Analysis of the Damage Domains of SBO Sequences With RCP Passive Thermal Shutdown

Author(s):  
C. Queral ◽  
L. Mena-Rosell ◽  
G. Jimenez ◽  
M. Sánchez-Perea ◽  
J. Hortal ◽  
...  

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.

Author(s):  
C. Queral ◽  
L. Mena-Rosell ◽  
G. Jiménez Varas ◽  
M. Sánchez-Perea ◽  
J. Hortal ◽  
...  

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 (recovery of AC). 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. Several damage limits have been taken into account within the analysis: 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 results show the capability and necessity of the ISA methodology, or similar, in order to obtain accurate results that take into account time uncertainties.


Author(s):  
J. Gonzalez-Cadelo ◽  
C. Queral ◽  
J. Montero-Mayorga

The Integrated Safety Assessment (ISA) methodology, developed by the Spanish Nuclear Regulatory Body, consists of a dynamic methodology of probabilistic safety assessment. It has been applied to a thermal-hydraulic analysis of small-break and medium-break LOCA sequences without HPSI, for different break locations, in a three-loop PWR. ISA methodology allows to obtain the damage domain, defined as the region of the space of uncertain parameters where the damage condition is exceeded, for each sequence of interest. In this work, damage domain relates two uncertain parameters (starting time of secondary-side depressurization and break size) to damage exceedance condition (PCT > 1477 K). Several damage domains have been obtained, each one for each break location of interest (cold leg, hot leg, vessel upper head and vessel lower head). Simulations have been performed with TRACE v5.0 patch 1 code, and the results show the capability and convenience of ISA methodology, in order to obtain accurate results that take into account time uncertainties.


Author(s):  
C. Queral ◽  
J. Mula ◽  
J. Gómez-Magán ◽  
J. Gil ◽  
I. Fernández ◽  
...  

As part of the collaboration between Universidad Politécnica de Madrid (UPM), Indizen Technologies and the Spanish Nuclear Safety Council (CSN), the Integrated Safety Assessment (ISA) methodology, developed by the Spanish Nuclear Safety Council (CSN), has been applied to a thermo-hydraulical analysis of Total Loss of Feed Water sequences in a Westinghouse 3-loop PWR. The ISA methodology allows among others obtaining the Damage Domain, i.e., the region of the space of uncertain parametere where certain damage limit is exceeded. Apart of typical uncertain parameters for the physical model, ISA considers operator actuations times (like RCP trip, AFW recovery and begining of Feed & Bleed…) and /or stochastic uncertain phennomena. Given a particular safety limit, several data of every sequence are necessary in order to obtain the damage exceedance frequency of that limit. These data are obtained from the results of the simulations of transients performed with MAAP code. 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.


2017 ◽  
Vol 2017 ◽  
pp. 1-16 ◽  
Author(s):  
Siniša Šadek ◽  
Davor Grgić ◽  
Zdenko Šimić

The integrity of the containment will be challenged during a severe accident due to pressurization caused by the accumulation of steam and other gases and possible ignition of hydrogen and carbon monoxide. Installation of a passive filtered venting system and passive autocatalytic recombiners allows control of the pressure, radioactive releases, and concentration of flammable gases. Thermal hydraulic analysis of the containment equipped with dedicated passive safety systems after a hypothetical station blackout event is performed for a two-loop pressurized water reactor NPP with three integral severe accident codes: ASTEC, MELCOR, and MAAP. MELCOR and MAAP are two major US codes for severe accident analyses, and the ASTEC code is the European code, joint property of Institut de Radioprotection et de Sûreté Nucléaire (IRSN, France) and Gesellschaft für Anlagen und Reaktorsicherheit (GRS, Germany). Codes’ overall characteristics, physics models, and the analysis results are compared herein. Despite considerable differences between the codes’ modelling features, the general trends of the NPP behaviour are found to be similar, although discrepancies related to simulation of the processes in the containment cavity are also observed and discussed in the paper.


2014 ◽  
Vol 986-987 ◽  
pp. 334-338
Author(s):  
Jia Xu Zuo ◽  
Xu Xu ◽  
Jian She Chai ◽  
Chun Ming Zhang ◽  
Jian Ping Jing

The nuclear safety became the more important after the Fukushima accident. The development of nuclear safety culture is one of the most effective methods to improve the safety. Also the assessment shows the level of the nuclear safety culture. The nuclear safety assessment model is described. The assessment model was described from its time and space, and it was also discussed from macro and micro levels. The internal and external environment parameters are considered and the assessment of different levels of safety culture is described. According to established principles of nuclear safety culture, the selection rules of the different levels evaluation indicators of nuclear safety culture are discussed too. And it is pointed that the assessment of nuclear safety culture is a long time project and an integrated system.


Equipment ◽  
2006 ◽  
Author(s):  
D. Sujish ◽  
C. Meikandamurthy ◽  
T. R. Ellappan ◽  
M. Rajan ◽  
G. Vaidyanathan

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