I. Ricapito - Academia.edu (original) (raw)

Papers by I. Ricapito

Research paper thumbnail of An Integral Methodological Framework for the Thermo-Mechanical Analysis and Structural Integrity Assessment of the European Tbm Sets

Research paper thumbnail of Tritium processing systems for the helium cooled pebble bed test blanket module

Fusion Engineering and Design, 2008

Research paper thumbnail of Experimental Investigation of ZrCo Getter Beds as Candidate Process for the Tritium Extraction Systems of the European Test Blanket Modules

Fusion Science and Technology, 2017

Abstract ZrCo is a well-known tritium storage material and has been studied intensively in the li... more Abstract ZrCo is a well-known tritium storage material and has been studied intensively in the literature. The most interesting properties with regards to the thermodynamics of the ZrCo-H system are the very low H2 partial pressure in equilibrium with ZrCoH3 at room temperature and the ease to reach sufficiently high temperature to completely release the stored H2. These properties motivate also to use ZrCo not as a simple storage, but rather as a concentrator of hydrogen isotopologues from inert gases like He. With such function, ZrCo getter beds are the reference solution adopted in the conceptual design of the tritium extraction system of the European Test Blanket Modules (TBM) to replace the cryogenic molecular sieve bed previously proposed. An experimental campaign was carried out on ZrCo in order to consolidate this choice. The results confirmed that ZrCo performs well as getter material but only substantially below the maximum loading capacity. They revealed that the dynamic thermo-mechanical response of the material, controlled by temperature and H2 concentration, is the main limiting factor for the component performance.

Research paper thumbnail of Preliminary assessment of tritium permeation and retention in the European Water Cooled Lithium Lead Test Blanket Module with TESSIM-X

Nuclear Materials and Energy

Research paper thumbnail of Consolidaded Design of the HCPB Breeding Blanket for the pre-Conceptual Design of the EU DEMO and Haermonization with the ITER HCPB TBM Program

Research paper thumbnail of On the numerical and experimental characterisation of the ITER divertor cassette plasma facing component steady state hydraulic behaviour

Research paper thumbnail of Infragilimento da idrogeno di acciai per reattori a fusione

Research paper thumbnail of Supporting analysis for WCLL test blanket system safety

Fusion Engineering and Design, 2021

Abstract The Water-Cooled Lithium Lead Breeding Blanket is one of the most promising concepts to ... more Abstract The Water-Cooled Lithium Lead Breeding Blanket is one of the most promising concepts to be used as a key component in fusion power devices. It provides tritium breeding and nuclear power conversion and extraction. Before its integration in a DEMO plant, a testing phase is required to gain data on both thermal and neutronic performances. For this purpose, a Test Blanket System is currently under conceptual design and will be integrated into dedicated ITER equatorial ports. Early insight on the safety performance of the Test Blanket System remains an essential element for integrating into ITER, and accident analysis is one of its critical components. Preliminary safety studies have been performed in the framework of Safety And Environment (SAE) work package of the EUROfusion consortium program to support the design and the integration of the Water-Cooled Lithium Lead Test Blanket System into ITER. At the present conceptual design phase, these studies are tailored to give insight on the effects of some parameters, such as postulated initiating event timing, plasma termination system or valves intervention, and possible safety provision implementation. A MELCOR model of the Water-Cooled Lithium Lead Test Blanket System has been developed, and two accident scenarios have been studied: an ex-vessel LOCA inside the port cell and a loss of flow accident because of pump seizure. The impact of mentioned parameters on accident evolution and consequences is investigated in support of the safety logic definition.

Research paper thumbnail of Magnetohydrodynamic velocity and pressure drop in manifolds of a WCLL TBM

Nuclear Fusion, 2021

In the frame of the EUROfusion breeding blanket research activities, the water cooled lead lithiu... more In the frame of the EUROfusion breeding blanket research activities, the water cooled lead lithium (WCLL) blanket is considered as reference liquid metal blanket design to be tested in ITER and to be used in a DEMO reactor. The design of breeding blankets represents a major challenge for fusion reactor engineering because of the performance requirements and the severe operating conditions in terms of heat load and neutron flux. Liquid metal alloys such as lead-lithium, PbLi, are considered as breeder material, due to their lithium content required for tritium production, and as heat transfer medium because of their large thermal conductivity and the possibility to be operated at high temperature. On the other hand, the motion of the electrically conducting breeder in the plasma-confining magnetic field induces electric currents and generates strong electromagnetic forces that modify significantly the velocity distribution in the blanket compared to hydrodynamic conditions and increase pressure losses. Magnetohydrodynamic (MHD) pressure drops have to be carefully quantified, since excessive values can jeopardize the feasibility of the considered blanket concept. The present work investigates numerically liquid metal MHD flows in manifolds of a WCLL test blanket module. Velocity and pressure distributions are analyzed.

