Luigino PETRIZZI - Academia.edu (original) (raw)

Papers by Luigino PETRIZZI

Research paper thumbnail of SEAFP, Safety and Environmental Assessment of Fusion Power, Final Report

The Safety and Environmental Assessment of Fusion Power (SEAFP) was undertaken for the Commission... more The Safety and Environmental Assessment of Fusion Power (SEAFP) was undertaken for the Commission of the European Union in the framework of the Fusion Programme 1990-94. Its terms of reference were in accordance with the programme decision of the Council of Ministers which followed a request by the European Parliament and a recommendation of the Fusion Programme Evaluation Board. The SEAFP is part of an ongoing effort to consider the safety and environmental aspects of fusion power. SEAFP was carried out by the NET Team, the Euratom/UKAEA Association, and by Industry, with contributions from other Associated Laboratories, the Joint Research Centre and the Canadian fusion programme.

Research paper thumbnail of Activation analysis and characteristics of the European community water cooled ceramic breeder blanket design proposal for ITER

The European Community (EC) Home Team has proposed various alternative blanket designs to the bas... more The European Community (EC) Home Team has proposed various alternative blanket designs to the basic concept (essentially integrated first wall, cooled by liquid metal, with structures made by vanadium alloys). One of the EC proposal is the Water Cooled Ceramic Blanket developed on the basis of a common action between NET and ENEA. It is based on a more conservative approach, but involving well proven technologies and qualified materials: SS-316L as structural material, Li{sub 2}ZrO{sub 3} as first breeder material choice (50% Li{sup 6} enrichment) and low temperature water coolant (160/200{degrees}C). Beryllium has been chosen as multiplying material. The nominal performance are: 1 MW/m{sup 2} as average neutron wall load, corresponding to 1.5 GW fusion power, 1 MW-y/m{sup 2} beneath it has been proved to withstand power excursion till 5 GW. The proposed blanket concept is based on a Breeder Inside Tube (BIT) type technology, with poloidal breeding elements, each one consisting of t...

Research paper thumbnail of EFD-P(13)38 Response of LaBr3(Ce) Scintillators to 2.5MeV Fusion Neutrons

Measurements of the response function of LaBr3(Ce) to 2.5MeV neutrons have been carried out at th... more Measurements of the response function of LaBr3(Ce) to 2.5MeV neutrons have been carried out at the Frascati Neutron Generator and at tokamak facilities with deuterium plasmas. The observed spectrum has been interpreted by means of a MCNP model. It is found that the main contributor to the measured response is neutron inelastic scattering on 79Br, 81Br and 139La. An extrapolation of the count rate response to 14MeV neutrons from deuterium-tritium plasmas is also presented. The results are of relevance for the design of gamma-ray diagnostics of fusion burning plasmas.

Research paper thumbnail of Three-Dimensional Neutronics and Shielding Analyses for the ITER Divertor

Fusion Technology, 1996

3-D neutronics and shielding analyses have been performed for the divertor region of the ITER int... more 3-D neutronics and shielding analyses have been performed for the divertor region of the ITER interim design. The peak neutron wall loading in the divertor region is 0.6 MW/m² at the divertor cassette dome. The total nuclear heating in the 60 divertor cassettes is 102.4 MW. The peak helium production in the VV behind the pumping ducts is 0.5 He

Research paper thumbnail of European Reference Design of the Water-Cooled Lithium-Lead Blanket for a Demonstration Reactor

Fusion Technology, 1994

The water-cooled Pb-17Li blankets represent one of the blanket lines selected within the European... more The water-cooled Pb-17Li blankets represent one of the blanket lines selected within the European Union for DEMO-relevant design and R&D activities. This paper gives a presentation of the reference conceptual design for water-cooled Pb-17Li DEMO blankets and an overview on the results of its performance assessments. Moreover, a critical discussion about the technical aspects requiring further improvements and/or modifications is performed taking into account the present status of the associated R&D. This concept appears to be a very promising candidate for a DEMO reactor breeding blanket.

Research paper thumbnail of The ITER Radial Neutron Camera Detection System

AIP Conference Proceedings, 2008

Research paper thumbnail of Nuclear analysis in support of ITER design

Proceedings of 16th International Symposium on Fusion Engineering

This paper summarizes the main results of nuclear analysis calculations performed during the Inte... more This paper summarizes the main results of nuclear analysis calculations performed during the International Thermonuclear Experimental Reactor (ITER) Engineering Design Activity (EDA). Major efforts were devoted to fulfilling the General Design Requirements to minimize the nuclear heating rate in the superconducting magnets and ensuring that radiation conditions at the cryostat are suitable for hands-on-maintenance after reactor shutdown .

