Abstract The accurate prediction of tritium inventory and permeation fluxes in the breeding blank... more Abstract The accurate prediction of tritium inventory and permeation fluxes in the breeding blanket of a D-T fusion reactor is a key aspect for future thermonuclear power plants licensing. Tritium permeation into structural materials could give rise to potential issues concerning the fuel self-sufficiency and can be lost into the environment with resulting radiological risks for the population. In the frame of hydrogen isotopes transport modelling, the Tritium Migration Analysis Program (TMAP) code is considered a referred code for the safety design of nuclear fusion power plants. It is mainly applied to Plasma Facing Components (PFCs), which are subjected to intense particles implantation and surface heat fluxes. Due to some code limitations and to the need of having a more easy-to-use computer code in a multiphysics framework, in the last years the commercial software COMSOL Multiphysics has been used to assess tritium transport studies as an application to ITER, DEMO and CFETR fusion reactors, but it was never validated against TMAP. Within this paper, a COMSOL-based code named mHIT (multi-trapping Hydrogen Isotopes Transport code) is presented. A set of verification and validation (V&V) problems were addressed, with the aim of substantiating the capabilities of the code in common fusion-relevant experimental set-ups and to include different physics according to the level of detail needed for a given application.
In HCLL (Helium Cooled Lithium Lead) blanket for ITER fusion reactor, an efficient management ope... more In HCLL (Helium Cooled Lithium Lead) blanket for ITER fusion reactor, an efficient management operation of tritium is necessary, in order to minimize the release of tritium to the environment and to improve the economy of the overall balance of the reactor tritium inventory. In this context, optimization techniques for the measurement of tritium in the various stages of the fuel cycle of the reactor must be carried out in parallel with the research and development of processing systems for tritium. For example, the TES, Tritium Extraction System, is a critical system and the determination of the concentration of hydrogen isotopes is required for its correct design [1]. Therefore, the need of a fast and reliable sensor for the measurement of the tritium concentration in the liquid breeder is well established. For this application, a sensor based on the principle of tritium permeation in a capsule of suitable material seems to be, at least in principle, the easiest and most reliable one. Furthermore, many data on tritium permeability through different materials and their compatibility in an environment of liquid metal are available, thanks to the experimental studies performed for the selection of structural materials for blanket modules
An analytical study, corroborated by Monte Carlo simulations, is presented which describes the in... more An analytical study, corroborated by Monte Carlo simulations, is presented which describes the interaction of a 300 keV–830 mA proton beam with a Cu and Al dumping system. The analytical calculations rely on the theoretical framework of the Particle Induced X-ray Emission, while the Monte Carlo simulations are performed by means of the GEANT4 toolkit. The case study is related to the project SORGENTINA-RF fusion neutron source and in particular to the tests devoted to assess the performance of the ion source of the plant. The results provide a detailed physical insight of the main processes occurring in the beam dump material, and are also important to give some realistic figures of the radiation emission expected during operation.
The latest progress in the design of the water-cooled lithium–lead (WCLL) tritium extraction and ... more The latest progress in the design of the water-cooled lithium–lead (WCLL) tritium extraction and removal (TER) system for the European DEMO tokamak reactor is presented. The implementation and optimization of the conceptual design of the TER system are performed in order to manage the tritium concentration in the LiPb and ancillary systems, to control the LiPb chemistry, to remove accumulated corrosion and activated products (in particular, the helium generated in the BB), to store the LiPb, to empty the BB segments, to shield the equipment due to LiPb activation, and to accommodate possible overpressure of the LiPb. The LiPb volumes in the inboard (IB) and outboard (OB) modules of the BB are separately managed due to the different pressure drops and required mass flow rates in the different plasma operational phases. Therefore, the tritium extraction is managed by 6 LiPb loops: 4 loops for the OB segments and 2 loops for the IB segments. Each one is a closed loop with forced circul...
