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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

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. Source

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

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. Source

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

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. Source

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

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. Source

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

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. Source

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