On the improved current pulse method for the dynamic assessment of thermal diffusive properties

Implementazione ed utilizzo di simulatori semplificati “Desktop”

Experimental tests and thermo-mechanical analyses on the HEXCALIBER mock-up

Abstract Within the framework of the R&D activities promoted by European Fusion Development Agreement on the helium-cooled pebble bed test blanket module to be irradiated in ITER, ENEA Brasimone and the Department of Nuclear Engineering of the University of Palermo performed intense research activities on the modelling of the thermo-mechanical behaviour of both beryllium and lithiated ceramics pebble beds, which are envisaged to be used, respectively, as neutron multiplier and tritium breeder. In particular, at the DIN a thermo mechanical constitutive model was developed for both lithiated ceramics and beryllium pebble beds and it was successfully implemented on a commercial finite element …

ANALYSES OF THE SPES-3 DESIGN AND BEYOND DESIGN ACCIDENT CONDITION BY USING TRACE CODE

Fattibilità di una diversa configurazione della facility SPES-3

Analyses of the OSU-MASLWR Natural Circulation Phenomena By Using TRACE Code

Analysis of a full-scale integral test in PERSEO facility by using TRACE code

Over the last decades a lot of experimental researches have been done to increase the reliability of passive decay heat removal systems implementing in-pool immersed heat exchanger. In this framework, a domestic research program on innovative safety systems was carried out leading the design and the development of the PERSEO facility at the SIET laboratories. The configuration of the system consists of an heat exchanger contained in a small pool which is connected both at the bottom and at the top to a large water reservoir pool. Within the frame of a national research program funded by the Italian minister of economic development, the DEIM department of the University of Palermo in coopera…

MHD Free Convection in Helium-Cooled Lithium-Lead Blanket Modules for the Demonstration Fusion Reactor

TRACE Analysis of the MASLWR Primary/Containment Coupling Phenomena in Beyond Design Accident Scenario

The HELICA mock-up. Experimental results and numerical thermo-mechanical analyses

A Comparison of Direct Numerical Simulation and Turbulence Models for Liquid Metal Free Convection in Volumetrically Heated Enclosures

Neutronic and photonic analysis of the water-cooled Pb17Li test blanket module for ITER-FEAT

Abstract Within the European Fusion Technology Program, the Water-Cooled Lithium Lead (WCLL) DEMO breeding blanket line was selected in 1995 as one of the two EU lines to be developed in the next decade, in particular with the aim of manufacturing a Test Blanket Module (TBM) to be implemented in ITER. This specific goal has been maintained also in ITER-FEAT program even if the general design parameters of the TBMs have reported some changes. This paper is focused on the investigation of the WCLL-TBM nuclear response in ITER-FEAT through detailed 3D-Monte Carlo neutronic and photonic analyses. A 3D heterogeneous model of the most recent design of the WCLL-TBM has been set-up simulating reali…

Analysis of the Multi-Application Small Light-Water Reactor (MASLWR) Design Natural Circulation Phenomena

First Flight Escape Probability and Uncollided Flux of Nuclear Particles in Convex Bodies with Spherical Symmetry

This paper deals with the evaluation of the first flight escape probability of nuclear particles from convex bodies with spherical symmetry by means of some geometrical arguments and very simple probability considerations. The cases of a full sphere, a one-region spherical shell with an empty central zone, a spherical shell region containing a black central zone, and a full sphere with a sourceless shell have been considered. In all the aforementioned cases, a homogeneous medium and uniform isotropic source have been taken into account. Moreover, a simple and general formula has been derived for the calculation of the uncollided flux that is presupposed to be valid for arbitrary geometries.…

Out of Pile Thermo-Mechanical Testing of Breeder Pebble Beds for HCPB TBM for ITER

A constitutive model for the thermo-mechanical behaviour of fusion-relevant pebble beds and its application to the simulation of HELICA mock-up experimental results

Abstract Within the framework of the R&D activities promoted by EFDA on the Helium-Cooled Pebble Bed Test Blanket Module to be irradiated in ITER, attention has been focused on the modelling of the thermo-mechanical behaviour of both beryllium and lithiated ceramic pebble beds that are envisaged to be used respectively as neutron multiplier and tritium breeder. This behaviour depends, mainly, on the reactor-relevant conditions, the pebble sizes and the breeder cell geometries and a general constitutive model has not yet been validated, especially for fusion-relevant applications. ENEA-Brasimone and the Department of Nuclear Engineering (DIN) of the University of Palermo have performed inten…

Final report on the test campaigns HELICA I-II performed in HEFUS facility

Analysis of a Generation 3+ Pressurised Water Reactor plant response to a postulated Station Black Out

