An attempt to introduce a resuspension model in MELCOR 1.8.6 for fusion applications

During normal plasma operation the erosion of the ”plasma facing components” occurs and the dusts formed tends to deposit onto the divertor surface. In case of an In-vessel LOCA, these dusts may resuspend and transported to the VV Pressure Suppression System. Define the maximum amount of mobilized dust is an issue of main concern. MELCOR v1.8.6 hasn’t a resuspension model and an attempt to introduce a resuspension model in MELCOR was performed

Standalone Containment Analysis of Four Phébus Tests with the ASTEC and the MELCOR Codes

After the severe accident (SA) occurred at the Three-Miles Island Nuclear Power Plant (NPP), important efforts on the investigation of the different phenomena during this kind of accidents have been started. Several experimental campaigns investigating one phenomenon at time or the combination of two or more phenomena have been performed. Today, the Phébus experimental campaign is probably the most important activity on the evaluation of the coupling among different phenomena. Four out of five tests investigated the degradation of an intact Pressurized Water Reactor (PWR) fuel bundle and the subsequent transport of Fission Products (FP) and Structural Materials (SM) through the primary circuit and into the containment, while the fifth test was only the degradation of a bed of PWR fuel bundle debris. These tests were performed between 1990 and 2010 at the CEA Cadarache laboratories (France) in a 5000:1 scaled facility. The main four tests varied the employed control rod materials, the fuel burn-up, and the oxidizing conditions of the atmosphere (strongly or weakly). The outcomes of this experimental campaign created a solid base for the understanding of the involved phenomena and allowed the development of models and software codes capable of simulating the evolution of a SA in a real NPP. ASTEC and MELCOR were two of the main SA codes profiting from the results of this Phébus campaign. These two codes were further improved in the latest years to account for the findings obtained in more recent experimental campaigns. A continuous verification and validation work is then necessary to check how the newer code's versions reproduce the tests performed in these older experimental campaigns such as Phébus one. The present work is intended to be the final step of a series of publications covering the activities carried out at University of Pisa with the ASTEC and the MELCOR SA codes on the four Phébus tests employing an intact PWR fuel bundle. Because of the complexity and the extent of these tests, only the containment aspects were considered in the precedent works, i.e., only the thermal-hydraulics transient and its coupling with the FP and SM behavior. Then, general conclusions based on the outcomes of these precedent works are summarized in this work

Normal and Accidental Scenarios Analyses with MELCOR 1.8.2 and MELCOR 2.1 for the DEMO Helium-Cooled Pebble Bed Blanket Concept

As for Light Water Reactors (LWRs), one of the most challenging accidents for the future DEMOnstration power plant is the Loss of Coolant Accident, which can trigger the pressurization of the confinement structures and components. Hence, careful analyses have to be executed to demonstrate that the confinement barriers are able to withstand the pressure peak within design limits and the residual cooling capabilities of the Primary Heat Transfer System are sufficient to remove the decay heat. To do so, severe accident codes, as MELCOR, can be employed. In detail, the MELCOR code has been developed to cope also with fusion reactors, but unfortunately, these fusion versions are based on the old 1.8.x source code. On the contrary, for LWRs, the newest 2.1.x versions are continuously updated. Thanks to the new features introduced in these latest 2.1.x versions, the main phenomena occurring in the helium-cooled blanket concepts of DEMO can be simulated in a basic manner. For this purpose, several analyses during normal and accidental DEMO conditions have been executed. The aim of these analyses is to compare the results obtained with MELCOR 1.8.2 and MELCOR 2.1 in order to highlight the differences among the results of the main thermal-hydraulic parameters

Stand-alone containment analysis of the Phébus FPT-0 test with the ASTEC V2.1 and the MELCOR v2.2 codes

During the last 40 years, several efforts have been carried out to investigate the different phenomena occurring during a severe accident in a Nuclear Power Plant (NPP). Within this framework, the execution of different experimental campaigns, investigating only specific phenomena or the coupling among two or more phenomena, has been one of the main activity and the integral Phébus FP tests were probably the most important experiences in this field. In these tests, the degradation of a PWR fuel bundle and the related phenomena in the primary circuit and in the containment system were investigated, employing different control rod materials and fuel burn-up levels in strongly or weakly oxidizing conditions. Such Phébus integral tests were of fundamental importance to understand the key aspects of each phenomena and to develop numerical codes capable to simulate the evolution of a severe accident in a real NPP. Two of the main codes international employed (ASTEC and MELCOR) for severe accident analysis were intensively benchmarked basing on the findings of the different Phébus FP tests. In the latest years, these two codes were furthermore improved, to implement the more recent research findings after the termination of the Phébus experimental campaign, as the results obtained in the SARNET projects. Therefore, a continuous verification and validation work is still needed for the codes to check that the new improvements introduced in such codes really allow a better prediction of the Phébus tests and of the other tests forming the validation test matrix. The aim of the present paper is to re-analyze the first Phébus FPT-0 test employing the latest ASTEC V2.1 and MELCOR V2.2 code versions. The performed analysis focuses on the thermal-hydraulics /aerosol coupling, and only the stand-alone containment aspects of the test have been investigated. Three different spatial nodalizations of the Phébus containment vessel have been employed, showing that at least 15/20 control volumes are necessary for the vessel spatial schematization to correctly predict thermal-hydraulics and aerosol behavior. Furthermore, the paper summarizes the main thermal-hydraulic results and presents the different sensitivity analyses carried out on the iodine and aerosols behavior. When possible, a comparison among the results obtained during this work and by different authors in previous works is also performed, to highlight the improvements in the physical models implemented in the two codes

