Review of current Severe Accident Management (SAM) approaches for Nuclear Power Plants in Europe

The Fukushima accidents highlighted that both the in-depth understanding of such sequences and the development or improvement of adequate Severe Accident Management (SAM) measures are essential in order to further increase the safety of the nuclear power plants operated in Europe. To support this effort, the CESAM (Code for European Severe Accident Management) R&D project, coordinated by GRS, started in April 2013 for 4 years in the 7th EC Framework Programme of research and development of the European Commission. It gathers 18 partners from 12 countries: IRSN, AREVA NP SAS and EDF (France), GRS, KIT, USTUTT and RUB (Germany), CIEMAT (Spain), ENEA (Italy), VUJE and IVS (Slovakia), LEI (Lithuania), NUBIKI (Hungary), INRNE (Bulgaria), JSI (Slovenia), VTT (Finland), PSI (Switzerland), BARC (India) plus the European Commission Joint Research Center (JRC). The CESAM project focuses on the improvement of the ASTEC (Accident Source Term Evaluation Code) computer code. ASTEC,, jointly developed by IRSN and GRS, is considered as the European reference code since it capitalizes knowledge from the European R&D on the domain. The project aims at its enhancement and extension for use in severe accident management (SAM) analysis of the nuclear power plants (NPP) of Generation II-III presently under operation or foreseen in near future in Europe, spent fuel pools included. In the frame of the CESAM project one of the tasks consisted in the preparation of a report providing an overview of the Severe Accident Management (SAM) approaches in European Nuclear Power Plants to serve as a basis for further ASTEC improvements. This report draws on the experience in several countries from introducing SAMGs and on substantial information that has become available within the EU “stress test”. To disseminate this information to a broader audience, the initial CESAM report has been revised to include only public available information. This work has been done with the agreement and in collaboration with all the CESAM project partners. The result of this work is presented here.JRC.F.5-Nuclear Reactor Safety Assessmen

Evaluation of human reliability analysis methods addressing cognitive error modelling and quantification

Actions and errors by the operating personnel, which are of significance for the safety of a technical system, are classified according to various criteria. Each type of action thus identified is roughly discussed with respect to its quantifiability by state-of-the-art human reliability analysis (A) within a probabilistic safety assessment (PSA). Thereby, the principal limits of quantifying human actions are discussed with special emphasis on data quality and cognitive error modelling. In this connection, the basic procedure for a IIRA is briefly described under realistic conditions. With respect to the quantitative part of a IRA - the determination of error probabilities - an evaluating description of the standard method THERP (Technique of Human Error Rate Prediction) is given using eight evaluation criteria. Furthermore, six new developments (EdF'sPFHRA, HCR, HCR/ORE, SLIM, HEART, INTENT) are briefly described and roughly evaluated. The report concludes with a catalogue of requirements for HRA methods

Development of advanced methods and related software for human reliability evaluation within probabilistic safety analyses

Human Reliability Analysis (HRA) is an important part of Probabilistic Safety Analysis (PSA). The first part of this report consists of an overview of types of human behaviour and human error including the effect of significant performance shaping factors on human reliability. Particularly with regard to safety assessments for nuclear power plants a lot of HRA merhods have been developed. The most imortant of these methods are presented and discussed in the report, together with techniques for incorporating HRA into PSA and with models of operator cognitive behaviour. Based on existing HRA methods the concept of a software system is described. For the development of this system the utilization of modern programmig tools is proposed; the essential goal is th effective application of HRA methods. A possible integration of computeraided HRA within PSA is discussed. The features of Expert System Technology and examplas of applications (PSA, HRA) are presented in four appendices

Entscheidungsfehler des Betriebspersonals von Kernkraftwerken als Objekt probabilistischer Risikoanalysen Methodische Erweiterungen auf der Basis einer differenzierten Betrachtungsweise sicherheitsgerichteter Ziele