Analysis, by RELAP5 code, of Boron Dilution Phenomena in Small Break LOCA and in Mid Loop Operation Transients, Performed in PKL III Test Facility

On the influence of the supporting frame on the nuclear response of the Helium-Cooled Lithium Lead Test Blanket Module for ITER

Abstract Within the European Fusion Technology Programme, very intense research activities have been promoted on the Helium-Cooled Lithium Lead (HCLL) breeding blanket concept with the specific aim of manufacturing a Test Blanket Module (TBM) to be irradiated in ITER. HCLL-TBM is foreseen to be located in an ITER equatorial port, being housed inside a proper steel-supporting frame. In particular, since that frame has been designed to provide two cavities separated by a dividing plate and HCLL-TBM is foreseen to fill just one of them, its nuclear response could vary accordingly to the filling status of the other one, unless the dividing plate is thick enough to isolate the components housed …

A Semi-Theoretical Approach to a Correlation for the Thermal Conductivity of a Beryllium Pebble Bed

In the framework of the European Fusion Technology Programme, Lithium ceramics and Beryllium packed pebble beds are foreseen to be used as Tritium breeders and neutron multipliers, respectively, for the Helium Cooled Pebble Bed breeding blanket of a fusion power reactor operating with a D-T plasma. The present work is focused on the semi-theoretical investigation of the thermal conductivity of single size Beryllium pebble beds, starting from the main hypothesis that this conductivity depends linearly on pebble bed local temperature and total volumetric strain and introducing a method to determine the coefficients of such dependence on the basis of the results obtained by the SUPER-PEHTRA ex…

Mixed magnetohydrodynamic convection in poloidal Helium-Cooled Lithium Lead blanket modules of a fusion reactor

The hydraulic behaviour of the ITER full-scale divertor cassette. The steady state analyses

Study of a water-cooled convective divertor prototype for the DEMO fusion reactor

The plasma facing components of a fusion power reactor have a large impact on the overall plant design, its performance and availability and on the cost of electricity. The present work concerns a study of feasibility for a water-cooled prototype of the convective divertor component of the DEMO fusion reactor. The study has been carried out in two steps. In the first one thermal-hydraulic and neutronic parametric analyses have been performed to find out the prototype optimized configuration. In the second step thermo-mechanical analyses have been carried out on the obtained configuration to investigate the potential and limits of the proposed prototype, with a particular reference to the ma…

Steady state and transient thermal–hydraulic analyses on ITER divertor module

Abstract One of the most challenging components of ITER is the divertor devoted at controlling the characteristics of the plasma boundary, exhausting the α particles and reducing the impurities in the plasma. The thermal–hydraulic design of the divertor is particularly, demanding because of the high heat loads and the cooling flow margin in the plasma-facing components (PFCs). The pressure drop is limited by the pumping power and also avoiding the risk of reaching critical heat flux (CHF). Furthermore, for maintenance operation foreseen, each single divertor cassette should be drained and dried before withdrawing it out from the vacuum vessel. To address these requirements, European Fusion …

Hydraulic characterization of the full scale divertor cassette prototype

In the frame of the activities related to ITER divertor R&D, ENEA C.R. Brasimone was in charge by Fusion For Energy (F4E) to perform the assembly, the hydraulic tests and the theoretical simulation of the hydraulic behavior of the full scale divertor cassette prototype. The objective of these activities was aimed at the investigation of the thermal-hydraulic behavior of the full-scale divertor cassette both under steady state condition and during draining and drying operational transient. In particular, the steady state tests were focused on finally check whether the hydraulic design of the divertor components is able to ensure a uniform and proper cooling for the plasma facing components, …

Post Test Analysis and Accuracy Quantification of PKL III F2.1 RUN1 Test

Effectiveness of mRNA COVID-19 Vaccines in Adolescents Over 6 Months

BACKGROUND AND OBJECTIVES On June 4, 2021, Italy launched the coronavirus disease 2019 (COVID-19) vaccination of adolescents to slow down the COVID-19 spread. Although clinical trials have evaluated messenger ribonucleic acid (mRNA) vaccine effectiveness in adolescents, there is limited literature on its real-world effectiveness. Accordingly, this study aimed to estimate the effectiveness of mRNA COVID-19 vaccines against severe acute respiratory syndrome coronavirus 2 (SARS-CoV-2) infection and mild or severe COVID-19 in a cohort of Sicilian adolescents within a 6 month observation period. METHODS A retrospective cohort study was conducted with adolescents aged 12 to 18 years, residents o…

On the theoretical–numerical study of the HEXCALIBER mock-up thermo-mechanical behaviour

