1,721,404 research outputs found
Code assessment and modelling for Design Basis Accident Analysis of the European sodium fast reactor design. Part I: System description, modelling and benchmarking
No full textThe new reactor concepts proposed in the Generation IV International Forum (GIF) are conceived to improve the use of natural resources, reduce the amount of high-level radioactive waste and excel in their reliability and safe operation. Among these novel designs sodium fast reactors (SFRs) stand out due to their technological feasibility as demonstrated in several countries during the last decades. As part of the contribution of EURATOM to GIF the CP-ESFR is a collaborative project with the objective, among others, to perform extensive analysis on safety issues involving renewed SFR demonstrator designs. The verification of computational tools able to simulate the plant behaviour under postulated accidental conditions by code-to-code comparison was identified as a key point to ensure reactor safety. In this line, several organizations employed coupled neutronic and thermal-hydraulic system codes able to simulate complex and specific phenomena involving multi-physics studies adapted to this particular fast reactor technology. In the "Introduction" of this paper the framework of this study is discussed, the second section describes the envisaged plant design and the commonly agreed upon modelling guidelines. The third section presents a comparative analysis of the calculations performed by each organisation applying their models and codes to a common agreed transient with the objective to harmonize the models as well as validating the implementation of all relevant physical phenomena in the different system codes. � 2013 Elsevier B.V. All rights reserved
ICARE/CATHARE Calculation of the QUENCH-11 Experiment in the Frame of the IRSN Participation in the SARNET Benchmark
No full textIn the frame of IRSN-Cadarache participation in the SARNET Benchmark, the QUENCH11 experiment, performed at FZK/Karlsruhe in 2005, has been analyzed using the French ICARE/CATHARE code. The ICARE/CATHARE calculation has been performed according to benchmark specifications and requirements. The code results, presented at the 12th International QUENCH Workshop, Karlsruhe, 24-26 October 2006, have been delivered to benchmark coordinator within the expected deadline (30 October 2006). This report presents the information on ICARE/CATHARE calculation and data files, as well as the calculation results presented in the QUENCH Workshop
Core loss during a severe accident (COLOSS)
No full textThe core loss during a severe accident (COLOSS) project is a 3-year shared-cost action which started in February 2000. The project is concerned with the consequences that core degradation, occurring under severe accident conditions, may have on H2 production, melt generation and the source term. Unresolved in-vessel risk-relevant issues are studied, through a large number of experiments such as (a) UO2 and MOX dissolution by molten zircaloy and burn-up effects, (b) simultaneous dissolution of UO2 and ZrO2 in rod geometry, (c) oxidation of U-O-Zr mixtures, (d) oxidation of pure B4C material and (e) degradation and oxidation of B4C control rods. A parallel effort is devoted to model developments for severe accident (SA) computer codes. These codes are finally used for plant calculations to assess SA code capabilities and to apply results produced in this project to evaluate their consequences on key SA sequences occurring in different plants such as PWR-1300, BWR, VVER-1000, EPR and in the TMI-2 accident. Following significant results have been produced at the mid-term of the project: Several B4C oxidation experiments have improved the understanding of oxidation mechanisms. Preliminary models have been developed and implemented in SA codes. Separate-effects tests (SET) on simultaneous UO2 and ZrO2 dissolution and on U-O-Zr oxidation by steam enabled progress to be made on the understanding and modelling of these interactions. There is experimental evidence that the oxidation of mixtures can contribute significantly to the large H2 production occurring during the reflood of a reactor core under severe accident conditions. Two large-scale tests CODEX-B4C and QUENCH-07 have been carried out with a central B4C control rod. The B4C effects on VVER and PWR core degradation and on the related gas production have been evaluated. Twelve plant calculations of key SA sequences illustrate the current status of SA codes to predict core degradation, in particular B4C effects
TMI-2 Accident Analysis with ASTEC V1.1 Code and Comparison with ICARE/CATHARE and SCDAP/RELAP5