A study devoted to assessing the neutronic performance of the SORGENTINA-RF fusion neutron source... more A study devoted to assessing the neutronic performance of the SORGENTINA-RF fusion neutron source is here presented. The calculation is performed considering the characteristics of the accelerated mixed deuterium/tritium ion beam from the ion source, the transport of ions into the titanium layer that constitutes the ion beam target, their implantation and the fusion reaction cross section. An investigation has been done by varying the ion composition of the beam in order to assess if and to what extent diatomic and triatomic ion species affect the neutron emission rate of the plant. This calculation is also instrumental to evaluate the cost-effectiveness of equipping the ion source with a filtering system to remove secondary ions from the beam.
The SORGENTINA-RF project aims to develop a 14 MeV fusion neutron source to produce medical radio... more The SORGENTINA-RF project aims to develop a 14 MeV fusion neutron source to produce medical radioisotopes with special focus on $$^{99}$$ 99 Mo. The facility is based on a positive ion source with an acceleration stage to produce a deuterium (D$$^{+})$$ + ) and tritium (T$$^{+})$$ + ) ion beam that, impacting on a titanium-coated rotating target, allows fusion reactions to take place. Maximizing the neutron production rate is one of the main issues to be addressed in the project and the optimization of some key parameters of the ion beam is of paramount importance in this regard. In this study, a methodology is discussed to reach a definition of the beam characteristics for an effective and sustainable operation of the plant. The most convenient layout that has been found out is based on a single ion source fed by a deuterium and tritium gas mixture. Eventually, a series of considerations about the operation of the ion source and fuel cycle have been drawn.
The SORGENTINA-RF project aims at developing a 14 MeV fusion neutron source featuring an emission... more The SORGENTINA-RF project aims at developing a 14 MeV fusion neutron source featuring an emission rate in the order of 5–7 × 1013 s−1. The plant relies on a metallic water-cooled rotating target and a deuterium (50%) and tritium (50%) ion beam. Beyond the main focus of medical radioisotope production, the source may represent a multi-purpose neutron facility by implementing a series of neutron-based techniques. Among the different engineering and technological issues to be addressed, the production of incondensable gases and corrosion product into the rotating target deserves a dedicated investigation. In this study, a preliminary analysis is carried out, considering the general layout of the target and the present choice of the target material.
The experimental qualification of the tritium extraction unit (TEU) from the LiPb eutectic alloy ... more The experimental qualification of the tritium extraction unit (TEU) from the LiPb eutectic alloy (15.7 at. % Li), the breeder material of the water-cooled lithium–lead breeding blanket concept, is one of the fundamental items for the demonstration of tritium balance sustainability for ITER and DEMO fusion reactors. Several technologies have been proposed as TEU, but the selection of the reference technology can be carried out only after the experimental measurement of the tritium extraction efficiency. For this purpose, a dedicated facility, called TRIEX-II, was designed and installed at ENEA Brasimone Research Centre, Italy. The facility is able to qualify gas–liquid contactor (GLC), permeator against vacuum and liquid-vacuum contactor technologies at different temperatures, lithium–lead mass flow rates and hydrogen isotopes concentrations. In TRIEX-II, the hydrogen or deuterium, used to simulate tritium, are solubilised inside the LiPb with a dedicated saturator and are then extra...
Abstract A crucial issue for the design of a fission Generation IV Sodium Fast Reactor (SFR) is t... more Abstract A crucial issue for the design of a fission Generation IV Sodium Fast Reactor (SFR) is to efficiently characterize the processes and the technical solutions devoted to the management and the control of tritium transfers and releases occurring in SFR circuits. In fact, the tritium balance in sodium fast reactor is a key aspect and the determination of the hydrogen isotopes (hydrogen/tritium) concentration in the liquid sodium has to be investigated for the operational control of tritium balance in the reactor. Therefore, the need for a fast and reliable sensor for the online measurement of hydrogen isotopes concentration in sodium is evident. In the present work, a new concept of hydrogen isotopes permeation sensor (HPS) is shown. The physical operating principle of the HPS is the permeation of hydrogen isotopes throughout a thin metallic membrane. Since the limit of detection and the accuracy of the integrated sensor has to be compatible with the range of tritium concentrations in sodium circuits, it is necessary for the sensor to be operated in high vacuum (HV) conditions in order to guarantee hydrogen isotopes measurements down to 10−5 Pa or even lower. The paper presents two designs foreseen not only for SFRs, but also for an experimental sodium loop where higher H 2 concentrations will be adopted.