Abstract Within the framework of the R&D activities promoted by European Fusion Development Agreement on the Helium-Cooled Pebble Bed Test Blanket Module to be irradiated in ITER, ENEA Brasimone and the Department of Nuclear Engineering of the University of Palermo (DIN) performed intense research activities on the modelling of the thermo-mechanical behaviour of both beryllium and lithiated ceramics pebble beds, that are envisaged to be used, respectively, as neutron multiplier and tritium breeder. In particular, the DIN developed a thermo-mechanical constitutive model for these pebble beds to be validated against the HEXCALIBER mock-up test campaign, carried out at the ENEA HE-FUS3 facilit…

Post Test Analysis and Accuracy Quantification of PKL III E3.1 Test

Validazione e Verifica (V&V) di CATHARE2 e TRACE sul Programma Sperimentale SPES-2

Preparation of HEXCALIBER tests and preliminary thermo-mechanical analyses

On the hyperporous non-linear elasticity model for fusion-relevant pebble beds

Abstract Packed pebble beds are particular granular systems composed of a large amount of small particles, arranged in irregular lattices and surrounded by a gas filling interstitial spaces. Due to their heterogeneous structure, pebble beds have non-linear and strongly coupled thermal and mechanical behaviours whose constitutive models seem limited, being not suitable for fusion-relevant design-oriented applications. Within the framework of the modelling activities promoted for the lithiated ceramics and beryllium pebble beds foreseen in the Helium-Cooled Pebble Bed breeding blanket concept of DEMO, at the Department of Nuclear Engineering of the University of Palermo (DIN) a thermo-mechani…

Final report on the contract EFDA 682 - Preparation of integration and hydraulic tests of full-scale divertor components.

Analysis of the OSU-MASLWR Natural circulation phenomena using TRACE code

Modifiche del Codice RELAP5 per lo Studio delle Perdite di Carico in Generatori di Vapore a Tubi Elicoidali Interessati da Flussi Bifase

TRACE input model for SPES3 facility

The hydraulic behaviour of the simulacrum of a Plasma Vessel Module of the W 7-X Reactor

Further improvements of the water-cooled Pb–17Li blanket

Abstract The water-cooled lithium–lead (WCLL) blanket is based on reduced-activation ferritic–martensitic steel as the structural material, the liquid alloy Pb–17Li as breeder and neutron multiplier, and water at typical PWR conditions as coolant. It was developed for DEMO specifications and shall be tested in ITER. In 1999, a reactor parameter optimization was performed in the EU which yielded improved specifications of what could be an attractive fusion power plant. Compared to DEMO, such a power reactor would be different in lay-out, size and performance, thus requiring to better exploit the potential of the WCLL blanket concept in conjunction with a water-cooled divertor. Several new ap…

TRACE and RELAP5 Codes for Beyond Design Accident Condition Simulation in the SPES3 Facility

Code validation on qualified experimental data is a fundamental issue in the design and safety analyses of nuclear power plants. The SPES3 facility is being built at the SIET laboratories for an integral type SMR simulation, in the frame of an R&D program on nuclear fission, funded by the Italian Ministry of Economic Development and led by ENEA. The facility, based on the IRIS reactor design, reproduces the primary, secondary and containment systems with 1:100 volume scale, full elevation and prototypical fluid and thermal-hydraulic conditions. It is suitable to test the plant response to design and beyond design accidents in order to verify the effectiveness of the primary and containm…

On the use of an improved current pulse method for the experimental determination of the thermal diffusive properties of packed pebble beds

On the computational assessment of the IFMIF-EVEDA Target Assembly thermal behaviour

On the use of tin–lithium alloys as breeder material for blankets of fusion power plants

Abstract Tin–lithium alloys have several attractive thermo-physical properties, in particular high thermal conductivity and heat capacity, that make them potentially interesting candidates for use in liquid metal blankets. This paper presents an evaluation of the advantages and drawbacks caused by the substitution of the currently employed alloy lead–lithium (Pb–17Li) by a suitable tin–lithium alloy: (i) for the European water-cooled Pb–17Li (WCLL) blanket concept with reduced activation ferritic–martensitic steel as the structural material; (ii) for the European self-cooled TAURO blanket with SiCf/SiC as the structural material. It was found that in none of these blankets Sn–Li alloys woul…

On the numerical assessment of the thermal-hydraulic behaviour of ITER Upper Port Plug Main Body cooling circuit

Study of the helium-cooled lithium lead test blanket module nuclear behaviour under irradiation in ITER

Abstract The present paper deals with the detailed investigation of the helium-cooled lithium lead test blanket module (HCLL-TBM) nuclear behaviour under irradiation in ITER, carried out at the Department of Nuclear Engineering of the University of Palermo adopting a numerical approach based on the Monte Carlo method. A realistic 3D heterogeneous model of the HCLL-TBM was set-up and inserted into an ITER 3D semi-heterogeneous model that realistically simulates the reactor lay-out up to the cryostat. A Gaussian-shaped neutron source was adopted for the calculations. The main features of the HCLL-TBM nuclear response were assessed, paying a particular attention to the neutronic and photonic d…