No full textIn the frame of WP4 of SARNET Project regarding ASTEC benchmarks on reactor applications, the V1.1 version of ASTEC code has been assessed on TMI-2 accident to evaluate the capability of the code to simulate severe accident in nuclear power plant. The present analysis has been limited to the first two phases of the accident just prior to core reflood. A sensitivity analysis to most important and uncertain core degradation parameters has been carried out to investigate their influence on hydrogen generation and core melting. Finally, the ASTEC results have been compared with the results of new applications of ICARE/CATHARE and SCDAP/RELAP5 codes
TMI-2 Accident Analysis with SCDAP/RELAP5 Code
No full textAnalysis of the TMI-2 accident has been performed using the SCDAP/RELAP5/MOD3.2 system code, which is being developed by INEEL for the U.S. NRC. The nodalization of the TMI-2 plant included a detailed modeling of the primary coolant system with the reactor pressure vessel, the two coolant loops with once-through steam generators and main coolant pumps, the pressurizer with surge-line and pilot operated relief valve. The nodalization of the secondary system was limited to the secondary side of steam generators. The boundary conditions for the calculations were fixed according to the OECD TMI-2 Analysis Exercise. The first three phases of the accident including small break LOCA, core heat-up and melting, and core reflood were adequately modeled with SCDAP/RELAP5. Hydrogen generation and relevant primary system pressure increase were well predicted during both core heat-up and reflood. The code predicted the formation of a large molten pool in the central region of the core and complete quenching and fragmentation of embrittled fuel rod in the upper part of the core, in reasonable agreement with the TMI-2 accident scenario. No molten core material slumping to the lower head was computed by the code in the late phase of the accident
Analysis of the OECD LOFT Fission Product Experiment LP-FP-2 with ASTEC VI .2.1 Code
No full textIn the frame of WP4 of SARNET European Project the LOFT LP-FP2 experiment has been analysed with ASTEC V1.2.1 for code validation purposes. The LOFT input deck provided by IRSN has been improved according to standard PWR reactor applications and the experience gained in the analysis of Bethsy experiments. The calculation results have been compared with experimental data in order to assess the ability of ASTEC code for simulating thermal-hydraulic behaviour, core degradation, hydrogen generation and fission product release which might occur during a severe accident in a PWR. Furthermore, the ASTEC results have been compared with the results of ICARE/CATHARE code. Finally, some sensitivity studies have been performed to evaluate the impact of uncertain key parameters on code response
LBE-Water Interaction in Sub-Critical Reactors: First Experimental and Modelling Results
No full textThis paper concerns the study of the phenomena involved in the interaction between LBE and pressurised water which could occur in some hypothetical accidents in accelerator driven system type reactors. The LIFUS 5 facility was designed and built at ENEA-Brasimone to reproduce this kind of interaction in a wide range of conditions. The first test of the experimental program was carried out injecting water at 70 bar and 235 °C in a reaction vessel containing LBE at 1 bar and 350 °C. A pressurisation up to 80 bar was observed in the test section during the considered transient. The SIMMER III code was used to simulate the performed test. The calculated data agree in a satisfactory way with the experimental results giving confidence in the possibility to use this code for safety analyses of heavy liquid metal cooled reactors
Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis
No full textThe new reactor concepts proposed in the Generation IV International Forum require the development and validation of computational tools able to assess their safety performance. In the first part of this paper the models of the ESFR design developed by several organisations in the framework of the CP-ESFR project were presented and their reliability validated via a benchmarking exercise. This second part of the paper includes the application of those tools for the analysis of design basis accident (DBC) scenarios of the reference design. Further, this paper also introduces the main features of the core optimisation process carried out within the project with the objective to enhance the core safety performance through the reduction of the positive coolant density reactivity effect. The influence of this optimised core design on the reactor safety performance during the previously analysed transients is also discussed. The conclusion provides an overview of the work performed by the partners involved in the project towards the development and enhancement of computational tools specifically tailored to the evaluation of the safety performance of the Generation IV innovative nuclear reactor designs
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