5,081 research outputs found
Analysis of the influence of the heat transfer phenomena on the late phase of the ThAI Iod-11 and Iod-12 tests
No full textIodine is one of the major contributors to the source term during a severe accident in a Nuclear Power Plant for its volatility and high radiological consequences. Therefore, large efforts have been made to describe the Iodine behaviour during an accident, especially in the containment system. Due to the lack of experimental data, in the last years many attempts were carried out to fill the gaps on the knowledge of Iodine behaviour. In this framework, two tests (ThAI Iod-11 and Iod-12) were carried out inside a multi-compartment steel vessel. A quite complex transient characterizes these two tests; therefore they are also suitable for thermal-hydraulic benchmarks. The two tests were originally released for a benchmark exercise during the SARNET2 EU Project. At the end of this benchmark a report covering the main findings was issued, stating that the common codes employed in SA studies were able to simulate the tests but with large discrepancies. The present work is then related to the application of the new versions of ASTEC and MELCOR codes with the aim of carry out a new code-to-code comparison vs. ThAI Iod-12 experimental data, focusing on the influence of the heat exchanges with the outer environment, which seems to be one of the most challenging issues to cope with
Stand-alone containment analysis of the Phébus FPT-1 test with the ASTEC and the MELCOR codes
No full textThe estimation of Fission Products (FPs) release from the containment system of a nuclear plant to the external environment during a Severe Accident (SA) is a quite complex task. In the last 30-40 years several efforts were made to understand and to investigate the different phenomena occurring in such a kind of accidents in the primary coolant system and in the containment. These researches moved along two tracks: understanding of involved phenomenologies through the execution of different experiments, and creation of numerical codes capable to simulate such phenomena. These codes are continuously developed to reflect the actual SA state-of-the-art, but it is necessary to continuously check that modifications and improvements are able to increase the quality of the obtained results. For this purpose, a continuous verification and validation work should be carried out.
Therefore, the aim of the present work is to re-analyze the Phébus FPT-1 test employing the ASTEC (F) and MELCOR (USA) codes. The analysis focuses on the stand-alone containment aspects of the test, and three different modellisations of the containment vessel have been developed showing that at least 15/20 Control Volumes (CVs) are necessary for the spatial schematization to correctly predict thermal-hydraulics and the aerosol behavior. Furthermore, the paper summarizes the main thermal-hydraulic results, and presents different sensitivity analyses carried out on the aerosols and FPs behavior
An attempt to introduce a resuspension model in MELCOR 1.8.6 for fusion applications
No full textDuring 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
Comparison among MELCOR 1.8.2 fusion version and MELCOR 2.1 during normal and accidental scenarios for the Primary Heat Transfer System of the DEMO Helium Cooled Pebble Bed blanket
No full textThe future introduction of fusion power plants requires also the demonstration that all the radiological risks, in term of potential hazards to the staff, the population and the environment, are below the limits established by national authorities in both normal and accidental conditions. As for Light Water Reactors (LWR), one of the most challenging accidents is the Loss of Coolant Accident (LOCA), which causes the depressurization of the Primary Heat Transfer System (PHTS) and the pressurization of the confinement structures and components, as the Vacuum Vessel (VV), the Expansion Volume (EV), and the Tokamak Building (TB). Hence, several analyses should be executed to demonstrate that the confinement barriers are able to withstand the accident pressure peak within design limits and the residual cooling capabilities of the PHTS are sufficient to remove the decay heat coming from the In-Vessel components. In LWRs these analyses are commonly executed employing "Severe Accident" codes, which are mainly integral codes able to simulate the incidental scenario from the initiating event to the release of radionuclides outside the containment. In principle, the same codes could be also employed for fusion reactors, but due to the intrinsic and deeper differences among the two reactor types it cannot be assured the quality and the correctness of the results obtained. For this purpose, some codes have been expanded to cope also with fusion reactors and their specific phenomena. One of the codes which have undergone this adaptation process is MELCOR, but unfortunately the only version developed under a quality assurance program is the old 1.8.2, which is no longer maintained. On the contrary, for LWRs, the latest MELCOR version available is 2.1, which was also expanded by SNL (Sandia National Laboratories) to cope with HTGRs (High Temperature Gas Reactor). Hence, this latest available MELCOR version should be also capable to treat, in a basic manner, the main phenomena occurring in the helium-cooled blanket concepts of DEMO. For this purpose, several analyses during normal and accidental (Ex-Vessel LOCAs) conditions have been executed considering the Primary Heat Transfer System (PHTS) of the DEMO Helium Cooled Pebble Bed (HPCB) blanket concept. The aim of these analyses is to compare the results obtained with MELCOR 1.8.2 fusion version vs. the MELCOR 2.1 standard version, in order to highlight the differences among the results on the main thermal-hydraulic parameters. In particular, the heat transfer coefficients predicted among the heat structures and the control volumes are investigated, being one of the main indicator of the differences among the two code versions
Stand-alone Containment Analysis of four Phébus tests with the ASTEC and MELCOR codes
No full textAfter 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 to simulate 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 thermalhydraulics
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
Stand-Alone Containment Analysis of the Phébus FPT Tests with the ASTEC and the MELCOR Codes - The FPT-0 Test
No full textDuring the last 40 years, several efforts have been carried out to investigate the different phenomena occurring during