Analisi Mediante Il Codice TRACE Delle Principali Fenomenologie Caratterizzanti Il Transitorio Conseguente Ad Una Rottura A Ghigliottina Nella Linea DVI dell’Impianto Sperimentale SPES-3

Identificazione Di Componenti Di Piccola Taglia In Sistemi Di Tipo Passivo e Possibili Attività Sperimentali Per La Loro Caratterizzazione

Water-cooled Pb–17Li test blanket module for ITER: Impact of the structural material grade on the neutronic responses

Abstract The Water-Cooled Lithium Lead (WCLL) DEMO blanket is one of the two EU lines to be further developed with the aim of manufacturing by 2010 a Test Blanket Module for ITER (TBM). In this paper results of a 3D-Monte Carlo neutronic analysis of the TBM design are reported. A fully 3D heterogeneous model of the WCLL–TBM has been inserted into an existing ITER model accounting for a proper D–T neutron source. The structural material assumed for the calculations was martensitic 9% Cr steel code named Z 10 CDV Nb 9-1. Results have been compared with those obtained using MANET. The main nuclear responses of the TBM have been determined, such as detailed power deposition density, material da…

Mixed MHD convection and Tritium transport in fusion-relevant configurations

Mixed MHD flow and Tritium transport were computed for a slender poloidal duct, representative of a DEMO HCLL blanket element. 2-D flow and temperature fields were computed in the duct's cross section under the assumption of parallel, fully developed flow, while Tritium concentration C was found by solving a fully 3-D problem with simplifying assumptions at the duct's ends. The spatial distribution of C depended on the intensity and direction of the forced flow. Significant peak factors were obtained if the net flow rate was so low that re-circulation occurred; C maxima were attained near the walls for upward flow, in the core region for downward flow.

Piano di emergenza per il reattore nucleare di ricerca AGN-201 COSTANZA dell’Università di Palermo: presupposti tecnici e valutazioni radiologiche

On the nuclear response of the helium-cooled lithium lead test blanket module in ITER

Abstract The helium-cooled lithium lead (HCLL) concept has been recently selected as one of the two European reference designs foreseen for the breeding blanket of a demonstration fusion reactor. In particular, within the framework of the research and development activities on this blanket line, an HCLL test blanket module (TBM) has to be designed and manufactured to be implemented in ITER. At the Department of Nuclear Engineering (DIN) of the University of Palermo, a research campaign has been carried out to investigate the nuclear response of HCLL-TBM inside ITER by a numerical approach based on the Monte Carlo method. A realistic 3D heterogeneous model of HCLL-TBM has been set-up and ins…

ANALYSES OF TRACE-PARCS COUPLING CAPABILITY

IAEA International Collaborative Standard Problem on Integral PWR Design Natural Circulation Flow Stability and Thermo-hydraulic Coupling of Containment and Primary System during Accidents - Double Blind Calculation Results

The hydraulic behaviour of the prototype of on a ITER divertor cassette

On the Improved Current Pulse method for the thermal diffusive characterization of lithiated ceramic pebble beds

Abstract Packed pebble beds are granular systems composed of small particles generally arranged in irregular lattices and surrounded by a gas filling their interstitial spaces. They show non-linear and coupled thermal and mechanical behaviours, which are under theoretical and experimental investigation to set-up a realistic constitutive model to be adopted for design-oriented purposes. At the Department of Nuclear Engineering (DIN) of the University of Palermo a realistic constitutive model of fusion-relevant pebble beds thermo-mechanical behaviour was developed adopting a “continuous” approach, based on the assumption that a pebble bed could be considered as a continuous, homogeneous and i…

On the effects of the supporting frame on the radiation-induced damage of HCLL-TBM structural material

Within the European Fusion Technology Programme, research activities have been conducted on the Helium-Cooled Lithium Lead (HCLL) breeding blanket concept with the aim of manufacturing a Test Blanket Module (TBM) to be irradiated in ITER. HCLL-TBM is planned to be located in an ITER equatorial port, housed inside an AISI 316 stainless steel-supporting frame. Since that frame has been designed to provide two positions separated by a dividing plate and the HCLL-TBM is expected to fill one of them, its nuclear response could vary depending on the filling status of the other position and on the plate thickness. A parametric study has been carried out to investigate the potential effects on the …

On the numerical assessment of the thermo-mechanical performances of the DEMO Helium-Cooled Pebble Bed breeding blanket module

