1,721,004 research outputs found
Safety and Radioactive Waste Management Aspects of the Ignitor Fusion Experiment
No full textIgnitor is a nuclear fusion experiment aimed at studying Deuterium-Tritium plasmas. If European proposed waste management strategies were applied, all Ignitor radioactive materials could be recycled or declassified to non-radioactive material. We have applied the Italian waste management regulations to the IGNITOR experiment radioactive materials: none of them should be classified in the High Level Waste category but the vessel, and most materials are classified as LLW (Low Level Waste). The machine has very low radiological risks and environmental impac
Tritium transport issues for helium-cooled breeding blankets
No full textTritium mobility through breeding blanket (BB) and steam generator heat transfer areas is a crucial aspect for the design of the next generation DEMO fusion power plants. Tritium is generated inside the breeder, dissolves in and permeates through materials, thus leading to a potential hazard for the environment. For this reason, it is important to carry out the tritium migration analysis for a specific DEMO blanket configuration to predict the released amount of tritium during the plant operation. Unfortunately, tritium assessments are often affected by several uncertainties implying very important modeling and parametric issues. In this paper, the main permeation issues are identified and possible solutions are discussed to address the modeling issues and the parametric uncertainties affecting the T migration assessments for the two DEMO helium-cooled BBs: 1) helium-cooled pebble beds and 2) helium-cooled lithium-lead. For these two helium-cooled blanket concepts various tritium migration analyses will be carried out by means of the computational tool FUS-TPC to define proper and feasible tritium mitigation techniques, which are needed to keep the tritium losses lower than the allowable environmental release (i.e., 20 Ci/d). © 1973-2012 IEEE
Tritium transport analysis in HCPB DEMO blanket with the FUS-TPC code
No full textIn the breeding areas a significant fraction of the tritium produced is extracted out from the Breeding Zone by the He gas purging the breeding ceramic in the helium cooled pebble bed (HCPB) blanket concept or transported in solution by the flowing liquid metal alloy in the helium cooled lead lithium (HCLL) blanket concept and then extracted outside the blanket box. Tritium produced inside the breeder can enter the metal structures and can be lost by permeation to the environment. It should therefore be kept under close control throughout the fusion reactor lifetime. In this study the problem of tritium transport in HCPB DEMO blanket has been studied and analysed by means of the computational code FUS-TPC, which has been originally developed to study the tritium transport in HCLL blanket and it has the main goal to provide the total tritium losses into the environment and the tritium inventories inside several HCPB blanket locations (e.g. purge gas and main coolant loops). © 2013 Elsevier B.V
Analisi del trasporto del trizio nei sistemi SFR
Full text linkUno degli isotopi radioattivi più difficili da contenere all'interno dei reattori SFR è il trizio in quanto diffonde prontamente attraverso i materiali strutturali alle temperature operative e conseguentemente può essere rilasciato in ambiente. AI fine di studiare ed analizzare possibili accorgimenti necessari per ridurre l'entità dei rilasci di trizio in ambiente è stato sviluppato un codice di calcolo in linguaggio MATLAB (SFR-TPC), il quale viene utilizzato per valutare i rilasci di trizio in ambiente e gli inventari di trizio all'interno dei reattori SFR. L'analisi condotta con l'utilizzo del codice SFR-TPC su un reattore di tipologia "a vasca" denominato PFBR (Prototype Fast Breeder Reactor), ha mostrato che, sotto alcuni ipotesi conservtive, solo circa 1.6 mg/y di trizio (su 3.874 g/y prodotti nel core e rilasciati nel refrigerante primario) raggiungono il ciclo del vapore e vengono conservativamente considerati come rilasci in ambiente. Nonostante le ipotesi conservative assunte il dato di rilascio ottenuto non costituisce un elevato rischio radiologico purché si adotti un opportuno set di misure mitigative per il trasporto di trizi
Pb-16Li/water interaction: Experimental results and preliminary modelling activities
No full textThe Water Cooled Lithium Lead (WCLL) blanket is based on the eutectic liquid alloy Pb16Li as breeder material and neutron multiplier, and pressurised water as coolant. The liquid breeder flows at few mm/s in the blanket module while the pressurised water is circulated inside double-wall tubes. In spite of the adoption of double-wall tubes for the coolant, the probability of a water large leak because of a tube rupture accident cannot be considered negligible. As a consequence, the Pb16Li/water interaction due to a large break in one or more cooling tubes still remains one of the biggest concerns for this blanket concept. This paper reports the results of three experimental tests on Pb16Li/water interaction carried out at ENEA-Brasimone operating the LIFUS 5 facility. Water was injected into the reaction tank, containing Pb16Li at 330 C, at a pressure of 155 bar with different values of sub-cooling and with different free volumes in the reaction system. In addition, post test analyses with SIMMER III code are presented in order to compare the pressure evolution measured during the experiments with that calculated by the code. © 2013 Elsevier B.V
Tritium production in breeding blankets
