468 research outputs found
Tritium and dust source term inventory evaluation issues in the European DEMO reactor concepts
No full textFusion reactors represent a future evolution of the nuclear technology improving the world-wide energy portfolio. The experimental fusion reactor under construction (ITER) and the planned industrial fusion reactors (DEMO) are large and complex facilities. For their operation it is necessary to ensure safety limiting of radiological and mobilizable source terms inventory, such as tritium and radioactive dust. The source term inventories shall be assumed in the establishment of operational and safety requirements for DEMO, as well as performing safety analyses for commercial fusion devices. In the last few years a methodology for evaluation of tritium and dust source term inventory has been proposed in the framework of the EUROFusion project. The basis of the methodology is a semi-empirical approach to scale the radioactive inventory limits implemented in ITER. The source term amounts derived for DEMO will supply a guidance for safe operation of future fusion reactors. The development of methodology has to be completed and refined because of the lack of validation versus adequate experimental data and rules for extension to different scenarios. The aim of this work is assessment of the developed methodology for evaluation of the source terms inventory versus JET operating limits for tritium and versus the dust control strategy adopted in ITER for the dust. The updating of the methodology for the dust and T inventory was carried out in line with the current DEMO design. The new values of the in-VV source terms were achieved. These values are recommended to be used for the safety assessment of the fusion reactor. Calculated DST and TST inventories’ values show a reasonable agreement with scaled ITER or assessed JET limits
Experimental and Numerical Analysis of the Air Inflow Technique for Dust Removal from the Vacuum Vessel of a Tokamak Machine
No full textIn fusion facilities, the dust production inside the plasma chamber is a concern from the viewpoint of both machine performance and safety. To the purpose of a correct handling of the experimental devices the problem of its removal must be properly solved. This work deals with the experiments carried out in the STARDUST facility by using as dust removal technique an air inflow into the volume representing the vacuum vessel. The goal was to evaluate the effectiveness of such an approach, less invasive as compared to all the others so far. These experiments, performed by using characterized carbon, tungsten and stainless steel dusts, show that the mobilization capability of the air inflow is between few percent and 100%, mainly depending on dust type of and deposit shape. The capture efficiency in a filter reached a maximum of about 7.5% in the STARDUST geometrical configuration. In conclusion, this simple and clean (from the radioactive point of view) removing technique needs particular care to be more effective and is not the perfect solution due to its low efficiency in the collection of removed powder in proper surfaces (i.e., filters). Nevertheless improvements are possible and worthwhile
Analysis of an Ex-Vessel Break in the ITER Divertor Cooling Loop
No full textIn the present work the integrated ECART code, developed for severe accident analysis in LWRs, is applied on the analysis of a
large ex-vessel break in the divertor cooling loop of the international thermonuclear experimental reactor (ITER).Acomparison of
the ECART results with those obtained by Studsvik Nuclear AB (S), utilizing the MELCOR code, was also performed in the general
framework of the quality assurance program for the ITER accident analyses. This comparison gives a good agreement in the
results, both for thermal-hydraulics and the environmental radioactive releases. Mainly these analyses, from the point of view of
the ITER safety, confirm that the accidental overpressure inside the vacuum vessel and theTokamak coolingwater system (TWCS)
Vault is always well below the design limits and that the radioactive releases are adequately confined below the ITER guidelines
Ice layer growth on a cryogenic surface in a fusion reactor during a loss of water event
No full textThe design of fusion devices often includes water as primary coolant of the first wall/blanket system and a potential accident scenario is the steam/water injection from the primary circuit into the magnets' cryogenic chamber (cryostat). An important question to be answered for the above scenario is whether the pressure suppression created by the cryogenic surfaces is sufficient to prevent cryostat damage. The computer codes used for the assessment of ITER safety were validated in the past years against the EVITA (Experimental Vacuum Ingress Test Apparatus) experiment at CEA-Cadarache, which was designed for the simulation of the physical phenomena occurring during a coolant ingress into the cryostat, namely ice formation on a cryogenic structure, heat transfer coefficient between walls and fluid, flashing, two-phase critical flow. The paper presents the results obtained by the CONSEN computer program for seven post-test calculations of the EVITA facility relating to the cryogenic experiments carried out in 2004 and 2005, in which the kinetics of the ice layer formation was analysed. The comparison with the experimental data has been performed and the main agreements and differences are remarked
Estimation of tritium and dust source term in european DEMOnstration fusion reactor during accident scenarios
No full textThe safety features of the future nuclear fusion reactors are one of the key issues for their attractiveness if compared with the fission plants. In fusion devices, accidents with high release of radioactive materials have low probabilities because the most part of abnormal transients lead to passive plasma shutdown. It does not mean that radiological source terms such tritium and activated dust are not generated and released, but their inventory does not increase during abnormal events. Therefore, the source term inventory has to be assessed during normal operation and traced when accidents occur. For this reason, a study for qualification and quantification of the tritium and dust source term (DTS) was established with the aim to understand their production, deposition, and penetration in the vacuum vessel (VV) and in the breeding blanket (BB). The main concern is source term release during the main accident scenarios to comply with a future licensing process. In case of abnormal event scenarios, the source term inventory involved in the release changes and requires a different confinement approach and mitigation. For the estimation of the source term in the DEMOnstration Fusion Power Station (DEMO), a methodology was developed. The methodology scales the tritium and DTS inside the VV from the International Thermonuclear Experimental Reactor, the European Power Plant Conceptual Study, and reports the tritium generated inside the breeder blanket from data quantified in other studies for DEMO. In this article, the methodology was updated and tritium and DTS for DEMO 2016 design were estimated. Moreover, the tritium and dust release pathways were highlighted according to different accidental scenarios. These results were obtained for all blanket concepts, which are analyzing in the ongoing DEMO EUROFusion project. The values estimated in this article will be used in the safety analyses to evaluate releases or to quantify the operational limits starting from values postulated in International Thermonuclear Experimental Reactor