Tritium permeation into the Primary Heat Transfer System (PHTS) of DEMO and ITER reactors is one ... more Tritium permeation into the Primary Heat Transfer System (PHTS) of DEMO and ITER reactors is one of the challenging issues to be solved in order to demonstrate the feasibility of nuclear fusion power plants construction. Several technologies were investigated as antipermeation and corrosion barriers to reduce the tritium permeation flux from the breeder into the PHTS. Within this frame, alumina coating manufactured by Pulsed Laser Deposition (PLD) and Atomic Layer Deposition (ALD) are two of the main candidates for the Water Cooled Lithium Lead (WCLL) Breeder Blanket (BB). In order to validate the performance of the coatings on relevant WCLL BB geometries, a mock-up was designed and will be characterized in an experimental facility operating with flowing lithium-lead, called TRIEX-II. The present work aims to illustrate the preliminary engineering design of a WCLL BB mock-up in order to deeply investigate permeation of hydrogen isotopes through PHTS water pipes. The permeation tests...
Fusion power offers the prospect of an almost inexhaustible source of energy for future generatio... more Fusion power offers the prospect of an almost inexhaustible source of energy for future generations. Although this may be true, it also presents so far serious scientific and engineering challenges. Breeding Blanket is one of key components of future fusion reactors, since it directly involves tritium breeding and energy extraction, both of them critical to development of fusion power. To deal with the development of Breeding Blankets, ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development) designed and built a lead lithium loop (IELLLO: Integrated European Lead Lithium LOop): the experiment aims to support R&D activities for the conceptual and preliminary design of the European test blanket systems. IELLLO was completed, installed and operated at ENEA Brasimone Research Centre. The purpose of this work is to describe the experimental activities that have been carried out concerning the qualification of the facility, and relative instrumentat...
Abstract The accurate prediction of tritium inventory and permeation fluxes in the breeding blank... more Abstract The accurate prediction of tritium inventory and permeation fluxes in the breeding blanket of a D-T fusion reactor is a key aspect for future thermonuclear power plants licensing. Tritium permeation into structural materials could give rise to potential issues concerning the fuel self-sufficiency and can be lost into the environment with resulting radiological risks for the population. In the frame of hydrogen isotopes transport modelling, the Tritium Migration Analysis Program (TMAP) code is considered a referred code for the safety design of nuclear fusion power plants. It is mainly applied to Plasma Facing Components (PFCs), which are subjected to intense particles implantation and surface heat fluxes. Due to some code limitations and to the need of having a more easy-to-use computer code in a multiphysics framework, in the last years the commercial software COMSOL Multiphysics has been used to assess tritium transport studies as an application to ITER, DEMO and CFETR fusion reactors, but it was never validated against TMAP. Within this paper, a COMSOL-based code named mHIT (multi-trapping Hydrogen Isotopes Transport code) is presented. A set of verification and validation (V&V) problems were addressed, with the aim of substantiating the capabilities of the code in common fusion-relevant experimental set-ups and to include different physics according to the level of detail needed for a given application.