Research paper thumbnail of An overview of the EU breeding blanket design strategy as an integral part of the DEMO design effort

Fusion Engineering and Design, 2019

Research paper thumbnail of Experimental investigation on HCLL-TBS In-box LOCA

Fusion Engineering and Design, 2018

Abstract The experimental facility THALLIUM was designed and installed at ENEA C.R. Brasimone to ... more Abstract The experimental facility THALLIUM was designed and installed at ENEA C.R. Brasimone to investigate the consequence of a HCLL-TBS In box LOCA, ensuring the geometrical relevance for HCLL-TBM. Within the framework of the contractual activities with Fusion for Energy, a first experimental campaign was carried out in HCLL-TBS relevant conditions. The experiments simulated a rupture in a cooling plate of the HCLL-TBM. Thereafter, a new experimental campaign was developed by installing a larger injection valve, with the aim of analyzing a worse accidental He injection. The experimental parameters that were varied during the first campaign were frozen, varying instead injection time and pressure. The pressure trend in different points of the facility and the effectiveness of the mitigation strategy are the most important outcomes of the experiments. A different behavior with respect to the first campaign was observed. In particular, the first pressure peaks are smaller as a percentage of the injection pressure and the pressure is increasing during the whole transient. The relief valve on the expansion tank drives the transient, confirming the importance of the design of this component to mitigate the effects of an In-box LOCA. The outcomes of this experimental investigation are not only supporting the design of HCLL-TBS but also provide important data in view of the conceptual design of HCLL breeding blanket for DEMO.

Research paper thumbnail of Tritium in Fusion: R&D in the EU

Fusion Science and Technology, 2008

Tritium as one of the two fuel components for fusion power plays a special role in any fusion dev... more Tritium as one of the two fuel components for fusion power plays a special role in any fusion device. Due to its volatile character, radioactivity and easy incorporation as HTO it needs to be controlled with special care and due to its scarcity on earth it has to be produced in-situ in future fusion power plants. The paper discusses the present tritium R&D activities in fusion ongoing in the EU and presents the various processes/techniques envisaged for controlling tritium in future fusion reactors focusing mainly on the issues of breeding blankets and the fuel cycle in DEMO.

Research paper thumbnail of The Adsorption of Gas Mixtures in Zeolites: Experimental Determination and Theoretical Prediction of Equilibria

Fusion Technology 1992, 1993

ABSTRACT

Research paper thumbnail of Status of the EU R&D programme on the blanket-shield modules for ITER

Fusion Engineering and Design, 2008

A research and development (R&amp... more A research and development (R&D) programme for the ITER blanket-shield modules has been implemented in Europe to provide input for the design and the manufacture of the full-scale production components. It involves in particular the fabrication and testing of mock-ups and full-scale prototypes of shield blocks and first wall (FW) panels. This paper summarises the main achievements obtained so far

Research paper thumbnail of Sensitivity Study for Tritium Permeation in Helium-Cooled Lead-Lithium DEMO Blanket with the FUS-TPC Code

Fusion Science and Technology

Abstract Hydrogen dissolves in and permeates through most materials, thus it is important to unde... more Abstract Hydrogen 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 lead-lithium 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.

Research paper thumbnail of Tritium management in the European test blanket systems and extrapolation to DEMO

One of the main objectives of the experimental campaign of the TBS (Test Blanket Systems) in ITER... more One of the main objectives of the experimental campaign of the TBS (Test Blanket Systems) in ITER is the efficient management and accurate accountancy of tritium from its source, the Test Blanket Module (TBM), up to its final routing to the ITER inner fuel cycle system. Indeed, the data collected by the tritium accountancy system, interpreted through comprehensive modelling tools, will be one of the most relevant outcomes in support of the blanket design for DEMO and beyond. This paper describes various aspects of HCLL (Helium Cooled Lithium Lead) and HCPB (Helium Cooled Pebble Bed)-TBS activities that have a direct impact on tritium management systems and then discusses their potential extrapolation in support of DEMO design. It includes: i) the design baseline of TBS sub-systems, also in light of new interface requirements coming from ITER Organization, relevant to the tritium management, focusing on the components potentially DEMO relevant; ii) the preliminary design of Tritium A...