Within the framework of the European DEMO Breeder Blanket Programme, a research campaign has been launched by University of Palermo, ENEA-Brasimone and Karlsruhe Institute of Technology to theoretically investigate the thermo-mechanical behavior of the Helium-Cooled Pebble Bed (HCPB) breeding blanket module of the DEMO1 blanket vertical segment, under normal operation and over-pressurization loading scenarios. The research campaign has been carried out following a theoretical-computational approach based on the finite element method (FEM) and adopting a qualified commercial FEM code. A realistic 3D FEM model of the HCPB blanket module central poloidal-radial region has been developed, inclu…

Steady state and transient thermal-hydraulic characterization of full-scale ITER divertor plasma facing components

Abstract In the frame of the activities related to ITER divertor R&D, ENEA CR Brasimone was in charge by EFDA (European Fusion Development Agreement) to investigate the thermal-hydraulic behaviour of the full-scale divertor plasma facing components, i.e. the outer vertical target, the inner vertical target and the dome-liner, both in steady state and during draining and drying transient. The investigation was performed by means of both experimental test campaigns performed at ENEA CR Brasimone and theoretical simulation developed in RELAP5 Mod.3.3 environment at the Department of Nuclear Engineering of the University of Palermo (DIN). This paper presents the achieved experimental results fo…

Steady state and transient thermal-hydraulic analyses on ITER divertor cassette

Modifiche del codice RELAP5/MOD3.2.b per lo studio delle perdite di carico e dello scambio termico in condotti elicoidali interessati da deflussi bifase: validazione attraverso gli esperimenti effettuati dal Politecnico di Milano presso l’impianto SIET di Piacenza

L’attività di ricerca svolta nell’ambito della linea progettuale LP2-B.2 del programma PAR 2008-09 (CERSE III) ha visto una prima fase in cui è stato ulteriormente sviluppato il lavoro svolto nel corso dei precedenti programmi CERSE [1, 2], riguardante la validazione del codice termoidraulico avanzato Relap5/Mod3.2.b, modificato per il calcolo delle cadute di pressione in tubi elicoidali interessati da deflussi monofase e bifase, ed una seconda fase che ha comportato l’implementazione di nuove procedure valide per lo studio dello scambio termico bifase in condotti elicoidali, in aggiunta a quelle relative al solo scambio termico monofase, in precedenza implementate. Per quanto riguarda ques…

Environmental Radioactivity Measurements in the Mediterranean Area

In the framework of an environmental radioactivity monitoring program, a research activity concerning an optimization of techniques for sampling and measurement of environmental radioactivity was performed at the Nuclear Engineering Department (DIN) of Palermo University. Al-though initially aimed to the analysis of monitoring data of the local network installed at the nuclear research reactor AGN-201 "COSTANZA" of Palermo University, in the following years the studies were oriented to improve the sampling facilities and measurement systems with the aim to achieve also detection limits as low as possible. The activity has allowed to collect numerous data of nuclide activity concentration fo…

Assessment of the Possible Lay-Out Influence on the HCLL-TBM Nuclear Response

The Department of Nuclear Engineering of the University of Palermo (DIN) was involved, since several years, in the study of the nuclear response of the helium-cooled lithium lead (HCLL) test blanket module (TBM) which will be tested in ITER. In this framework a research campaign was performed, at the DIN, to asses the nuclear response of the TBM in a toroidal lay-out, with the specific aim to investigate the possible lay-out influence on the module nuclear behaviour by comparing the results obtained with those presented in a similar previous work focussed on the most recent design of the poloidal HCLL-TBM. A computational approach based on the Monte Carlo method was followed and a realistic…

A study of the potential influence of frame coolant distribution on the radiation-induced damage of HCLL-TBM structural material

Abstract Within the European Fusion Technology Programme, the Helium-Cooled Lithium Lead (HCLL) breeding blanket concept is one of the two EU lines to be developed for a Long Term fusion reactor, in particular with the aim of manufacturing a Test Blanket Module (TBM) to be implemented in ITER. The HCLL-TBM is foreseen to be located in an ITER equatorial port, being housed inside a steel-supporting frame, actively cooled by pressurized water. That supporting frame has been designed to house two different TBMs, providing two cavities separated by a dividing Plate 20 cm thick. As the nuclear response of HCLL-TBM might vary accordingly to the supporting frame configuration and composition, at t…

Thermal–mechanical and thermal–hydraulic integrated study of the Helium-Cooled Lithium Lead Test Blanket Module

Abstract The Helium-Cooled Lithium Lead Test Blanket Module (HCLL-TBM) is one of the two TBM to be installed in an ITER equatorial port since day 1 of operation, with the specific aim to investigate the main concept functionalities and issues such as high efficiency helium cooling, resistance to thermo-mechanical stresses, manufacturing techniques, as well as tritium transport, magneto-hydrodynamics effects and corrosion. In particular, in order to show a DEMO-relevant thermo-mechanical and thermal–hydraulic behavior, the HCLL-TBM has to meet several requirements especially as far as its coolant thermofluid-dynamic conditions and its thermal–mechanical field are concerned. The present paper…

