scholarly journals Spreading of Excellence in SARNET Network on Severe Accidents: The Education and Training Programme

2012 ◽  
Vol 2012 ◽  
pp. 1-9 ◽  
Author(s):  
Sandro Paci ◽  
Jean-Pierre Van Dorsselaere
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Training Programme ◽  
Education And Training ◽  
Severe Accident ◽  
Research Network ◽  
Safety Issues ◽  
Severe Accidents ◽  
Management Procedures ◽  
And Training

The SARNET2 (severe accidents Research NETwork of Excellence) project started in April 2009 for 4 years in the 7th Framework Programme (FP7) of the European Commission (EC), following a similar first project in FP6. Forty-seven organisations from 24 countries network their capacities of research in the severe accident (SA) field inside SARNET to resolve the most important remaining uncertainties and safety issues on SA in water-cooled nuclear power plants (NPPs). The network includes a large majority of the European actors involved in SA research plus a few non-European relevant ones. The “Education and Training” programme in SARNET is a series of actions foreseen in this network for the “spreading of excellence.” It is focused on raising the competence level of Master and Ph.D. students and young researchers engaged in SA research and on organizing information/training courses for NPP staff or regulatory authorities (but also for researchers) interested in SA management procedures.

scholarly journals The European Research on Severe Accidents in Generation-II and -III Nuclear Power Plants

2012 ◽  
Vol 2012 ◽  
pp. 1-12 ◽  
Author(s):  
Jean-Pierre Van Dorsselaere ◽  
Ari Auvinen ◽  
David Beraha ◽  
Patrick Chatelard ◽  
Christophe Journeau ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Computer Code ◽  
Severe Accident ◽  
Research Network ◽  
Combustion Source ◽  
Joint Research ◽  
Severe Accidents ◽  
Young Researchers

Forty-three organisations from 22 countries network their capacities of research in SARNET (Severe Accident Research NETwork of excellence) to resolve the most important remaining uncertainties and safety issues on severe accidents in existing and future water-cooled nuclear power plants (NPP). After a first project in the 6th Framework Programme (FP6) of the European Commission, the SARNET2 project, coordinated by IRSN, started in April 2009 for 4 years in the FP7 frame. After 2,5 years, some main outcomes of joint research (modelling and experiments) by the network members on the highest priority issues are presented: in-vessel degraded core coolability, molten-corium-concrete-interaction, containment phenomena (water spray, hydrogen combustion…), source term issues (mainly iodine behaviour). The ASTEC integral computer code, jointly developed by IRSN and GRS to predict the NPP SA behaviour, capitalizes in terms of models the knowledge produced in the network: a few validation results are presented. For dissemination of knowledge, an educational 1-week course was organized for young researchers or students in January 2011, and a two-day course is planned mid-2012 for senior staff. Mobility of young researchers or students between the European partners is being promoted. The ERMSAR conference is becoming the major worldwide conference on SA research.

scholarly journals ANALYTICAL SIMULATION OF A PWR USING MELCOR

2021 ◽  
Vol 8 (3A) ◽  
Author(s):  
Maritza Rodríguez Gual ◽  
Nathalia N. Araújo ◽  
Marcos C. Maturana
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cooling System ◽  
Nuclear Regulatory Commission ◽  
Severe Accident ◽  
Normal Operation ◽  
Detailed Knowledge ◽  
Pressurized Water ◽  
Severe Accidents ◽  
Reference Plant