In HCLL (Helium Cooled Lithium Lead) blanket for ITER fusion reactor, an efficient management ope... more In HCLL (Helium Cooled Lithium Lead) blanket for ITER fusion reactor, an efficient management operation of tritium is necessary, in order to minimize the release of tritium to the environment and to improve the economy of the overall balance of the reactor tritium inventory. In this context, optimization techniques for the measurement of tritium in the various stages of the fuel cycle of the reactor must be carried out in parallel with the research and development of processing systems for tritium. For example, the TES, Tritium Extraction System, is a critical system and the determination of the concentration of hydrogen isotopes is required for its correct design [1]. Therefore, the need of a fast and reliable sensor for the measurement of the tritium concentration in the liquid breeder is well established. For this application, a sensor based on the principle of tritium permeation in a capsule of suitable material seems to be, at least in principle, the easiest and most reliable one. Furthermore, many data on tritium permeability through different materials and their compatibility in an environment of liquid metal are available, thanks to the experimental studies performed for the selection of structural materials for blanket modules
An analytical study, corroborated by Monte Carlo simulations, is presented which describes the in... more An analytical study, corroborated by Monte Carlo simulations, is presented which describes the interaction of a 300 keV–830 mA proton beam with a Cu and Al dumping system. The analytical calculations rely on the theoretical framework of the Particle Induced X-ray Emission, while the Monte Carlo simulations are performed by means of the GEANT4 toolkit. The case study is related to the project SORGENTINA-RF fusion neutron source and in particular to the tests devoted to assess the performance of the ion source of the plant. The results provide a detailed physical insight of the main processes occurring in the beam dump material, and are also important to give some realistic figures of the radiation emission expected during operation.
The latest progress in the design of the water-cooled lithium–lead (WCLL) tritium extraction and ... more The latest progress in the design of the water-cooled lithium–lead (WCLL) tritium extraction and removal (TER) system for the European DEMO tokamak reactor is presented. The implementation and optimization of the conceptual design of the TER system are performed in order to manage the tritium concentration in the LiPb and ancillary systems, to control the LiPb chemistry, to remove accumulated corrosion and activated products (in particular, the helium generated in the BB), to store the LiPb, to empty the BB segments, to shield the equipment due to LiPb activation, and to accommodate possible overpressure of the LiPb. The LiPb volumes in the inboard (IB) and outboard (OB) modules of the BB are separately managed due to the different pressure drops and required mass flow rates in the different plasma operational phases. Therefore, the tritium extraction is managed by 6 LiPb loops: 4 loops for the OB segments and 2 loops for the IB segments. Each one is a closed loop with forced circul...
A study devoted to assessing the neutronic performance of the SORGENTINA-RF fusion neutron source... more A study devoted to assessing the neutronic performance of the SORGENTINA-RF fusion neutron source is here presented. The calculation is performed considering the characteristics of the accelerated mixed deuterium/tritium ion beam from the ion source, the transport of ions into the titanium layer that constitutes the ion beam target, their implantation and the fusion reaction cross section. An investigation has been done by varying the ion composition of the beam in order to assess if and to what extent diatomic and triatomic ion species affect the neutron emission rate of the plant. This calculation is also instrumental to evaluate the cost-effectiveness of equipping the ion source with a filtering system to remove secondary ions from the beam.
The SORGENTINA-RF project aims to develop a 14 MeV fusion neutron source to produce medical radio... more The SORGENTINA-RF project aims to develop a 14 MeV fusion neutron source to produce medical radioisotopes with special focus on $$^{99}$$ 99 Mo. The facility is based on a positive ion source with an acceleration stage to produce a deuterium (D$$^{+})$$ + ) and tritium (T$$^{+})$$ + ) ion beam that, impacting on a titanium-coated rotating target, allows fusion reactions to take place. Maximizing the neutron production rate is one of the main issues to be addressed in the project and the optimization of some key parameters of the ion beam is of paramount importance in this regard. In this study, a methodology is discussed to reach a definition of the beam characteristics for an effective and sustainable operation of the plant. The most convenient layout that has been found out is based on a single ion source fed by a deuterium and tritium gas mixture. Eventually, a series of considerations about the operation of the ion source and fuel cycle have been drawn.
The SORGENTINA-RF project aims at developing a 14 MeV fusion neutron source featuring an emission... more The SORGENTINA-RF project aims at developing a 14 MeV fusion neutron source featuring an emission rate in the order of 5–7 × 1013 s−1. The plant relies on a metallic water-cooled rotating target and a deuterium (50%) and tritium (50%) ion beam. Beyond the main focus of medical radioisotope production, the source may represent a multi-purpose neutron facility by implementing a series of neutron-based techniques. Among the different engineering and technological issues to be addressed, the production of incondensable gases and corrosion product into the rotating target deserves a dedicated investigation. In this study, a preliminary analysis is carried out, considering the general layout of the target and the present choice of the target material.
