto a PSA model any failures to comply with rules and regulations. A requantification of the model yields the impact of the performance deficiency in the plant risk. The outcome is used to take decisions on further regulatory actions. In addition, whenever applicable, inspection scope is driven by PSA importance, be it the baseline inspections or special inspections, reactive to events. This regulatory framework requires PSA information to be disseminated throughout the organization, even among non PSA experts. A web based information system has been brought up that presents PSA hypotheses, methods and results in a consistent manner for all Spanish plants. Inspectors have ready access to this information tool within CSN's internal website. The information in the web based system stems from the licensees' PSA models. Since the Spanish NPPs use several PSA codes, a common interface to feed the web base information system is required. The OpenPSA initiative ( proposed an XML standard format for PSA model exchange, dubbed OPSA-MEF. At CSN, an XML format derived from the OpenPSA work, suitable for the quantification tools used at CSN has been the choice. Moreover, the OPSA-MEF is a platform for developing tools that convert PSA models between PSA codes, allowing the same model to be viewed, modified and quantified by different programs. This has been demonstrated by tools developed at CSN. Furthermore, as regulators, CSN staff need independent views on licensee models. These can be best accomplished by in-house tools that perform batch checks for consistency of the models.

Benikhlef T.,University of Boumerdès | Benazzouz D.,University of Boumerdès | Izquierdo J.M.,Consejo de Seguridad Nuclear | Sanchez M.,Consejo de Seguridad Nuclear
Asia-Pacific Journal of Chemical Engineering | Year: 2012

This paper focuses on the improvement of solutions to some of the problems that arise in chemical and petrochemical risk assessment studies, namely those associated with potentially undue grouping in sequences of events, caused by different time evolutions. It provides an adequate work-horse simulation model of risky scenarios able to explore a large amount of transients at a reasonable and viable cost. It is specifically suitable for the transfer of modern dynamic reliability techniques, originated in the nuclear domain, to the chemical engineering environment. As an application, it presents results of a case study to analyse Bhopal-like scenarios. We found the model adequate to discriminate and filter out the myriad of success scenarios expected in well-protected installations, screening necessary to focus the risk studies on damage situations. © 2011 Curtin University of Technology and John Wiley & Sons, Ltd.

Queral C.,Technical University of Madrid | Mena-Rosell L.,Technical University of Madrid | Jimenez G.,Technical University of Madrid | Sanchez-Perea M.,Consejo de Seguridad Nuclear | And 2 more authors.
Annals of Nuclear Energy | Year: 2016

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 parameters space where the damage limit is exceeded, for each sequence of interest as a function of the operator actuation times. Several damage limits have been taken into account within this analysis: cladding embrittlement criteria (Peak cladding temperature >1477 K); Inadequate core cooling conditions (Core Exit Thermocouples temperature >922 K); local fuel melting (fuel temperature >2499 K); fuel relocation in lower plenum and vessel failure. Other continuous damages, such as percentage of relocated fuel are also studied. Every one of these damage variables provides a specific DD. The application to the severe accident management (SAM) actions shows the capability of a methodology such as ISA in order to analyze the impact of different SAM strategies and to obtain the available times for different operator actions. © 2016 Elsevier Ltd

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