After the two most significant nuclear accidents in history – the Chernobyl Reactor Four explosion in Ukraine(1986) and the Fukushima Daiichi accident in Japan (2011) –, the Final Safety Analysis Report (FSAR) included a new chapter (19) dedicated to the Probabilistic Safety Assessment (PSA) and Severe Accident Analysis (SAA), covering accidents with core melting. FSAR is the most important document for licensing of siting, construction, commissioning and operation of a nuclear power plant. In the USA, the elaboration of the FSAR chapter 19 is according to the review and acceptance criteria described in the NUREG-0800 and U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.200. The same approach is being adopted in Brazil by National Nuclear Energy Commission (CNEN). Therefore, the FSAR elaboration requires a detailed knowledge of severe accident phenomena and an analysis of the design vulnerabilities to the severe accidents, as provided in a PSA – e.g., the identification of the initiating events involving significant Core Damage Frequency (CDF) are made in the PSA Level 1. As part of the design and certification activities of a plant of reference, the Laboratory of Risk Analysis, Evaluating and Management (LabRisco), located in the University of São Paulo (USP), Brazil, has been preparing a group of specialists to model the progression of severe accidents in Pressurized Water Reactors (PWR), to support the CNEN regulatory expectation – since Brazilian Nuclear Power Plants (NPP), i.e., Angra 1, 2 and 3, have PWR type, the efforts of the CNEN are concentrated on accidents at this type of reactor. The initial investigation objectives were on completing the detailed input data for a PWR cooling system model using the U.S. NRC MELCOR 2.2 code, and on the study of the reference plant equipment behavior – by comparing this model results and the reference plant normal operation main parameters, as modeled with RELAP5/MOD2 code.

scholarly journals Safety assessments and severe accidents, impact of external events on nuclear power plants and on mitigation strategies

2020 ◽  
Vol 6 ◽  
pp. 39
Author(s):  
Jean-Pierre Van Dorsselaere ◽  
Ahmed Bentaib ◽  
Thierry Albiol ◽  
Florian Fichot ◽  
Alexei Miassoedov ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Mitigation Strategies ◽  
Severe Accident ◽  
Severe Accidents ◽  
Accident Management ◽  
External Events ◽  
Safety Assessments ◽  
The University

The Fukushima-Daiichi accidents in 2011 underlined the importance of severe accident management (SAM), including external events, in nuclear power plants (NPP) and the need of implementing efficient mitigation strategies. To this end, the Euratom work programmes for 2012 and 2013 was focused on nuclear safety, in particular on the management of a possible severe accident at the European level. Relying upon the outcomes of the successful Euratom SARNET and SARNET2 projects, new projects were launched addressing the highest priority issues, aimed at reducing the uncertainties still affecting the main phenomena. Among them, PASSAM and IVMR project led by IRSN, ALISA and SAFEST projects led by KIT, CESAM led by GRS and sCO2-HeRO lead by the University of Duisburg-Essen. The aim of the present paper is to give an overview on the main outcomes of these projects.

scholarly journals Development of Instrument Transmitter Protecting Device against High-Temperature Condition during Severe Accidents

2014 ◽  
Vol 2014 ◽  
pp. 1-7 ◽  
Author(s):  
Min Yoo ◽  
Sung Min Shin ◽  
Hyun Gook Kang
Keyword(s):  
Heat Transfer ◽  
High Temperature ◽  
Nuclear Power ◽  
Power Plants ◽  
Severe Accident ◽  
Transfer Model ◽  
Potential Threat ◽  
Severe Accidents ◽  
Development Of Instrument ◽  
Novel Concept

Reliable information through instrumentation systems is essential in mitigating severe accidents such as the one that occurred at the Fukushima Daiichi nuclear power plant. There are five elements which might pose a potential threat to the reliability of parameter detection at nuclear power plants during a severe accident: high temperature, high pressure, high humidity, high radiation, and missiles generated during the evolution of a severe accident. Of these, high temperature apparently poses the most serious threat, since thin shielding can get rid of pressure, humidity, radiation (specifically, alpha and beta radiations), and missile effects. In view of this fact, our study focused on designing an instrument transmitter protecting device that can eliminate the high-temperature effect on transmitters to maintain their functional integrity. We present herein a novel concept for designing such a device in terms of heat transfer model that takes into account various heat transfer mechanisms associated with the device.

