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Farmer, M.T.; Corradini, M.; Gauntt, R.O.
In-vessel Melt Retention and Ex-vessel Corium Cooling. Summary of a Technical Meeting. Supplementary Files
In-vessel Melt Retention and Ex-vessel Corium Cooling. Summary of a Technical Meeting. Supplementary Files
AbstractAbstract
[en] Goals: – Develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structures and components (SSCs) as they age in environments; – Apply this knowledge to develop and demonstrate methods and technologies that support safe and economical long-term operation of existing reactors; – Research new technologies that enhance plant performance, economics, and safety.
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International Atomic Energy Agency, Safety Assessment Section, Vienna (Austria); vp; ISBN 978-92-0-106320-5;
; ISSN 1011-4289;
; May 2020; 24 p; Technical Meeting on Phenomenology and Technologies Relevant to In-Vessel Melt Retention and Ex-Vessel Corium Cooling; Shanghai (China); 17-21 Oct 2016; Also available on-line: https://www.iaea.org/publications/13576/in-vessel-melt-retention-and-ex-vessel-corium-cooling; Enquiries should be addressed to IAEA, Marketing and Sales Unit, Publishing Section, E-mail: sales.publications@iaea.org; Web site: http://www.iaea.org/books; 6 refs.


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AbstractAbstract
[en] This article presents the INES scale and shows how the gravity of a reactor accident is assessed. The INES scale is made of 7 levels: 1 - anomaly, 2 - incident, 3 - severe incident, 4 - accident with local consequences, 5 - accident with extended consequences, 6 - severe accident, and 7 - major accident. Each level is illustrated by a real accident. Only 2 accidents ranked at the highest level (7): Tchernobyl (Ukraine - 1986) and Fukushima Daiichi (Japan - 2011), they were characterized by a complete meltdown of the reactor core. A partial meltdown occurred for the following reactor accidents: Three Mile Island (USA - 1979), Saint Laurent des Eaux (France - 1969 and 1980), Lucens (Switzerland - 1969), Chapelcross (UK - 1967), Windscale (UK - 1957), EBR-1 (US - 1955) and NRX (Canada - 1952). A quick methodology to assess the gravity of an accident is to estimate the quantity of equivalent radioactivity released in the atmosphere, for instance the equivalent of a few thousands Tera decay/second of iodine 131 corresponds to the level 5. The number of people irradiated is also an important parameter as well as the radiation exposure, for instance for an individual dose over 200 mSv, involving 100 people or more corresponds to a level 5 event, 10 people or more to a level 4 event and less than 10 people to a level 3 event. (A.C.)
Original Title
Pourquoi le fonctionnement d'un reacteur nucleaire peut-il etre dangereux?
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2 refs.
Record Type
Journal Article
Journal
Rayonnements Ionisants, Techniques de Mesures et de Protection; ISSN 0397-9210;
; CODEN RITMB3; (no.2); p. 7-13

