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Sistema economia RIM
No full textLibro di testo di Economia per il secondo biennio del corso di Relazioni internazional
Analysis of an unmitigated 2-inch cold leg LOCA transient with ASTEC and MELCOR codes
Full text linkThe analyses of postulated severe accident sequences play a key role for the
international nuclear technical scientific community for the study of the effect of possible actions
to prevent significant core degradation and mitigate source term release. To simulate the
complexity of phenomena involved in a severe accident, computational tools, known as severe
accident codes, have been developed in the last decades. In the framework of NUGENIA TA-2
ASCOM project, the analysis of an unmitigated 2-inch cold leg LOCA transient, occurring in a
generic western three-loops PWR-900 MWe, has been carried out with the aim to give some
insights on the modelling capabilities of these tools and to characterize the differences in the
calculations results. The ASTEC V2.2b code (study carried out with ASTEC V2, IRSN all rights
reserved, [2021]), and MELCOR 2.2 code have been used in this code-to-code benchmark
exercise. In the postulated transient, the unavailability of all active injection coolant systems has
been considered and only the injection of accumulators has been assumed as accident mitigation
strategy
Analysis of BDBA sequences in a generic IRIS reactor using ASTEC code
Full text linkIntegral Severe Accident (SA) codes are aimed at providing an exhaustive coverage of all the main phenomena taking place in a core melt accident. Today, these deterministic codes have reached a high level of maturity for the simulation of operating reactors and the nuclear technical community is starting to extend their applicability to advanced reactor designs, as Small Modular Reactor (SMR). In the framework of the NUGENIA TA-2 ASCOM (ASTEC COMmunity) collaborative project, a generic input-deck based on the IRIS design has been developed for the ASTEC code. The generic SMR ASTEC model has been already proved able to simulate the main thermal–hydraulic phenomena driving the passive mitigation of a SBLOCA in Design Basis Accident (DBA) conditions. The same initiator event, regarding the guillotine break of a Direct Vessel Injection (DVI) line, will be assumed for the simulation of beyond-design scenarios by considering the unavailability of selected passive safety systems. The results of the ASTEC simulations of four Beyond Design Basis Accidents (BDBAs) (study carried out with ASTEC V2.2, IRSN all rights reserved, [2021]) will be analyzed and discussed against the reference DBA sequence in the present paper. This study is aimed at proving the first insights about the capability of the ASTEC model of a generic IRIS reactor to be used in BDBA and in SA analyses, if significant core degradation takes place. In addition, it characterizes the role played by each safety system in SMR passive mitigation strategy and give the possibility to characterize the phenomenologies specific of SMR designs
Coupling of ASTEC V2.1 and RASCAL 4.3 Codes to Evaluate the Source Term and the Radiological Consequences of a Loss-of-Cooling Accident at a Spent Fuel Pool
Full text linkThis paper deals with a general methodology to evaluate the Source Term (ST) and the Radiological Consequences (RC) of a hypothetical Severe Accident (SA) at a Fukushima-like Spent Fuel Pool (SFP) by coupling ASTEC 2.1 and RASCAL 4.3 SA and consequence projections (CP) codes, respectively. The methodology consists of the following sequential steps: the ST provided by a prior simulation performed by ASTEC V2.1 code was used as input to RASCAL 4.3 code to make a RC analysis. This approach was developed as a preparatory study for the Management and Uncertainties in Severe Accident (MUSA) H2020 European Project, coordinated by CIEMAT, where the ENEA's Nuclear Installations safety laboratory is committed to performing an analysis on a Fukushima-like SFP with the aim to apply innovative management of SFP accidents (WP6) to mitigate the RC of the accident itself. To perform the RC studies that could have an impact on Italy, a Fukushima-like SFP was assumed located in one of the Italian cross-border NPP sites. The weather data adopted are both standard and real hourly meteorological data taken from more than one geographical location. The results of the RC for 96 h of ST release in a range of 160 km from the emission point are reported in terms of Total Effective Dose Equivalent (TEDE), Thyroid dose, and Cs-137 total ground deposition. The mitigating effect on ST and on RC of the cooling spray system (CSS) actuated with several pH values (i.e., 4, 7, 10) was also investigated
Validation and uncertainty analysis of ASTEC in early degradation phase against QUENCH-06 experiment
Full text linkSevere Accident (SA) integral codes, such as the Accident Source Term Evaluation Code (ASTEC) developed by IRSN, are used to simulate the phenomena occurring during accident progression in Nuclear Power Plants (NPPs) up to the source term evaluation. Code validation against experimental data is fundamental to carry out deterministic safety analysis and apply these codes to NPPs. In addition, in the Best Estimate Plus Uncertainty (BEPU) framework, the quantification of the results uncertainty is needed. In the framework of the IAEA CRP I31033 “Advancing the State-of-Practice in Uncertainty and Sensitivity Methodologies for Severe Accident Analysis in Water-Cooled Reactors”, the QUENCH test-6 experiment, conducted at KIT, has been selected to develop an uncertainty analysis using the ASTEC v2.2b code. The accuracy of the best-estimate ASTEC simulation was evaluated with the Fast Fourier Transform Based Method (FFTBM) against the experimental data. Then, the uncertainty of the code results was quantified by using the probabilistic propagation of input uncertainties method, through the coupling of ASTEC with RAVEN (Risk Analysis and Virtual Environment). Beyond identifying the main sources of uncertainty affecting the simulated test, the outcomes of the work also include some general discussion on the uncertainty propagation in a SA sequence