Analysis of the ICE Experimental Tests Using the ECART Code
No full textIn the present paper the work performed to validate the ENEL/EDF ECART code on the base of a series of experimental tests performed in the Japanese ICE facility is presented. This activity has been carried out in the general framework of the validation phase of the ECART code, initially developed for integrated analysis of severe accidents in LWRs, for its application on incidental sequences related to fusion plants. The ICE facility consists of the cylindrical vacuum vessel, at horizontal axis (900. mm of external diameter and 600. mm length), the boiler, the blow-down tank and the corresponding piping and valves. The vessel has heat plates to maintain the required wall temperatures and initially contains dry air at a sub-atmospheric pressure of about 100. Pa. The employed ECART code full-couples the aerosol-vapour transport phenomena with thermal-hydraulics and chemical equilibrium. For the present work purposes, due to the ICE tests characteristics, only the thermal-hydraulic code section was activated and the dry aerosol or the chemical equilibrium modules were not employed. Two different conclusions have been highlighted by the work. The first one, being the main goal of the work, is related to the assessment of the ECART code in experimental conditions relevant for the future fusion reactors, against the traditional LWRs assessment previously performed for this code. The second one is related to the evaluation of the ICE experimental tests
Failure Mode and Effect Analysis for the Water Cooled Lithium-Led Blanket of Iter Test Module (ITM). Effects of Small Leaks on the ITM Safety. Final Report
No full textThe final Failure Mode and Effect Analysis (FMEA) on the Water Cooled Lithium-Lead Blanket of Iter Test Module (ITM) has been performed basing on Design Description Document. As a result a set of Postulated Initiating Events (PIEs) to be taken into account in the deterministic transient analyses has been defined.Detailed tables are reported about: list of components in ITM; possible failure modes of components together with related causes, consequences and preventive/mitigative actions; list of PIEs with related contributors, frequency and category values.Evaluation of the effects of small leaks of water in liquid lithium-lead alloy was performed to point out the safety concerns
Exploratory fire analysis in DONES lithium system
No full textThe exploitation of Nuclear Fusion energy in power plants will require the development and qualification of materials able to withstand outstanding neutronic loads for which no operating experience is currently available. Therefore, a Demo-Oriented early NEutron Source (DONES) facility for material irradiation is currently being designed within EUROfusion programme aimed at the production of neutrons with fusion relevant spectrum and fluence by means of D – Li stripping reactions occurring between a deuteron beam impacting a stable liquid lithium flowing film implementing the target. The Lithium System (LS) in DONES shall provide circulation of liquid lithium with suitable thermal-hydraulic characteristics and assure impurity control and heat removal. Given the hazard constituted by liquid lithium inventory and the potential risk of reactions with air eventually resulting into fire events, a preliminary evaluation of the modality of occurrence and evolution of such abnormal events in LS has been performed. In particular, two events have been selected for fire analysis. The first event considers a failure of off-line sampler equipment with water getting in contact with sampled lithium inventory. The initiating event is a failure in the glove box containing the off-line sampler. An air ingress occurs in the glovebox and a break of the Li flask is hypothesized at the same time. The second event assumes a leak at the outlet of the electromagnetic pump with loss of Lithium from the Li pipe in Heat Rejection System in LS room. Simultaneous air ingress is hypothesized as well in LS room (normally filled with inert argon). Fire loads were initially identified for the selected events and the room models developed considering dimensions, lithium inventory and fire compartment assumed to coincide with room limits. The ignition of lithium in contact with air occurs at liquid lithium operating temperature as reported in most conservative observations in literature and lithium fires were simulated as heat flux associated to lithium – air reactions rates observed in literature. A model for both events was implemented in Fire Dynamic Simulator (FDS) code to evaluate fire dynamics and a sensitivity analysis was performed on relevant inventories in the lithium loop area to investigate possible consequences
Validation of the ECART code for the safety analysis of fusion reactors
No full textRealistic calculation of the radiotoxic substances transport within a fusion facility requires a coupling among thermal–hydraulic, chemistry and aerosol-vapour models. The paper introduces the methodology adopted for the simulation of these phenomena with ECART, a code developed by a pool of Italian institutions with the support of European Union and EDF. In the past, this tool was successfully validated against the major source term tests applied to Light Water Reactor (LWR) safety and it is now employed in investigations about advanced LWRs and non-nuclear risk studies. It also contains information about chemical compounds involved in fusion reactor safety and simulates the related oxidation reactions. With regard to its application on fusion, a large validation activity was performed, mainly based on the analyses of experimental programs promoted inside the EURATOM Fusion Technology Programme. The correct simulation of main phenomena occurring in ICE and STARDUST facilities demonstrates the applicability of ECART in performing a realistic prediction of the whole sequence (thermal–hydraulics and dust transport) inside fusion plants
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