IAEA International Collaborative Standard Problem on Integral PWR Design Natural Circulation Flow Stability and Thermo-hydraulic Coupling of Containment and Primary System during Accidents

Education and Research in Nuclear Engineering and Radiological Protection at Nuclear Engineering Department of Palermo University

Mixed MHD Convection and Tritium Transport in HCLL TBM Breeder Units for the ITER Fusion Reactor

Modelling of the thermal mechanical behaviour of a single size beryllium pebble bed

The Helium Cooled Pebble Bed (HCPB) Blanket for fusion power reactors and the ITER breeding blanket are based on the use of pebble beds of lithium ceramics as breeder and beryllium as neutron multiplier. Experimental activities were performed at Forschungszentrum Karlsruhe concerning the measurement of pebble bed heat transfer parameters. At the Department of Nuclear Engineering of the University of Palermo, the experimental results have been reproduced by means of the ABAQUS finite element code. Moreover, a thermal-mechanical theoretical model has been developed for single size beryllium pebble beds. In the paper the results from the numerical and theoretical analyses and the comparison wi…

Adeguamento dello SPES2 per Prove di Sicurezza. Analisi Preliminari per La Simulazione di un Incidente Tipo Fukushima su SPES-2

The ITER divertor cassette. Steady-state characterisation and draining and drying transient hydraulic analyses

A Training Experience of Operators with the AGN-201 “Costanza” Research Reactor of Palermo University

The nuclear reactor AGN-201 named “Costanza”, installed at the Nuclear Engineering Section of the Department of Energy of the University of Palermo, is a “zero power” research reactor designed to be mainly used for education purposes as well as for research applications, such as activation analysis and irradiation tests, and last, but not the least, for radionuclide production to be used in nuclear medical applications. Due to its intrinsic safety and low margin of reactivity (less than 350 p.c.m.) so as to the absence of start-up and shut-down time limits, it represents a useful training tool for operators, too. This paper reports some of the activities carried out within the framework of …

On the investigation of Lithium Methatitanate pebble bed thermal diffusive properties by the Improved Current Pulse method

On the AGN - 201 "COSTANZA" Research Reactor at the Department of Nuclear Engineering of the University of Palermo

Analyses of the OSU-MASLWR Experimental Test Facility

Today, considering the sustainability of the nuclear technology in the energy mix policy of developing and developed countries, the international community starts the development of new advanced reactor designs. In this framework, Oregon State University (OSU) has constructed, a system level test facility to examine natural circulation phenomena of importance to multi-application small light water reactor (MASLWR) design, a small modular pressurized water reactor (PWR), relying on natural circulation during both steady-state and transient operation. The target of this paper is to give a review of the main characteristics of the experimental facility, to analyse the main phenomena characteri…

Analysis of Primary/Containment Coupling Phenomena Characterizing the MASLWR Design During a SBLOCA Scenario

Today considering the world energy demand increase, the use of advanced nuclear power plants, have an important role in the environment and economic sustainability of country energy strategy mix considering the capacity of nuclear reactors of producing energy in safe and stable way contributing in cutting the CO2 emission (Bertel & Morrison, 2001; World Energy Outlook-Executive Summary, 2009; Wolde-Rufael & Menyah, 2010; Mascari et al., 2011d). According to the information’s provided by the “Power Reactor Information System” of the International Atomic Energy Agency (IAEA), today 433 nuclear power reactors are in operation in the world providing a total power installed capacity of 366.610 G…

Analysis of the SPES-3 direct vessel injection line break by using TRACE code

Analysis of a Station Black-Out transient in SMR by using the TRACE and RELAP5 code

The present paper deals with the investigation of the evolution and consequences of a Station Black-Out (SBO) initiating event transient in the SPES3 facility [1]. This facility is an integral simulator of a small modular reactor being built at the SIET laboratories, in the framework of the R&D program on nuclear fission funded by the Italian Ministry of Economic Development and led by ENEA. The SBO transient will be simulated by using the RELAP5 and TRACE nodalizations of the SPES3 facility. Moreover, the analysis will contribute to study the differences on the code predictions considering the different modelling approach with one and/or three-dimensional components and to compare the capa…

Analisi di Sistemi Passivi, Utilizzati in Impianti ad Acqua Pressurizzata di Tipo Avanzato. Identificazione di Componenti di Piccolo Diametro

Transient analysis of "2 inch Direct Vessel Injection line break" in SPES-2 facility by using TRACE code