No full textThis chapter describes the equipment referred to as breeding blankets that will be used in the next fusion reactors. The mechanisms occurring in such components are addressed, including simplified diagrams to illustrate their functioning. The breeding function for an industrial size machine - the next step after ITER - is crucial because tritium is not available in sufficient quantities on the earth's surface. Only mock-ups - Test Blanket Modules (TBM) - will be tested in ITER to demonstrate the tritium breeding function and to collect experimental data. The TBM are supported by processes aiming to recover the heat generated by the plasma operations and to recover the tritium produced by neutron bombardment. The first section of this chapter focuses on the various TBM concepts proposed by the members of the ITER experimental and scientific programme. Characteristics and expected performance levels are summarized and discussed. In the second section, the processes selected to support the TBM operations are described. The technical issues and/or technical locks regarding tritium technologies are then highlighted in the third and final section. © 2013 Nova Science Publishers, Inc. All rights reserved
Tritium technologies for pbli based breeding blankets
No full textA lead-lithium based breeding blanket for DEMO and its corresponding Test Blanket Module (TBM) to be tested in ITER is being developed in EU. The same approach is pursued in other countries involved in the development of fusion nuclear technologies. The eutectic lithium lead alloy (Pb-16Li) includes the tritium breeder material (Li) and the neutron multiplier (Pb) and, used in liquid form, it can be recirculated outside the blanket module to facilitate the tritium extraction and the alloy purification. However, it shows also some drawbacks. In fact, besides the issues related to the magnetohydrodynamic (MHD) effects and the compatibility of liquid lead lithium alloy with structural materials, it is known that the tritium Sieverts' constant in Pb-16Li is low. As a consequence, an high tritium permeation rate from the liquid metal to the primary cooling system is expected, at least in absence of efficient tritium permeation barriers. This aspect affects the whole blanket tritium cycle, whose main steps consist of tritium extraction from the liquid breeder (TES, Tritium Extraction System) and tritium removal from He primary coolant (CPS, Cooling Purification System). This chapter is focused on the description of tritium processing systems and tritium technologies related to the adoption of the lead-lithium alloy in DEMO breeding blankets. The first section describes the DEMO fuel cycle giving a particular accuracy to the description of the breeding blanket part. The second section introduces the hydrogen isotopes transport parameters in Pb-16Li which have a strong impact on the process strategy of the blanket fuel cycle. Then, the above-mentioned tritium processing systems and technologies will be described in detail. In particular, an assessment of the tritium systems that can meet the requirements of a lead-lithium based DEMO blanket is carried out presenting the suitable technologies together with their drawbacks and advantages. © 2013 Nova Science Publishers, Inc. All rights reserved
Analysis of the LBE-Water Interaction in the Lifus 5 Facility to Support the Investigation of a SGTR Event in LFRs
Full text linkThe main purpose of this experimental and computational study is to obtain an assessment of the physical effects following energetic interaction between Iead-bismuth eutectic alloy (LBE) and water; this phenomenon is of great relevance in the analysis of a SGTR (Steam Generator Tube Rupture) accident in a Lead-cooled Fast Reactor (LFR) as ELSY and for the qualification of the models developed for the SIMMER code. The experiment named ELSY-2, performed with the facility LIFUS 5, was analysed with SIMMER code using both the 2D and 3D versions. The calculation models employed in the simulations were realized on the basis of the facility's real technical specifications. In particular, the calculated pressure time trend inside the reaction vessel of LIFUS 5 facility resulted as being in quite good agreement with the experimental data, although both versions of the code tended to overestimate the pressure in the first stage of the transient
Tritium permeation issues for helium-cooled breeding blankets
No full textTritium permeation through Breeding Blanket and Steam Generator heat transfer areas is a crucial aspect for the design of the next generation DEMO fusion power plants. Tritium is generated inside the breeder, dissolves in and permeates through materials, thus leading to a potential hazard for the environment. For this reason it is important to carry out the tritium migration analysis for a specific DEMO blanket configuration in order to predict the released amount of tritium during the plant operation. Unfortunately, tritium assessments are often affected by several uncertainties implying very important modelling and parametric issues. In this study the main permeation issues are identified and possible solutions are discussed to face the modelling issues and the parametric uncertainties affecting the T migration assessments for the two DEMO helium-cooled breeding blankets, i.e.: 1) Helium-Cooled Pebble Beds (HCPB) and 2) Helium-Cooled Lithium-Lead (HCLL). For these two helium-cooled blanket concepts various tritium migration analyses will be carried out by means of the computational tool FUS-TPC in order to define proper and feasible tritium mitigation techniques which are needed to keep the tritium losses lower than the allowable environmental release (i.e. 20 Ci/d). © 2013 IEEE
Sensitivity study for tritium permeation in helium-cooled lead-lithium DEMO blanket with the FUS-TPC code
No full textHydrogen dissolves in and permeates through most materials, thus it is important to understand the permeation, diffusion and dissolution phenomena of hydrogen and its isotopes in materials. We address the problem of tritium transport in Helium Cooled Lead-Lithium (HCLL) DEMO blanket from leadlithium breeder through different heat transfer surfaces to the environment by developing a computational code (FUS-TPC). The main features of the code are briefly described and a parametric study is performed in order to identify the most influencing parameters in terms of tritium releases into the environment and of tritium inventories. The results showed that the results are strongly affected by the tritium Sievert's constant in Lead-Lithium and the efficiency of permeation barriers
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