Analysis of the source term formation in a severe accident initiated by end fitting failure in CANDU type reactors
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M. Constantin
Abstract
CANDU type reactors have some peculiarities in initiation and progression of severe accidents. In the present paper one of the specificities – the End Fitting Failure accident – is analysed from the point of view of source term formation. The accident is initiated by a failure of the re-fueling machine. Fuel bundles are ejected in the re-fuelling machine room and fuel elements suffer a significant fragmentation by mechanical impact and by the rapid increase of the temperature. A direct transfer of the fission products occurs directly to the containment. The source term in the containment and also the source term to the environment is calculated supposing an open venting communication to the external atmosphere. The simulation is performed by using the ASTEC code in coupled calculation CPA-IODE-ISODOPE-DOSE option. The evolution of the distributions for the most important released fission products is presented for different regions and for different hosts. The most important factors of influence on the source term formation are identified and discussed.
Kurzfassung
CANDU-Reaktoren haben einige Besonderheiten beim Einsetzen und beim Ablauf schwerer Störfälle. In der vorliegenden Arbeit wird eine dieser Besonderheiten, der Störfall durch Probleme mit den Endstücken, analysiert in Bezug auf die Quelltermbildung. Der Störfall wird eingeleitet durch einen Defekt der Lademaschine. Brennelemente werden in den Maschinenraum geschleudert und erleiden durch mechanische Einwirkungen und durch den schnellen Temperaturanstieg erhebliche Beschädigungen. Der Eintritt von Spaltprodukten direkt in das Containment folgt. Der Quellterm im Containment und der Quellterm in der Umgebung wurden unter Annahme einer Freisetzung von Spaltprodukten berechnet. Die Simulation wurde durchführt mit Hilfe des ASTEC-Codes in Verbindung mit der CPA-IODE-ISODOPE-DOSE-Option. Die Entwicklung der Verteilungen der wichtigsten freigesetzten Spaltprodukte wird für verschiedene Regionen und verschiedene Wirte vorgestellt. Die wichtigsten Faktoren, die die Quelltermbildung beeinflussen, werden identifiziert und diskutiert.
References
1 Mathew, P. Mani; et al.: Severe Core Damage Accident Analysis for a CANDU Plant. OECD-ISAM, (2009)Suche in Google Scholar
2 Constantin, M.: CANDU Model Adaptations. FP6-SARNET, 3rd ASTEC Users’ Club Meeting, Aix-en-Provence, April 2008Suche in Google Scholar
3 http://www.candu.com/en/home/candureactors/candu6/default.aspxSuche in Google Scholar
4 Allelein, H.-J.; et al.: European Validation of the Integral Code ASTEC (EVITA) First experience in validation and plant sequence calculation. Nuclear Engineering and Design, 235 (2005) 285–30810.1016/j.nucengdes.2004.08.051Suche in Google Scholar
5 Klein-Heßling, W.; Schwinges, W. B.: CPA Module Program Reference Manual, Rev0. ASTEC-V0/DOC/01-34, TECHNISCHE NOTIZ ASTEC 01/06, (1998)Suche in Google Scholar
6 Cantrel, L.: ASTEC V2.0 – DOSE module: description of physical modelling, Rev0. Report ASTECV2/DOC/09-14, (2009)Suche in Google Scholar
7 Jacq, F.: ASTECV0 – ISODOP code: Isotope treatment. Report ASTEC-V0/DOC/00-10Suche in Google Scholar
8 Bosland, L.: IODE module: iodine and ruthenium behavior in the containment. Report ASTEC-V2/DOC/09-13, (2009)Suche in Google Scholar
9 Gauld, I. C.; et al.: ORIGEN-S, NUREG/CR-200, Rev7, Volume 2, Section F7, ORNL/NUREG/CSD-2/V2/R7, (2002)Suche in Google Scholar
10 Oh, D. J.; et al.: Fission product release assessment for end fitting failure in CANDU reactor loaded with CANFLEX-NU fuel bundles. Proceedings of the Korean Nuclear Society Autumn Meeting, vol. I, October 24–25, Taegu, Korea, (1997), p 797.Suche in Google Scholar
© 2014, Carl Hanser Verlag, München
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Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Technical Contributions/Fachbeiträge
- A model-based approach to operational event groups ranking
- Application of the method for uncertainty and sensitivity evaluation to results of PWR LBLOCA analysis calculated with the code ATHLET. Part 2: sensitivity analysis
- Statistical Large-Break LOCA analysis for PWRs with combined ECC injection
- Analysis of the CSF model for simulated loss of coolant accident conditions
- Performance of two fluid model based numeric tool with pressure and enthalpy as independent variables for simulation of two phase flow in axial and radial direction
- Pre-decommissioning complex engineering and radiation inspection of the WWR-M reactor
- Analysis of the source term formation in a severe accident initiated by end fitting failure in CANDU type reactors
- Neutron flux investigation on certain alternative fluids in a hybrid system by using MCNPX Monte Carlo transport code
- Analytical investigation of lignite and its ash samples taken from the Afşin-Elbistan coal basin in Turkey
- Explicit formulation of a nodal transport method for discrete ordinates calculations in two-dimensional fixed-source problems
- T1 and U1 approximations to neutron transport equation in one-dimensional spherical geometry