In the past few decades a lot of theoretical and experimental researches have been done to understand the physical phenomena characterizing nuclear accidents. In particular, after the Three Miles Island accident, several reactors have been designed to handle successfully LOCA events. This paper presents a comparison between experimental and numerical results obtained for the “2 inch Direct Vessel Injection line break” in SPES-2. This facility is an integral test facility built in Piacenza at the SIET laboratories and simulating the primary circuit, the relevant parts of the secondary circuits and the passive safety systems typical of the AP600 nuclear power plant. The numerical analysis her…

A neutron point kinetic model for fusion relevant calculations

Abstract In the framework of research activities on fusion reactors a great effort is dedicated by the scientific community to the development of tritium breeding blankets. One of the main goals is to assess the neutronic behaviour of such devices to analyse their tritium breeding performance and to evaluate the required data for their thermal–mechanic and thermal–hydraulic design. Many papers have been published on this topic considering some stationary condition to calculate such important quantities as heating power, gas production and dpa rates, tritium breeding ratio, etc., but not much attention has been focussed to neutronic transport analyses in transient conditions. The present pap…

Simulazione del Circuito Primario di un Impianto Nucleare LWR Mediante L'uso Integrato Dei Codici PARCS e TRACE

TRACE Code Analyses for the IAEA ICSP on “Integral PWR Design Natural Circulation Flow Stability and Thermo-Hydraulic Coupling of Containment and Primary System During Accidents”

Considering the world energy demand increase in order to fulfill an environmental and economic sustainability, the energy policy of each country has to diversify the sources of energy and use stable, safe energy production option able of producing electricity in a clean way contributing in cutting the CO2 emission. In the framework of the sustainable development, today the use of advanced nuclear power plant, have an important role in the environmental and economic sustainability of country energy strategy. In the last 20 years, in fact, the international community, taking into account the operational experience of the nuclear reactors, starts the development of new advanced reactor designs…

HEXCALIBER TEST SECTION. Thermo-mechanical analysis

Post Test Analisys by Relap5 Code & Accuracy Quantification of PKLIII E3.1 Experiments (Sensitivity calculation)

Applicazione del metodo Monte Carlo a problemi monodimensionali di conduzione termica stazionaria in sistemi con conducibilità dipendente dalla posizione

The hydraulic behaviour of the ITER full-scale divertor cassette. The transient analyses

Sensitivity Analysis of the MASLWR Helical Coil Steam Generator Using TRACE

Accurate simulation of transient system behavior of a nuclear power plant is the goal of systems code calculations, and the evaluation of a code's calculation accuracy is accomplished by assessment and validation against appropriate system data. These system data may be developed either from a running system prototype or from a scaled model test facility, and characterize the thermal hydraulic phenomena during both steady state and transient conditions. The identification and characterization of the relevant thermal hydraulic phenomena, and the assessment and validation of thermal hydraulic systems codes, has been the objective of multiple international research programs. The validation and…

Potential and limits of water-cooled Pb–17Li blankets and divertors for a fusion power plant

Abstract Blankets and divertors are key components of a fusion power plant. They have a large impact on the overall plant design, its performance and availability, and on the cost of electricity. The water-cooled Pb–17Li (WCLL) blanket uses reduced activation ferritic–martensitic steel as structural material. It was previously validated under numerous aspects such as TBR, mechanical and thermo-mechanical stability, thermal–hydraulics, MHD, safety and others. This was done assuming the specifications for a European DEMOnstration reactor which were fixed back in 1989. A WCLL blanket would best be combined with a water-cooled divertor so that a single coolant could be used for the entire react…

Analysis, by RELAP5 code, of Boron Dilution Phenomena in a Mid-Loop Operation Transient, Performed in PKL III F2.1 RUN 1 Test

Progress in the benchmark exercise for analyzing the lithiate breeder pebble bed thermo-mechanical behaviour

The Helium Cooled Pebble Bed (HCPB) Blanket is one of the reference concepts for the European Breeding Blanket Programme for DEMO. In the reference blanket module, alternate layers of lithiated ceramics and beryllium pebbles act, respectively, as tritium breeder and neutron multiplier. The thermo-mechanical behaviour of both the pebble beds and their performances in reactor relevant conditions are also dependent on the pebble size and cell geometries (bed thickness, pebble packing factor, bed thermal conductivity). Therefore, in the EU Fusion Technology Programme, several out-of-pile experimental test campaigns have been performed to determine these behaviours. Theoretical calculations have…

Natural Convection in Liquid Metal-Filled Rectangular Enclosures with Volumetric Heating

Passive safety systems in view of sustainable development

Studies on the AGN - 201 "COSTANZA" Research Reactor

Experimental tests on Li-ceramic breeders for the helium cooled pebble bed (HCPB) blanket design