Research paper thumbnail of Tritium Extraction from Liquid Pb-16Li: A Critical Review of Candidate Technologies for ITER and DEMO Applications

Fusion Science and Technology

Abstract Extraction of tritium from liquid lead lithium eutectic alloy is a key topic for the fea... more Abstract Extraction of tritium from liquid lead lithium eutectic alloy is a key topic for the feasibility of any PbLi based tritium breeding blanket (BB). Particularly in DEMO, high tritium extraction efficiency will be required in order to keep low the tritium concentration in the Pb-16Li loop. This is essential to minimize tritium release into the environment and tritium permeation from BB into the primary cooling system. In addition, the tritium extraction process needs to be highly reliable in order not to impact negatively on the operation of the whole fusion reactor, ITER or DEMO. In the present paper, a critical review of the main candidate technologies for tritium extraction from Pb-16Li, particularly gas liquid contactors and vacuum permeators, is accomplished. The intrinsic limits and possible advantages of these technologies are presented and discussed, in the light of considerations coming directly from mathematical models describing their behaviour as well as from the experimental results so far achieved. Needs in terms of R&D activities are identified.

Research paper thumbnail of Conceptual design of the main Ancillary Systems of the ITER Water Cooled Lithium Lead Test Blanket System

Fusion Engineering and Design

Research paper thumbnail of Consolidated design of the HCPB Breeding Blanket for the pre-Conceptual Design Phase of the EU DEMO and harmonization with the ITER HCPB TBM program

Fusion Engineering and Design

Research paper thumbnail of Thermal-hydraulic modeling and analysis of the Water Cooling System for the ITER Test Blanket Module

Fusion Engineering and Design

Research paper thumbnail of An Integral Methodological Framework for the Thermo-Mechanical Analysis and Structural Integrity Assessment of the European Tbm Sets

Research paper thumbnail of Tritium processing systems for the helium cooled pebble bed test blanket module

Fusion Engineering and Design, 2008

Research paper thumbnail of Experimental Investigation of ZrCo Getter Beds as Candidate Process for the Tritium Extraction Systems of the European Test Blanket Modules

Fusion Science and Technology, 2017

Abstract ZrCo is a well-known tritium storage material and has been studied intensively in the li... more Abstract ZrCo is a well-known tritium storage material and has been studied intensively in the literature. The most interesting properties with regards to the thermodynamics of the ZrCo-H system are the very low H2 partial pressure in equilibrium with ZrCoH3 at room temperature and the ease to reach sufficiently high temperature to completely release the stored H2. These properties motivate also to use ZrCo not as a simple storage, but rather as a concentrator of hydrogen isotopologues from inert gases like He. With such function, ZrCo getter beds are the reference solution adopted in the conceptual design of the tritium extraction system of the European Test Blanket Modules (TBM) to replace the cryogenic molecular sieve bed previously proposed. An experimental campaign was carried out on ZrCo in order to consolidate this choice. The results confirmed that ZrCo performs well as getter material but only substantially below the maximum loading capacity. They revealed that the dynamic thermo-mechanical response of the material, controlled by temperature and H2 concentration, is the main limiting factor for the component performance.