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Prasad, Mahendra; Vinod, Gopika; Gaikwad, Avinash J.; Ramarao, A., E-mail: mprasad@barc.gov.in
AbstractAbstract
[en] Highlights: • Core Damage Frequency and Large Early Release Frequency. • Multi –Unit Risk Metrics. • Aggregation of CDF of NPP through Mean Values. • Aggregation of CDF of NPP as Random Variable. - Abstract: The nuclear generating sites around the world are mostly twin unit and multi-unit sites. The PSA risk metrics Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) currently are based on per reactor reference. The models for level 1 and level 2 PSA have been developed based on single unit. The Fukushima accident has spawned the need to address the issue of site base risk metrics, Site Core Damage Frequency (SCDF) and Site Large Early Release Frequency (SLERF), on the site years rather than reactor years. It is required to develop a holistic framework for risk assessment of a site. In the context of current study, the holistic framework refers to integration of risk from all units, dependencies due to external events and operation time of individual units. There is currently no general consensus on how to arrive at site-specific risk metrics. Some documents provide suggestions for site CDF and site LERF. This paper proposes a new method of aggregation of risk metric from the consideration of operating time of individual units under certain assumptions with a purpose to provide a new conceptual aspect for multi-unit PSA. The result of a case study on hypothetical data shows that site level CDF is not sum of CDF of all units but around 18% higher than unit level CDF. When the CDF is considered to be a random variable then, the new methodology produces site CDF as 50% higher than single unit CDF. These two approaches have been detailed in the paper. For a general data set of CDF for individual units, site CDF would more than individual unit CDF however, it would not be multiples of a single unit value.
Primary Subject
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S0149197016302785; Available from http://dx.doi.org/10.1016/j.pnucene.2016.12.007; Copyright (c) 2017 Elsevier Science B.V., Amsterdam, The Netherlands, All rights reserved.; Country of input: International Atomic Energy Agency (IAEA)
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AbstractAbstract
[en] The objectives of DEC analysis can be summarized as follows: (c) DEC analyses are used to confirm that features, credited for DEC, have the requested performances to meet their relevant safety objectives, in addition to DBA studies: (i) No core melting such as to ensure prevention of core melting for DEC-A. (ii) Protection actions that are limited in terms of lengths of time and areas of application need to be sufficient to protect people and the environment, this meaning limitation of radiological consequences in DEC-B. (d) In particular, the demonstration needs to meet the following requirements: (i) DEC-A conditions are considered in emergency operating procedures (with other specific procedures or guidelines when applicable). DEC-B conditions need to be considered by SAM guidelines (with other specific procedures or guidelines when applicable). (ii) Any equipment credited in a DEC analysis needs to be adequately qualified to perform its safety functions in the environmental conditions resulting from this DEC situation. (iii) SSCs that are necessary to meet the safety requirements in DEC analyses need to be considered as items important to safety and to be safety classified accordingly.
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International Atomic Energy Agency, Safety Assessment Section, Vienna (Austria); 171 p; ISBN 978-92-0-133921-8;
; ISSN 1011-4289;
; Oct 2021; p. 83-91; Also available on-line: https://www.iaea.org/publications/14976/current-approaches-to-the-analysis-of-design-extension-conditions-with-core-melting-for-new-nuclear-power-plants; Enquiries should be addressed to IAEA, Marketing and Sales Unit, Publishing Section, E-mail: sales.publications@iaea.org; Web site: http://www.iaea.org/books; 2 refs.


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AbstractAbstract
[en] The phenomenon and mechanism of FCI (Fuel Coolant Interaction) has been widely studied around the world in the past few decades. A series of experiments were performed and several FCI models were developed on the basis of these experiments. However, there are still large uncertainties in the models of FCI and limitations to predict FCI process, especially the reactor scale process. To study the mechanism of FCI, a new FCI experimental facility was designed and further experiments were performed by Shanghai Jiao Tong University, China. The photography of FCI process were obtained by 2 high-speed cameras recording from 2 different directions vertical to each other. Water level changed can also be got from images of FCI process. Pressure peak produced by intense interaction and temperature of coolant are recorded. To discuss the influence of different factors for FCI, numbers of variables are considered in these experiments, including jet material, melt temperature, coolant type, coolant subcooled temperature, release heights, break size and interaction pool size. This paper focuses on the FCI responses for different melt temperatures. Tin, with the melt point of 231.9 C. degrees, was chosen as the melt material since it is possible to acquire a large temperature range of melt from 400 to 1300 C. degrees. The phenomenon responses of FCI process and the particle size responses for different melt temperature are discussed in the paper. (authors)
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Societe Francaise d'Energie Nucleaire (SFEN), 75 - Paris (France); 3390 p; 2015; p. 1391-1400; ICAPP 2015: Nuclear Innovations for a low-carbon future; Nice (France); 3-6 May 2015; Available (USB stick) from: SFEN, 103 rue Reaumur, 75002 Paris (France); 10 refs.; This record replaces 48079354
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Katsaounis, A.; Fuerst, H.D.; Weber, M.
Annual meeting on nuclear technology '83 - Reactor conference 1983
Annual meeting on nuclear technology '83 - Reactor conference 1983
AbstractAbstract
No abstract available
Original Title
Messung und Berechnung des einphasigen Druckverlustes in einer T-Rohrverbindung; Investigations with respect to the atws reactor accident
Primary Subject
Source
Deutsches Atomforum e.V., Bonn (Germany, F.R.); Kerntechnische Gesellschaft e.V., Bonn (Germany, F.R.); 900 p; 1983; p. 65-68; Fachinformationszentrum Energie, Physik, Mathematik; Eggenstein-Leopoldshafen (Germany, F.R.); Annual meeting on nuclear technology '83 - Reactor conference 1983; Berlin (Germany, F.R.); 14 - 16 Jun 1983; Published in summary form only.
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Book
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Tarasov, A.E.
Conference of young specialists Innovations in nuclear power engineering. Book of abstracts
Conference of young specialists Innovations in nuclear power engineering. Book of abstracts
AbstractAbstract
No abstract available
Original Title
Dinamicheskoe vozdejstvie volny termicheskoj detonatsii na pregradu
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Source
Aktsionernoe Obshchestvo Ordena Lenina Nauchno-Issledovatel'skij i Konstruktorskij Inst. Ehnergotekhniki imeni N.A. Dollezhalya, Moscow (Russian Federation); 91 p; ISBN 978-5-98706-098-8;
; 2015; p. 68; Conference of young specialists Innovations in nuclear power engineering; Konferentsiya molodykh spetsialistov Innovatsii v atomnoj ehnergetike; Moscow (Russian Federation); 25-26 Nov 2015