The Helium Cooled Pebble Bed (HCPB) Test Blanket Module (TBM) to be tested in ITER (International Thermonuclear Experimental Reactor) Reactor foresees the utilization of Lithiate ceramics as Tritium breeder in form of pebble beds. Since 1998, ENEA has launched many experimental activities for the evaluation of the breeder thermomechanics and the interaction between the pebble beds and the prismatic steel containment walls. Main objectives of these activities are the measurement of the pebble bed effective thermal conductivity, the wall heat transfer coefficient, the pressure loads and deformations on the lateral walls and their dependency from the mechanical constraints. The paper presents …

A study of the potential influence of frame coolant on HCLL-TBM nuclear response

Abstract Within the European Fusion Technology Programme, the Helium-Cooled Lithium Lead (HCLL) breeding blanket concept is one of the two EU lines to be developed for a long term fusion reactor, in particular with the aim of manufacturing a test blanket module (TBM) to be implemented in ITER. The HCLL-TBM is foreseen to be located in an ITER equatorial port, being housed inside a steel-supporting frame, actively cooled by pressurized water. This supporting frame has been designed to house two different TBMs providing two cavities separated by a dividing plate 20 cm thick. As the nuclear response of HCLL-TBM could vary with the supporting frame configuration and composition, a parametric st…

Sezione di prova HELICA. Analisi termomeccaniche

Analyses of the TRACE V5 capability for the simulation of natural circulation and primary/containment coupling in BDBA condition typical of the MASLWR

In the short term period the use of advanced Small Modular Reactor (SMR) is one of the most promising options for the deployment of nuclear technology. The validation and assessment of the best estimate thermal hydraulic system code TRACE against SMR thermal hydraulic phenomena is a novel effort. In this framework the use of the natural circulation database developed at the OSU-MASLWR test facility, simulating the MASLWR reactor prototype, is of interest for analyses of the TRACE code capability in predicting natural circulation and primary/containment coupled behavior in SMR. The target of this paper is to analyze the TRACE V5 capability for the simulation of natural circulation phenomena,…

Analyses Of The SPES-3 Accident Condition By Using TRACE Code

Neutronic and photonic analysis of the single box water-cooled lithium lead blanket for a DEMO reactor

Abstract The water-cooled Pb–17Li demonstration plant (DEMO) breeding blanket line was selected in 1995 as one of the two EU lines to be further developed in the next decade. In this paper the results of a neutronic and photonic analysis of the `single box' concept is presented. A full three-dimensional model, including the whole assembly and many of the DEMO reactor components, has been developed, together with a three-dimensional neutron source. A tritium breeding ratio (TBR) value of 1.16, with no ports and a Li6 enrichment of 90%, has been obtained and a further analysis has been performed to determine Li6 enrichment that would still ensure tritium breeding self-sufficiency. Selected po…

On the nuclear response of the water-cooled Pb-17Li test blanket module for ITER-FEAT

Abstract Within the European Fusion Technology Programme, the Water-Cooled Lithium Lead (WCLL) DEMO breeding blanket line was selected in 1995 as one of the two EU lines to be developed in the next decades, in particular with the aim of manufacturing a Test Blanket Module (TBM) to be tested in ITER-FEAT. The present paper is focused on the study of the WCLL-TBM nuclear response in ITER-FEAT, being specifically oriented to the investigation of the local effects due to the typical C-shaped tubes of the breeder zone, since they could play a pivotal role in the module-relevant thermo–mechanical design. A 3D heterogeneous model of the WCLL-TBM, realistically simulating its new lay out and taking…

Effectiveness of mRNA COVID-19 Vaccination on SARS-CoV-2 Infection and COVID-19 in Sicily over an Eight-Month Period

In order to reduce the spread of SARS-CoV-2 infection and the burden of disease, since 27 December 2020, Sicily has introduced a regional COVID-19 vaccination campaign. This study aimed at estimating the effectiveness of mRNA COVID-19 vaccines on SARS-CoV-2 infections and COVID-19. A retrospective cohort study was carried out on 3,966,976 Sicilian adults aged 18 years or more, who were followed-up from 1 January 2021 to 30 September 2021. The risk of SARS-CoV-2 infection, severe COVID-19, and COVID-19 death or intubation during the study period was compared among vaccinated with two mRNA doses and unvaccinated individuals. Cox regression, adjusted for age and sex, and a joint-point analysis…

Analisi Incidentali Deterministiche e Utilizzo di Simulatori di Impianto a Supporto delle Verifiche di Sicurezza. Sviluppo e Messa a Punto di un Modello di un Impianto PWR (EPR like) per Preliminari Analisi con il Codice TRACE di Eventi di Station Blackout