ASTEC code DBA analysis of a passive mitigation strategy on a generic IRIS SMR
No full textAdvanced small LWRs, are considered as one of the key design options for the development of a safer nuclear technology considering their inherent safety due to the adoption of passive mitigation strategies and lower nominal power. In the framework of deterministic safety analyses for Small Modular Reactors (SMRs), a model of a generic IRIS SMR was developed by using the severe accident code ASTEC (Study carried out with ASTEC V2, IRSN all rights reserved, [2019]). The ASTEC code thermal-hydraulics modules have been used for the reactor modelling, and the nodalization approach has been described in the present work. The objective of the paper is to analyze the code capability and the consequent applicability to model an integral-type reactor and to simulate the complex thermal–hydraulic phenomena occurring in a passive mitigation strategy. The analysis is based on 2-inches guillotine Direct Vessel Injection (DVI) line break transients
Analysis of BDBA sequences in a generic IRIS reactor using ASTEC code
No full textIntegral Severe Accident (SA) codes are aimed at providing an exhaustive coverage of all the main phenomena taking place in a core melt accident. Today, these deterministic codes have reached a high level of maturity for the simulation of operating reactors and the nuclear technical community is starting to extend their applicability to advanced reactor designs, as Small Modular Reactor (SMR). In the framework of the NUGENIA TA-2 ASCOM (ASTEC COMmunity) collaborative project, a generic input-deck based on the IRIS design has been developed for the ASTEC code. The generic SMR ASTEC model has been already proved able to simulate the main thermal-hydraulic phenomena driving the passive mitigation of a SBLOCA in Design Basis Accident (DBA) conditions. The same initiator event, regarding the guillotine break of a Direct Vessel Injection (DVI) line, will be assumed for the simulation of beyond-design scenarios by considering the unavailability of selected passive safety systems. The results of the ASTEC simulations of four Beyond Design Basis Accidents (BDBAs) (study carried out with ASTEC V2.2, IRSN all rights reserved, [2021]) will be analyzed and discussed against the reference DBA sequence in the present paper. This study is aimed at proving the first insights about the capability of the ASTEC model of a generic IRIS reactor to be used in BDBA and in SA analyses, if significant core degradation takes place. In addition, it characterizes the role played by each safety system in SMR passive mitigation strategy and give the possibility to characterize the phenomenologies specific of SMR designs
SBO analysis of a generic PWR-900 with ASTEC and MELCOR codes
Full text linkAfter the Fukushima accident, the interest of the public to nuclear safety has growth and the international technical nuclear community has increased his attention in the investigation and the characterization of Severe Accident (SA) scenarios. In order to simulate the different, complex and multi-physical phenomena involved in a SA, computational tools, known as SA codes, have been developed in the last decades. In order to give some insights on the modelling capabilities of these tools and the differences in the calculation results, also related to the user-effect, an analysis of an unmitigated Station Black Out (SBO) occurring in a generic Western three-loops PWR 900 MWe has been carried out by the authors in the framework of the NUGENIA TA-2 ASCOM project. The simulation results of ASTEC code (study carried out with ASTEC V2, IRSN all rights reserved, [2019]), developed by IRSN, and MELCOR 2.2 code, developed by SANDIA for USNRC, have been compared and analyzed. The SBO scenario considered takes into account the intervention of the accumulators as only accident mitigation strategy. Several figures of merits related to the thermal-hydraulic (e.g. primary pressure, cladding temperature, etc.) and to the core degradation (e.g. hydrogen production, etc.) have been considered to describe the accident evolution until the vessel failure, for the two codes comparison
ASTEC code DBA analysis of a passive mitigation strategy on a generic IRIS SMR
No full textAdvanced small LWRs, are considered as one of the key design options for the development of a safer nuclear technology considering their inherent safety due to the adoption of passive mitigation strategies and lower nominal power. In the framework of deterministic safety analyses for Small Modular Reactors (SMRs), a model of a generic IRIS SMR was developed by using the severe accident code ASTEC (Study carried out with ASTEC V2, IRSN all rights reserved, [2019]). The ASTEC code thermal-hydraulics modules have been used for the reactor modelling, and the nodalization approach has been described in the present work. The objective of the paper is to analyze the code capability and the consequent applicability to model an integral-type reactor and to simulate the complex thermal–hydraulic phenomena occurring in a passive mitigation strategy. The analysis is based on 2-inches guillotine Direct Vessel Injection (DVI) line break transients
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