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AbstractAbstract
[en] This publication results from a technical meeting on phenomenology and technologies relevant to in-vessel melt retention (IVMR) and ex-vessel corium cooling (EVCC). The purpose of the publication is to capture the state of knowledge, at the time of that meeting, related to phenomenology and technologies as well as the challenges and pending issues relevant to IVMR and EVCC for water cooled reactors by summarizing the information provided by the meeting participants in a form useful to practitioners in Member States.
Primary Subject
Source
May 2020; vp; Technical Meeting on Phenomenology and Technologies Relevant to In-Vessel Melt Retention and Ex-Vessel Corium Cooling; Shanghai (China); 17-21 Oct 2016; ISBN 978-92-0-106320-5;
; ISSN 1011-4289;
; Also available on-line: https://www.iaea.org/publications/13576/in-vessel-melt-retention-and-ex-vessel-corium-cooling; Enquiries should be addressed to IAEA, Marketing and Sales Unit, Publishing Section, E-mail: sales.publications@iaea.org; Web site: http://www.iaea.org/books; Refs., figs., tabs.


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AbstractAbstract
[en] The forecast of atmospheric diffusion described here allows statements to be made about the size and location of the area in which maximum radiation burdens have to be expected from emission after an accident. If the source strength can be estimated from the course of the accident, the value of the maximum burden to be expected can be predicted in addition. Based on these statements, first effective actions can be taken to protect the population in case of serious hypothetical accidents. (orig.)
[de]
Die hier vorgestellte Ausbreitungsprognose erlaubt Aussagen ueber Groesse und Lage des Gebietes, in dem die hoechsten Strahlenbelastungen nach einer unfallbedingten Emission zu erwarten sind. Ist aufgrund des Unfallverlaufs eine Abschaetzung der Quellstaerke moeglich, so kann zusaetzlich die Hoehe der zu erwartenden maximalen Belastungen vorausgesagt werden. Mit diesen Angaben koennen bei grossen hypothetischen Unfaellen erste wirksame Massnahmen zum Schutze der Bevoelkerung eingeleitet werden. (orig.)Original Title
Prognose der Schadstoffausbreitung in der Atmosphaere nach Unfaellen in kerntechnischen Anlagen
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4 figs.; 2 tabs.; 15 refs.
Record Type
Journal Article
Journal
Atomkernenergie; v. 29(4); p. 282-286
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Lee, Yong Bum; Kwon, Y. M.; Kim, E. K.; Suk, S. D.; Chang, W. P.; Jeong, H. Y.; Ha, K. S.
Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
AbstractAbstract
[en] This report contains the safety analyses of the KALIMER-600 conceptual design which KAERI has been developing under the Long-term Nuclear R and D Program. The analyses have been performed reflecting the design developments during the second year of the 4th design phase in the program. The specific presentations are the key design features with the safety principles for achieving the safety objectives, the event categorization and safety criteria, and results on the safety analyses for the DBAs and ATWS events, the containment performance, and the channel blockages. The safety analyses for both the DBAs and ATWS events have been performed using SSC-K version 1.3., and the results have shown the fulfillment of the safety criteria for DBAs with conservative assumptions. The safety margins as well as the inherent safety also have been confirmed for the ATWS events. For the containment performance analysis, ORIGEN-2.1 and CONTAIN-LMR have been used. In results, the structural integrity has been acceptable and the evaluated exposure dose rate has been complied with 10 CFR 100 and PAG limits. The analysis results for flow blockages of 6-subchannels, 24-subchannels, and 54- subchannels with the MATRA-LMR-FB code, have assured the integrity of subassemblies
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Feb 2007; 151 p; Also available from KINS; 23 refs, 65 figs, 24 tabs
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