Analysis of an unmitigated 2-inch cold leg LOCA transient with ASTEC and MELCOR codes

The 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 ASTEC-Na capabilities for simulating a loss of flow CABRI experiment

Abstract This paper presents simulation results of the CABRI BI1 test using the code ASTEC-Na, currently under development, as well as a comparison of the results with available experimental data. The EU-JASMIN project (7th FP of EURATOM) centres on the development and validation of the new severe accident analysis code ASTEC-Na (Accident Source Term Evaluation Code) for sodium-cooled fast reactors whose owner and developer is IRSN. A series of experiments performed in the past (CABRI/SCARABEE experiments) and new experiments to be conducted in the new experimental sodium facility KASOLA have been chosen to validate the developed ASTEC-Na code. One of the in-pile experiments considered for the validation of ASTEC-Na thermal–hydraulic models is the CABRI BI1 test, a pure loss-of-flow transient using a low burnup MOX fuel pin. The experiment resulted in a channel voiding as a result of the flow coast-down leading to clad melting. Only some fuel melting took place. Results from the analysis of this test using SIMMER and SAS-SFR codes are also presented in this work to check their suitability for further code benchmarking purposes

SBO analysis of a generic PWR-900 with ASTEC and MELCOR codes

Abstract After 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

Third Yearly Activity Report

The calculation work performed during the 3rd project year in WP2 as well as the R&D activities carried out in WP3, WP4 and WP5 are described in this report. In addition, the work dedicated to the project management (WP1) as well as to WP6 regarding the dissemination/communication activities and the education/training program (e.g. the follow-up of the mobility program between different organizations in the consortium, training on simulation tools and activities accomplished by PhD/post-doctoral students) is also reported

SIngle and two-phase sodium flow analysis for two TUCOP CABRI tests using the ASTEC-Na code

The development and validation of ASTEC-Na code as a safety code system for severe accident analysis is being performed in the framework of the JASMIN project supported by the European Commission. One of the main tasks of the modelling and validation tasks of this project is devoted to the sodium thermal hydraulic behavior both in single and two-phase regimes. In this paper we present the first ASTEC-Na results of E8 and EFM1 in-pile tests conducted in the CABRI experimental reactor. These two tests are LOF+TOP transients. We present for both CABRI tests a comparison between experimental data and ASTEC-Na results for transient coolant temperatures at different heights, boiling onset time, inlet and outlet flow rate and the evolution of the sodium two-phase front during the boiling time up to TOP onset. Besides, a benchmark is presented comparing ASTEC-Na simulation results with the results of other safety codes such as CATHARE, RELAP5-3D and SAS-SFR

Air oxidation of Zircaloy-4 in the 600-1000 °c temperature range: Modeling for ASTEC code application

cited By 35Progress in the treatment of air oxidation of zirconium in severe accident (SA) codes are required for a reliable analysis of severe accidents involving air ingress. Air oxidation of zirconium can actually lead to accelerated core degradation and increased fission product release, especially for the highly-radiotoxic ruthenium. This paper presents a model to simulate air oxidation kinetics of Zircaloy-4 in the 600-1000 °C temperature range. It is based on available experimental data, including separate-effect experiments performed at IRSN and at Forschungszentrum Karlsruhe. The kinetic transition, named "breakaway", from a diffusion-controlled regime to an accelerated oxidation is taken into account in the modeling via a critical mass gain parameter. The progressive propagation of the locally initiated breakaway is modeled by a linear increase in oxidation rate with time. Finally, when breakaway propagation is completed, the oxidation rate stabilizes and the kinetics is modeled by a linear law. This new modeling is integrated in the severe accident code ASTEC, jointly developed by IRSN and GRS. Model predictions and experimental data from thermogravimetric results show good agreement for different air flow rates and for slow temperature transient conditions. © 2010 Elsevier B.V. All rights reserved

ANALYSIS OF A 2-INCHES GUILLOTINE BREAK OF THE DVI LINE IN AN IRIS-LIKE DESIGN BY USING ASTEC CODE

Today the LWR passive Small Water Reactors (SMR) are one of the key design options for the deployment of nuclear technology considering their inherent safety due to the use of passive mitigation strategy (e.g. use of passive systems, etc). The integral nature and the passive mitigation strategy of these reactors require the analysis and the characterization of the phenomena characteristic both in design and beyond design basis accidents conditions. In the framework of deterministic safety analyses for integral small modular reactor a nodalization of IRIS-like SMR was developed by using the severe accident ASTEC code (study carried out with ASTEC V2, IRSN all rights reserved, [2019]). In particular, the IRIS-like primary/secondary circuit and the safety systems were modeled using CESAR module of ASTEC code, while the containment system was nodalized using CPA module. A 2-inches guillotine break of the Direct Vessel Injection (DVI) line was hypothesized and the effect of the safety systems considered available (ADS, EHRS, PSS, LGMS, EBS) was investigated on the transient progressions. The purpose of this work is to analyze the capability of ASTEC code to simulate the mitigation effect due to activation of the different passive systems and the consequent transient scenario