The Formulation of Material Characteristics of Austenitic Stainless Steels at Extremely High Temperature

2017 ◽  
Author(s):  
Kenta Shimomura ◽  
Takashi Onizawa ◽  
Shoichi Kato ◽  
Masanori Ando ◽  
Takashi Wakai
Keyword(s):  
High Temperature ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Structural Integrity ◽  
Austenitic Stainless Steels ◽  
Severe Accident ◽  
Integrity Assessment ◽  
Severe Accidents ◽  
Material Characteristics

This paper describes the formulation of material characteristics of austenitic stainless steels at extremely high temperature which meets in some kinds of severe accidents of nuclear power plants. After the severe accident in Fukushima dai-ichi nuclear power plants, it has been supposed to be very important not only to prevent the occurrence of abnormal conditions, i.e. from the first to the third layer safety, but also to prevent the expansion of the accident conditions, i.e. the fourth layer safety[1] [2]. In order to evaluate the structural integrity under the severe accident condition, material characteristics which can be used in the numerical analyses, such as finite element analysis, were required [3] [4]. However, there were no material characteristics applicable to the structural integrity assessment at extremely high temperature. Therefore, a series of tensile and creep tests was performed for austenitic stainless at extremely high temperature which meets in some kinds of severe accidents of nuclear power plants, namely up to 1000 °C. Based on the acquired data from the tests, monotonic stress-strain equation and creep rupture equation applicable to the structural analysis at extremely high temperature, up to 1000 °C were formulated. As a result, these formulae make it possible to conduct the structural integrity assessment using numerical analysis techniques, such as finite element method.

Overview of Revised Level 2 PRA Standard in Japan

2016 ◽  
Author(s):  
Koichi Nakamura ◽  
Yoshiyuki Narumiya ◽  
Yutaka Abe
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Technical Committee ◽  
Lessons Learned ◽  
Severe Accident ◽  
Severe Accidents ◽  
New Findings ◽  
Accident Research ◽  
Level 2

In this paper, we introduce the overview of the standard for Procedure of Level 2 Probabilistic Risk Assessment (PRA) for nuclear power plants established and issued by the Atomic Energy Society of Japan (AESJ). The first edition of the standard was published in 2008 through the discussions at the Level 2 Subcommittee under the Risk Technical Committee of the Standards Committee. As an enforcement standard based on the PRA procedure, the standard specifies the requirements which should have the PRA dealing with incidents resulting from internal events at nuclear power plants during power operation, and the concrete methods of meeting it. This new version standard is regular revision. In revising the 2008 version standard, we updated various requirements to reflect advancements in Level 2 PRA techniques based on new technological findings after the publication of the previous standard and to improve the quality and transparency of this standard. In particular, the lessons learned and new findings from the severe accidents of Fukushima Dai-ichi nuclear power plants, which occurred on March 11 of 2011, were significant. The reason was that three cores were melted down and large amounts of FP were released in the accidents. We investigated the latest documents relevant to severe accident research, and the measure against a severe accident established after the severe accidents of Fukushima Dai-ichi nuclear power plants. Furthermore, we extracted the matter, which should be reflected by comparison with international standard for PRA, ASME/ANS standard and IAEA SSG-4. Here, we introduce the outline and the feature of the AESJ standard for level 2 PRA. We also introduce the future renewal plan of the standard including the extension of the scope for external event, such as an earthquake and tsunami.

Discussing the Modification of Dedicated Lines Used for Severe Accident Depressurization

2013 ◽  
Author(s):  
Bo Zhang ◽  
Jing Liu ◽  
Xichao Liu ◽  
Yafu Gao ◽  
Aicheng Gong
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Second Generation ◽  
Numerical Results ◽  
Severe Accident ◽  
Excessive Heat ◽  
Severe Accidents ◽  
Present Structure

A barrier that limits the development of the nuclear power plants is the problem of the depressurization for severe accidents. The present work addresses this issue by adding the dedicated severe accident depressurization system that is placed on the top of the pressurizer. This design enables the excessive heat to be directly pumped into the atmosphere to lower the pressure in the containment. The theoretical and numerical results indicated that the addition of the dedicated severe accident depressurization system does not affect the performance of the second generation of PWRs. More importantly, this design not only significantly reduces the LERF value but also facilitates the modification of the present structure.

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