The integration of human error - also called man-machine system analysis (MMSA) - is an essential part of probabilistic risk assessment (PRA). A new method is presented which allows for a systematic and comprehensive PRA inclusions of decision-based errors due to conflicts or similarities. For the error identification procedure, new question techniques are developed. These errors are shown to be identified by looking at retroactions caused by subordinate goals as components of the overall safety relevant goal. New quantification methods for estimating situation-specific probabilities are developed. The factors conflict and similarity are operationalized in a way that allows their quantification based on informations which are usually available in PRA. The quantification procedure uses extrapolations and interpolations based on a poor set of data related to decision-based errors. Moreover, for passive errors in decision-making a completely new approach is presented where errors are quantified via a delay initiating the required action rather than via error probabilities. The practicability of this dynamic approach is demonstrated by a probabilistic analysis of the actions required during the total loss of feedwater event at the Davis-Besse plant 1985. The extensions of the ''classical'' PRA method developed in this work are applied to a MMSA of the decay heat removal (DHR) of the ''HTR-500''. Errors in decision-making - as potential roots of extraneous acts - are taken into account in a comprehensive and systematic manner. Five additional errors are identified. However, the probabilistic quantification results a nonsignificant increase of the DHR failure probability. (orig.)Die Einbeziehung von Operateurfehlern - Mensch-Maschine-Systemanalyse (MMSA) - ist wesentlicher Bestandteil einer probabilistischen Risikoanalyse (PRA). Es wird eine Methode vorgestellt, mit der sich Entscheidungsfehler aufgrund der Faktoren Konflikt und Aehnlichkeit systematisch und umfassend in eine MMSA integrieren lassen. Zur Identifizierung der entsprechenden Situationen im Stoerfallablauf wird eine neuartige Fragetechnik entwickelt und gezeigt, dass solche Fehler als Rueckwirkungen untergeordneter Ziele auftreten koennen. Solche Rueckwirkungen sind ueber eine differenzierte Betrachtung sicherheitsgerichteter Ziele identifizierbar. Zur Quantifizierung wird eine neue Methode entwickelt, mit der sich situationsspezifisch Wahrscheinlichkeiten fuer Entscheidungsfehler schaetzen lassen. Es gelingt, die Faktoren Konflikt und Aehnlichkeit so zu operationalisieren, dass sie mit den Informationen, die einem PRA-Anwender ueblicherweise zur Verfuegung stehen, quantitativ zugaenglich sind. Das Quantifizierungsverfahren basiert auf Extra- und Interpolationen zu den wenigen Daten, die zur Zeit ueber Entscheidungsfehler von Operateuren existieren. Ausserdem wird fuer passive Entscheidungsfehler (Unterlassungen notwendiger Handlungen) ein neuartiger Modellansatz vorgestellt, der solche Fehler ueber Verzoegerungszeiten quantifiziert. Die praktische Durchfuehrbarkeit dieses dynamischen Ansatzes wird am Beispiel einer probabilistischen Analyse der konkreten Operateurmassnahmen gezeigt, die waehrend des Stoerfalls im Kernkraftwerk Davis-Besse (1985) angefordert wurden. Die Erweiterungen der ''klassischen'' MMSA-Methodik werden am Beispiel der Nachwaermeabfuhr (NWA) des HTR-500 angewendet. Entscheidungsfehler (als Ursachen ungeplanter Handlungen) werden systematisch und umfassend beruecksichtigt. Fuenf zusaetzliche Entscheidungsfehler werden identifiziert. Unter Einbeziehung der entsprechenden Fehlerwahrscheinlichkeiten ergibt sich keine signifikante Erhoehung der Ausfallwahrscheinlichkeit bzw. Unverfuegbarkeit der NWA. (orig.)SIGLEAvailable from TIB Hannover: RA 831(2811) / FIZ - Fachinformationszzentrum Karlsruhe / TIB - Technische InformationsbibliothekBundesministerium fuer Forschung und Technologie (BMFT), Bonn (Germany)DEGerman

Zuverlaessigkeitsanalysen fuer Personalhandlungen und Bauteile als Beitraege zur systematischen Erweiterung der Methodik probabilistischer Risikoanalysen Endbericht

SIGLEAvailable from TIB Hannover: D.Dt.F. QN1(1,24) / FIZ - Fachinformationszzentrum Karlsruhe / TIB - Technische InformationsbibliothekDEGerman

Development of advanced methods and related software for human reliability evaluation within probabilistic safety analyses

Human Reliability Analysis (HRA) is an important part of Probabilistic Safety Analysis (PSA). The first part of this report consists of an overview of types of human behaviour and human error including the effect of significant performance shaping factors on human reliability. Particularly with regard to safety assessments for nuclear power plants a lot of HRA methods have been developed. The most important of these methods are presented and discussed in the report, together with techniques for incorporating HRA into PSA and with models of operator cognitive behaviour. Based on existing HRA methods the concept of a software system is described. For the development of this system the utilization of modern programming tools is proposed; the essential goal is the effective application of HRA methods. A possible integration of computeraided HRA within PSA is discussed. The features of Expert System Technology and examples of applications (PSA, HRA) are presented in four appendices. (orig.)Die Untersuchung menschlicher Zuverlaessigkeit (Human Reliability Analysis, HRA) ist ein bedeutsamer Teil probabilistischer Sicherheitsanalysen (PSA). Der erste Teil dieses Berichtes besteht aus einem Ueberblick zu den Arten menschlichen Verhaltens und menschlicher Fehler einschliesslich der signifikanten Einflussgroessen auf menschliche Zuverlaessigkeit. Insbesondere fuer Sicherheitsbewertungen von Kernkraftwerken wurde eine Reihe von HRA-Methoden entwickelt. Die wichtigsten von ihnen werden vorgestellt und diskutiert, zusammen mit Vorgehensweisen zur Einbeziehung von HRA in PSA sowie mit Modellierungen des kognitiven Verhaltens von Betriebsmannschaften. Auf der Basis vorhandener HRA-Methoden wird das Konzept eines Computer-Programmsystems beschrieben. Zur Entwicklung dieses Systems wird die Verwendung moderner Programmiermethoden vorgeschlagen; das Hauptziel ist die wirkungsvolle Anwendung von HRA-Methoden. Eine moegliche Einbeziehung computerunterstuetzter HRA in PSA wird diskutiert. Die grundsaetzliche Besonderheit von Expertensystemen und Anwendungsbeispiele (PSA, HRA) werden in vier Anhaengen vorgestellt. (orig.)Available from TIB Hannover: RA 831(2928) / FIZ - Fachinformationszzentrum Karlsruhe / TIB - Technische InformationsbibliothekSIGLEDEGerman