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

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

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. 2016 The Authors. Published by Elsevier Ltd. This is an open access article under the CC BY-NC-NDlicenseJRC.F.5-Nuclear Reactor Safety Assessmen

ASTEC-Na code: thermal-hydraulic model validation and benchmarking with other codes

This paper describes the work performed within the WP2.1 of the JASMIN project to validate the thermal-hydraulic models of ASTEC-Na code. The experiments used for validation purpose have been: BI1, E8 and EFM1, carried out in the CABRI reactor, BE+3, APL1 and APL3, carried out in the SCARABEE reactor and N02, conducted in the KNS facility. Moreover, the simulation of CABRI and SCARABEE tests has been the object of a code benchmark exercise. Thermal-hydraulic codes such as CATHARE, RELAP5-Na and RELAP5-3D as well as severe accident codes such as SIMMER-III and SAS-SFR have been used as benchmark to assess the ASTEC-Na performances. ASTEC-Na has been also successfully applied (pretest calculations) to assess the performances of the KASOLA sodium loop and verify its response under different operating conditions. Also in this case, the ASTEC-Na results have been bench marked with other thermal-hydraulic system codes (CATHARE, RELAP5-3D and RELAP5-Na).JRC.G.I.4-Nuclear Reactor Safety and Emergency Preparednes

ASTEC-Na code Thermal-hydraulic model validation and benchmarking with other codes

International audienceThis paper describes the work performed within the WP2.1 of the JASMIN project to validate the thermal-hydraulic models of ASTEC-Na code. The experiments used for validation purposes have been BI1, E8 and EFM1, carried out in the CABRI reactor; BE + 3, APL1 and APL3, carried out in the SCARABEE reactor and N02, conducted in the KNS facility. ASTEC-Na has been also successfully applied (pretest calculations) to assess the performances of the KASOLA sodium loop and to verify its response under different operating conditions. Finally, a PHENIX natural circulation test has also been performed with ASTEC-Na. Besides, these experimental tests have been used as a code benchmarking exercise where thermal-hydraulic codes (CATHARE, RELAP5-Na and RELAP5-3D) and severe accident codes (SIMMER-III and SAS-SFR) have been compared with ASTEC-Na. © 2017 Elsevier Lt

Development and assessment of ASTEC-Na fuel pin thermo-mechanical models performed in the European JASMIN project

International audienceASTEC-Na is a computer code system which evaluates protected and unprotected accidents in Sodium-cooled Fast Reactors throughout the Initiation Phase. The fuel pin behavior models implemented in ASTEC-Na simulate the essential aspects of the thermal and mechanical behavior of SFR fuel pins in nominal and accidental conditions. Besides the improvement of existing ASTEC-Na models, a specific cladding mechanical model and an in-pin fuel relocation model have been developed during the JASMIN project, where the latter has been already implemented in the code. ASTEC-Na fuel pin behavior models are described in this paper and their transient predictions for four CABRI transient tests are compared to available experimental data as well as to SAS-SFR and SIMMER-III code calculations. Conclusions on the overall performance of ASTEC-Na fuel pin models are also presented. © 2017 Elsevier Lt

Development and Assessment of ASTEC-Na Fuel Pin Thermo-Mechanical Models Performed in the European JASMIN Project

ASTEC-Na is a computer code system which evaluates protected and unprotected accidents in Sodium-cooled Fast Reactors throughout the Initiation Phase. The fuel pin behavior models implemented in ASTEC-Na sum up the essential aspects of the thermal and mechanical behavior of SFR fuel pins undergoing an accidental transient. A specific cladding mechanical model and an in-pin fuel relocation model have been developed all through the project where the later has been already implemented in the code and is already available. ASTEC-Na fuel pin behavior models are described in this paper and their transient predictions for four CABRI transient tests are compared to available experimental data as well as to SAS-SFR and SIMMER-III code calculations. Conclusions on the overall performance of ASTEC-Na fuel pin models are also presented.JRC.G.I.4-Nuclear Reactor Safety and Emergency Preparednes