Research paper thumbnail of Preliminary assessment of tritium permeation and retention in the European Water Cooled Lithium Lead Test Blanket Module with TESSIM-X

Nuclear Materials and Energy

Research paper thumbnail of Consolidaded Design of the HCPB Breeding Blanket for the pre-Conceptual Design of the EU DEMO and Haermonization with the ITER HCPB TBM Program

Research paper thumbnail of On the numerical and experimental characterisation of the ITER divertor cassette plasma facing component steady state hydraulic behaviour

Research paper thumbnail of Infragilimento da idrogeno di acciai per reattori a fusione

Research paper thumbnail of Supporting analysis for WCLL test blanket system safety

Fusion Engineering and Design, 2021

Abstract The Water-Cooled Lithium Lead Breeding Blanket is one of the most promising concepts to ... more Abstract The Water-Cooled Lithium Lead Breeding Blanket is one of the most promising concepts to be used as a key component in fusion power devices. It provides tritium breeding and nuclear power conversion and extraction. Before its integration in a DEMO plant, a testing phase is required to gain data on both thermal and neutronic performances. For this purpose, a Test Blanket System is currently under conceptual design and will be integrated into dedicated ITER equatorial ports. Early insight on the safety performance of the Test Blanket System remains an essential element for integrating into ITER, and accident analysis is one of its critical components. Preliminary safety studies have been performed in the framework of Safety And Environment (SAE) work package of the EUROfusion consortium program to support the design and the integration of the Water-Cooled Lithium Lead Test Blanket System into ITER. At the present conceptual design phase, these studies are tailored to give insight on the effects of some parameters, such as postulated initiating event timing, plasma termination system or valves intervention, and possible safety provision implementation. A MELCOR model of the Water-Cooled Lithium Lead Test Blanket System has been developed, and two accident scenarios have been studied: an ex-vessel LOCA inside the port cell and a loss of flow accident because of pump seizure. The impact of mentioned parameters on accident evolution and consequences is investigated in support of the safety logic definition.

Research paper thumbnail of Magnetohydrodynamic velocity and pressure drop in manifolds of a WCLL TBM

Nuclear Fusion, 2021

In the frame of the EUROfusion breeding blanket research activities, the water cooled lead lithiu... more In the frame of the EUROfusion breeding blanket research activities, the water cooled lead lithium (WCLL) blanket is considered as reference liquid metal blanket design to be tested in ITER and to be used in a DEMO reactor. The design of breeding blankets represents a major challenge for fusion reactor engineering because of the performance requirements and the severe operating conditions in terms of heat load and neutron flux. Liquid metal alloys such as lead-lithium, PbLi, are considered as breeder material, due to their lithium content required for tritium production, and as heat transfer medium because of their large thermal conductivity and the possibility to be operated at high temperature. On the other hand, the motion of the electrically conducting breeder in the plasma-confining magnetic field induces electric currents and generates strong electromagnetic forces that modify significantly the velocity distribution in the blanket compared to hydrodynamic conditions and increase pressure losses. Magnetohydrodynamic (MHD) pressure drops have to be carefully quantified, since excessive values can jeopardize the feasibility of the considered blanket concept. The present work investigates numerically liquid metal MHD flows in manifolds of a WCLL test blanket module. Velocity and pressure distributions are analyzed.

Research paper thumbnail of An overview of the EU breeding blanket design strategy as an integral part of the DEMO design effort

Fusion Engineering and Design, 2019

Research paper thumbnail of Experimental investigation on HCLL-TBS In-box LOCA

Fusion Engineering and Design, 2018

Abstract The experimental facility THALLIUM was designed and installed at ENEA C.R. Brasimone to ... more Abstract The experimental facility THALLIUM was designed and installed at ENEA C.R. Brasimone to investigate the consequence of a HCLL-TBS In box LOCA, ensuring the geometrical relevance for HCLL-TBM. Within the framework of the contractual activities with Fusion for Energy, a first experimental campaign was carried out in HCLL-TBS relevant conditions. The experiments simulated a rupture in a cooling plate of the HCLL-TBM. Thereafter, a new experimental campaign was developed by installing a larger injection valve, with the aim of analyzing a worse accidental He injection. The experimental parameters that were varied during the first campaign were frozen, varying instead injection time and pressure. The pressure trend in different points of the facility and the effectiveness of the mitigation strategy are the most important outcomes of the experiments. A different behavior with respect to the first campaign was observed. In particular, the first pressure peaks are smaller as a percentage of the injection pressure and the pressure is increasing during the whole transient. The relief valve on the expansion tank drives the transient, confirming the importance of the design of this component to mitigate the effects of an In-box LOCA. The outcomes of this experimental investigation are not only supporting the design of HCLL-TBS but also provide important data in view of the conceptual design of HCLL breeding blanket for DEMO.