Analysis of the THAI Iod-11 and Iod-12 tests: Advancements and limitations of ASTEC V2.0R3p1 and MELCOR V2.1.4803

This work is related to the application of the ASTEC V2.0R3p1 and MELCOR V2.1.4803 codes to the analysis of the THAI Iod-11 and Iod-12 containment tests characterised by an iodine release. The main scope of these two tests was to investigate the steel interaction on dry and wet surfaces, with an interaction supposed to be a two-steps process: an initial faster and reversible physisorption followed by a slower, and irreversible, chemisorption of the physisorbed I2. The aim of the present work is to highlight advancements and limitations of the current ASTEC and MELCOR code versions respect to the older code versions employed during the European SARNET projects. The investigation was carried out as a code-to-code comparison vs. the experimental THAI data, focusing on the evaluation of the code models treating the iodine behaviour. A similar spatial nodalisation was employed for both codes. As main result, ASTEC had shown an overall good agreement compared to the iodine related experimental data while, on contrary, MELCOR had shown poor results, probably due to unsolved numerical issues and unsatisfactory iodine modellisation

Development of a SIMMER\RELAP5 coupling tool

Abstract The In-Box Loss Of Coolant (LOCA) postulated accident is considered a major concern for the safety connected with the development of EU-DEMO fusion reactor. Relating to the renewed interest in the Water-Cooled Lithium Led blanket concept, an innovative experimental campaign is under development at ENEA Brasimone laboratories aiming at investigating the consequences related to the In-Box LOCA applied to the WCLL breeding blanket. In this frame, a new coupling tool between the SIMMER-III (modified version to implement the PbLi/water chemical interaction) and the RELAP5/Mod3.3 codes (modified version to implement PbLi thermo-physical properties) has been developed together with its preliminary application to simple test cases with water as working fluid. The coupling procedure can be defined as a "two-way", "non-overlapping", "online" technique aiming at investigating multi-physics and multi-scales phenomena in support of the development of fusion reactor technologies

Best-Estimate for System Codes (BeSYC): A New Software to Perform Best-Estimate Plus Uncertainty Analyses with Thermal-Hydraulic and Safety System Codes for Both Fusion and Fission Scenarios

The development and the validation of old and new software in relevant DEMO reactor conditions have been exploited in the latest years within the EUROfusion Consortium. The aim was to use—if possible—the software already validated for fission reactors and to fill the gaps with new ad-hoc software. As contribution to this effort, the Karlsruhe Institute of Technology (KIT) developed and tested a novel software to apply the Best-Estimate Model Calibration and Prediction through Experimental Data Assimilation methodology to the system codes RELAP5-3D, MELCOR 1.8.6, and MELCOR 2.2. This software is called Best-estimate for SYstem Codes (BeSYC), and it is developed as a MATLAB App. The application is in charge of applying the mathematical framework of the methodology, writing and executing the code runs required by the methodology, and printing the obtained results. The main goal of BeSYC is to wrap up the methodology in a software suitable to be used by any user through a simple graphical user interface. Albeit developed in the fusion research context, BeSYC can be applied to any reactor/scenario type supported by the specific system code. The goals of BeSYC, the mathematical framework, the main characteristics, and the performed verification and validation activities are described in this paper

Post-Test Numerical Analysis of a Helium-Cooled Breeding Blanket First Wall under LOFA Conditions with the MELCOR Fusion Code

The validation of numerical tools employed in the analysis of incidental transients in a fusion reactor is a topic of main concern. KIT is taking part in this task providing both experimental data and by performing numerical analysis in support of the main codes used for the safety analyses of the Helium Cooled Pebble Bed (HCPB) blanket concept. In recent years, an experimental campaign has been performed in the KIT-HELOKA facility to investigate the behavior of a First Wall Mock-Up (FWMU) under Loss Of Flow Accident (LOFA) conditions. The aim of the experimental campaign was twofold: to check the expected DEMO thermal-hydraulics conditions during normal and off-normal conditions and to provide robust data for code validation. The present work is part of these validation efforts, and it deals with the analysis of the LOFA experimental campaign with the system code MELCOR 1.8.6 for fusion. A best-estimate methodology has been used in support of this analysis to ease the distinction between user’s assumptions and code limitations. The numerical analyses are here described together with their goals, achievements, and lesson learn