a Severe Accident (SA) in a Nuclear Power Plant (NPP). Such efforts have been supported with the execution of different experimental campaigns investigating only specific phenomena or the coupling among two or more of them. The integral Phébus tests were probably one of the most important experiences in this field. In these tests, the degradation of a PWR fuel bundle was investigated employing different control rod materials and burn-up levels in strongly or weakly oxidizing conditions. Such 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 SA in a real NPP. Two of the main SA codes developed also basing on the findings of the Phébus tests were the ASTEC and the MELCOR codes. In the latest years, these two codes were furthermore expanded to implement the more recent findings after the termination of the Phébus experimental campaign. Therefore, a continuous verification and validation work is necessary to check that the new improvements introduced in such codes allow also a better prediction of these Phébus tests. Therefore, the aim of the present work is to re-analyze the Phébus FPT-0 test employing the up-to-date ASTEC and MELCOR codes. The analysis focuses only on the stand-alone containment aspects of this test, and three different spatial nodalizations of the containment vessel have been developed, showing that at least 15/20 Control Volumes (CVs) are necessary for the spatial schematization to predict thermal-hydraulics and the aerosol behavior. Then, the paper summarizes the main thermal-hydraulic results and presents different sensitivity analyses carried out on the aerosols and Fission Products (FPs) behavior. When possible, a comparison among the results obtained in this work and by different authors in previous works is also performed. This paper is part of a series of publications covering the four executed Phébus tests employing a solid PWR fuel bundle: FPT-0, FPT-1, FPT-2, and FPT-3
Implementation and Validation of a Resuspension Model in MELCOR 1.8.6 for Fusion Applications
No full textIn a fusion reactor, a continuous erosion of the Plasma Facing Surfaces (PFSs) occurs during normal plant operations. This erosion creates a particles debris (dust) spanning from 1 μm to 50 μm in dimension. These formed dust deposits on the bottom of the Vacuum Vessel (VV), i.e. on the divertor surface. In case of Beyond Design Basis Accident (BDBA), this dust may be mobilized and transported towards the confinement building, or even into the outer environment. Therefore, the evaluation of the maximum mobilized dust mass is a safety issue of main concern, because it may pose a radiological risk to plant operators and to the outer population. To investigate these incidental scenarios, lumped-parameters codes such MELCOR are commonly employed. A specific version of MELCOR, capable to treat the phenomena occurring in a fusion reactor, is developed by Idaho National Laboratory (INL). Although, a model to treat the mobilization of dust is still not yet implemented in this specific MELCOR version. This paper presents, after a review of the available resuspension models, the selection of a specific resuspension model and the efforts made for its validation against different experimental tests. The selected model is the semi-empirical “Force Balance” model, just implemented in the ASTEC and ECART codes, but specific modifications were introduced to allow its implementation in MELCOR through ad hoc developed Control Functions (CFs). Its validation shows a quite good agreement with most of the experimental tests investigated, highlighting its capabilities for the safety analysis of the forthcoming fusion reactors
Stand-alone containment analysis of the Phébus FPT tests with the ASTEC and the MELCOR codes: the FPT-2 test
No full textDuring the last 40 years, several efforts have been carried out to investigate the different phenomena occurring during a Severe Accident (SA) in a Nuclear Power Plant (NPP). Such efforts have been supported with the execution of different experimental campaigns, and the integral Phébus tests were probably some of the most important experiments in this field. In these tests, the degradation of a Pressurized Water Reactor (PWR) fuel bundle was investigated employing different control rod materials and burn-up levels in strongly or weakly oxidizing conditions. From the findings on these and previous tests, numerical codes such as ASTEC and MELCOR have been developed to analyze the evolution of a SA in real NPPs. After the termination of the Phébus campaign, these two codes have been furthermore improved to implement the more recent findings coming from different experimental campaigns. Therefore, continuous verification and validation is still necessary to check that the new improvements introduced in such codes allow also a better prediction of these Phébus tests. The aim of the present work is to re-analyze the Phébus FPT-2 test employing the updated ASTEC and MELCOR code versions. The analysis focuses on the stand-alone containment aspects of this test, and three different spatial nodalizations of the containment vessel have been developed. The paper summarizes the main thermal-hydraulic results and presents different sensitivity analyses carried out on the aerosols and fission products (FPs) behavior. When possible, a comparison among the results obtained during this work and by different authors in previous work is also performed. This paper is part of a series of publications covering the four Phébus FP tests using a PWR fuel bundle: FPT-0, FPT-1, FPT-2, and FPT-3, excluding the FPT-4 one, related to the study of the release of low-volatility FPs and transuranics from a debris bed and a bath of melted fuel
Normal and Accidental Scenarios Analyses with MELCOR 1.8.2 and MELCOR 2.1 for the DEMO Helium-Cooled Pebble Bed Blanket Concept
No full textAs 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-1 test with the ASTEC and the MELCOR codes
No full textThe estimation of fission products (FPs) release from the containment system of a nuclear
plant to the external environment during a severe accident (SA) is a quite complex task.
In the last 30–40 yr, several efforts were made to understand and to investigate the different
phenomena occurring in such a kind of accidents in the primary coolant system and
in the containment. These researches moved along two tracks: understanding of involved
phenomenologies through the execution of different experiments and creation of numerical
codes capable to simulate such phenomena. These codes are continuously developed
to reflect the actual SA state of the art, but it is necessary to continuously check that modifications
and improvements are able to increase the quality of the obtained results. For
this purpose, also a continuous verification and validation work should be carried out.
Therefore, the aim of the present work is to re-analyze the Phebus fission products test 1
(FPT-1) test employing the accident source term evaluation code (ASTEC) and MELCOR
codes (respectively, ASTEC v.2.0 revision 3 patch 3 and MELCOR V2.1.6840 version).
The analysis focuses on the stand-alone containment aspects of the test, and three different
modelizations of the containment vessel have been developed showing that at least
15/20 control volumes (CVs) are necessary for the spatial schematization to correctly
predict the test thermal hydraulics and the aerosol behavior. Furthermore, the paper
summarizes the main thermal-hydraulic results and presents different sensitivity analyses
carried out on the aerosols and FPs behavior
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