Research paper thumbnail of Tritium in Fusion: R&D in the EU

Fusion Science and Technology, 2008

Tritium as one of the two fuel components for fusion power plays a special role in any fusion dev... more Tritium as one of the two fuel components for fusion power plays a special role in any fusion device. Due to its volatile character, radioactivity and easy incorporation as HTO it needs to be controlled with special care and due to its scarcity on earth it has to be produced in-situ in future fusion power plants. The paper discusses the present tritium R&D activities in fusion ongoing in the EU and presents the various processes/techniques envisaged for controlling tritium in future fusion reactors focusing mainly on the issues of breeding blankets and the fuel cycle in DEMO.

Research paper thumbnail of The Adsorption of Gas Mixtures in Zeolites: Experimental Determination and Theoretical Prediction of Equilibria

Fusion Technology 1992, 1993

ABSTRACT

Research paper thumbnail of Status of the EU R&D programme on the blanket-shield modules for ITER

Fusion Engineering and Design, 2008

A research and development (R&amp... more A research and development (R&D) programme for the ITER blanket-shield modules has been implemented in Europe to provide input for the design and the manufacture of the full-scale production components. It involves in particular the fabrication and testing of mock-ups and full-scale prototypes of shield blocks and first wall (FW) panels. This paper summarises the main achievements obtained so far

Research paper thumbnail of Sensitivity Study for Tritium Permeation in Helium-Cooled Lead-Lithium DEMO Blanket with the FUS-TPC Code

Fusion Science and Technology

Abstract Hydrogen dissolves in and permeates through most materials, thus it is important to unde... more Abstract Hydrogen 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 lead-lithium 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.

Research paper thumbnail of Tritium management in the European test blanket systems and extrapolation to DEMO

One of the main objectives of the experimental campaign of the TBS (Test Blanket Systems) in ITER... more One of the main objectives of the experimental campaign of the TBS (Test Blanket Systems) in ITER is the efficient management and accurate accountancy of tritium from its source, the Test Blanket Module (TBM), up to its final routing to the ITER inner fuel cycle system. Indeed, the data collected by the tritium accountancy system, interpreted through comprehensive modelling tools, will be one of the most relevant outcomes in support of the blanket design for DEMO and beyond. This paper describes various aspects of HCLL (Helium Cooled Lithium Lead) and HCPB (Helium Cooled Pebble Bed)-TBS activities that have a direct impact on tritium management systems and then discusses their potential extrapolation in support of DEMO design. It includes: i) the design baseline of TBS sub-systems, also in light of new interface requirements coming from ITER Organization, relevant to the tritium management, focusing on the components potentially DEMO relevant; ii) the preliminary design of Tritium A...

Research paper thumbnail of Tritium Extraction from Liquid Pb-16Li: A Critical Review of Candidate Technologies for ITER and DEMO Applications

Fusion Science and Technology

Abstract Extraction of tritium from liquid lead lithium eutectic alloy is a key topic for the fea... more Abstract Extraction of tritium from liquid lead lithium eutectic alloy is a key topic for the feasibility of any PbLi based tritium breeding blanket (BB). Particularly in DEMO, high tritium extraction efficiency will be required in order to keep low the tritium concentration in the Pb-16Li loop. This is essential to minimize tritium release into the environment and tritium permeation from BB into the primary cooling system. In addition, the tritium extraction process needs to be highly reliable in order not to impact negatively on the operation of the whole fusion reactor, ITER or DEMO. In the present paper, a critical review of the main candidate technologies for tritium extraction from Pb-16Li, particularly gas liquid contactors and vacuum permeators, is accomplished. The intrinsic limits and possible advantages of these technologies are presented and discussed, in the light of considerations coming directly from mathematical models describing their behaviour as well as from the experimental results so far achieved. Needs in terms of R&D activities are identified.

Research paper thumbnail of Conceptual design of the main Ancillary Systems of the ITER Water Cooled Lithium Lead Test Blanket System

Fusion Engineering and Design

Research paper thumbnail of Consolidated design of the HCPB Breeding Blanket for the pre-Conceptual Design Phase of the EU DEMO and harmonization with the ITER HCPB TBM program

Fusion Engineering and Design

Research paper thumbnail of Thermal-hydraulic modeling and analysis of the Water Cooling System for the ITER Test Blanket Module

Fusion Engineering and Design