The experimental qualification of the tritium extraction unit (TEU) from the LiPb eutectic alloy ... more The experimental qualification of the tritium extraction unit (TEU) from the LiPb eutectic alloy (15.7 at. % Li), the breeder material of the water-cooled lithium–lead breeding blanket concept, is one of the fundamental items for the demonstration of tritium balance sustainability for ITER and DEMO fusion reactors. Several technologies have been proposed as TEU, but the selection of the reference technology can be carried out only after the experimental measurement of the tritium extraction efficiency. For this purpose, a dedicated facility, called TRIEX-II, was designed and installed at ENEA Brasimone Research Centre, Italy. The facility is able to qualify gas–liquid contactor (GLC), permeator against vacuum and liquid-vacuum contactor technologies at different temperatures, lithium–lead mass flow rates and hydrogen isotopes concentrations. In TRIEX-II, the hydrogen or deuterium, used to simulate tritium, are solubilised inside the LiPb with a dedicated saturator and are then extra...
Abstract A crucial issue for the design of a fission Generation IV Sodium Fast Reactor (SFR) is t... more Abstract A crucial issue for the design of a fission Generation IV Sodium Fast Reactor (SFR) is to efficiently characterize the processes and the technical solutions devoted to the management and the control of tritium transfers and releases occurring in SFR circuits. In fact, the tritium balance in sodium fast reactor is a key aspect and the determination of the hydrogen isotopes (hydrogen/tritium) concentration in the liquid sodium has to be investigated for the operational control of tritium balance in the reactor. Therefore, the need for a fast and reliable sensor for the online measurement of hydrogen isotopes concentration in sodium is evident. In the present work, a new concept of hydrogen isotopes permeation sensor (HPS) is shown. The physical operating principle of the HPS is the permeation of hydrogen isotopes throughout a thin metallic membrane. Since the limit of detection and the accuracy of the integrated sensor has to be compatible with the range of tritium concentrations in sodium circuits, it is necessary for the sensor to be operated in high vacuum (HV) conditions in order to guarantee hydrogen isotopes measurements down to 10−5 Pa or even lower. The paper presents two designs foreseen not only for SFRs, but also for an experimental sodium loop where higher H 2 concentrations will be adopted.
Tritium permeation into the Primary Heat Transfer System (PHTS) of DEMO and ITER reactors is one ... more Tritium permeation into the Primary Heat Transfer System (PHTS) of DEMO and ITER reactors is one of the challenging issues to be solved in order to demonstrate the feasibility of nuclear fusion power plants construction. Several technologies were investigated as antipermeation and corrosion barriers to reduce the tritium permeation flux from the breeder into the PHTS. Within this frame, alumina coating manufactured by Pulsed Laser Deposition (PLD) and Atomic Layer Deposition (ALD) are two of the main candidates for the Water Cooled Lithium Lead (WCLL) Breeder Blanket (BB). In order to validate the performance of the coatings on relevant WCLL BB geometries, a mock-up was designed and will be characterized in an experimental facility operating with flowing lithium-lead, called TRIEX-II. The present work aims to illustrate the preliminary engineering design of a WCLL BB mock-up in order to deeply investigate permeation of hydrogen isotopes through PHTS water pipes. The permeation tests...
Fusion power offers the prospect of an almost inexhaustible source of energy for future generatio... more Fusion power offers the prospect of an almost inexhaustible source of energy for future generations. Although this may be true, it also presents so far serious scientific and engineering challenges. Breeding Blanket is one of key components of future fusion reactors, since it directly involves tritium breeding and energy extraction, both of them critical to development of fusion power. To deal with the development of Breeding Blankets, ENEA (Italian National Agency for New Technologies, Energy and Sustainable Economic Development) designed and built a lead lithium loop (IELLLO: Integrated European Lead Lithium LOop): the experiment aims to support R&D activities for the conceptual and preliminary design of the European test blanket systems. IELLLO was completed, installed and operated at ENEA Brasimone Research Centre. The purpose of this work is to describe the experimental activities that have been carried out concerning the qualification of the facility, and relative instrumentat...
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