Research paper thumbnail of A Helium-Cooled Ceramic Breeder Blanket Concept with Lead/Graphite as Multiplier/Moderator

Fusion Technology 1990, 1991

Helium-cooled ceramic breeder blanket designs are commonly based on the use of beryllium as both ... more Helium-cooled ceramic breeder blanket designs are commonly based on the use of beryllium as both multiplying and moderating material. The possibility of using lead as multiplier and graphite as moderator, instead of beryllium, is investigated by means of a proper optimization code, aiming at maximizing the tritium breeding ratio (TBR) vs the material composition under the constraints of the thermal-hydraulic dimensioning. For a fixed value of the volume percentage of voids (25%, including helium coolant), and for upper and lower limits of ceramic breeder ( 10%), respectively, an optimal blanket configuration is obtained with a 1-D TBR of 1.40 (lithium silicate) and 1.30 (lithium alluminate), with a 75% Li6 enrichment of the breeder material. Results of the optimization code together with their design implications are discussed in the paper. A conceptual design of the optimal blanket configuration is developed, starting from a poloidal breeder-in-tube scheme based on the ongoing ENEA helium-cooled blanket design for DEMO. The main operating conditions and features of the DEMO-relevant ceramic breeder lead/graphite blanket design are max. neutron wall load = 2.6 MW/m2; helium coolant inlet/outlet temperature = 250/500°C; max. structure temperature = 550°C; cooling pumping power percentage = 4%; and helium purge circuit entirely separated from the coolant circuit.

Research paper thumbnail of Plasma Facing Component Activation in Net-II Initial D-T Pulsed Operations

Fusion Technology 1992, 1993

This paper presents the main results of the NET-II/ITER first wall stainless steel activation and... more This paper presents the main results of the NET-II/ITER first wall stainless steel activation and dose evaluation performed using different activation codes, nuclear cross section data, and data processing techniques. Basic results have been obtained by ANITA activation code. A benchmark exercise has been set up in order to compare nuclear cross section data libraries (AMPX, VITAMIN-C, and VITAMIN-J) and activation codes (ANITA, ORIGEN, and FISPACT-2). The update of the ORIGEN neutron Data Library was obtained by collapsing the 100-group GREAC-ECN-5 Activation Library with the flux-weighted spectrum provided by XSDRNPM-S code; this method allows the radioactivity inventory and the decay heat power evaluations to be done by the ORIGEN-S code. XSDOSE has been used in conjunction with a fixed-source XSDRNPM to evaluate dose rates outside the shield. The ORIGEN-S provides the gamma and neutron source strength and spectra. Various neutron power loads and fluences have been considered. The dose assessment methodology has been checked by comparison of calculated and experimental gamma-dose rate values following non-continuous irradiation of standard fuel elements (U, UO2 natural or slightly enriched) used over a period of 15 years in RB-3 research critical facility, at Montecuccolino, Bologna.

Research paper thumbnail of Thermal and hydraulic analysis of the cooling system for the ITER equatorial port plugs

Fusion Engineering and Design, 2009

Thermo-hydraulic analysis of the equatorial port plug (EPP) is currently being performed by CIEMA... more Thermo-hydraulic analysis of the equatorial port plug (EPP) is currently being performed by CIEMAT within the framework of the engineering activities related to the ITER port plug diagnostic integration launched by EFDA. This study is focussed on calculating the main hydraulic parameters for the reference cooling circuit, and analyzing different piping configurations, in order to achieve a balanced circuit, this being a key point for obtaining an homogeneous refrigeration of the port plug. It also analyzes the cooling requirements for the Diagnostic Shield Modules (DSM), considering not only the reference circuit design, where the piping cooling circuits run through those parts that need to be cooled, but also a new design proposal which entails a port plug cooling system based on the thermal contact between the DSM and the port plug itself. The overall study represents not only an improvement in the equatorial port plug cooling system design, but also a better understanding of the whole system so as to foresee the weak points that will need to be taken into account in the subsequent detailed design.

Research paper thumbnail of Neutronics experiment for the validation of activation properties of DEMO materials using real DT neutron spectrum at JET

Fusion Engineering and Design, 2006

ABSTRACT A key issue in the long-term programme of European Fusion Development Agreement (EFDA) i... more ABSTRACT A key issue in the long-term programme of European Fusion Development Agreement (EFDA) is the experimental validation of the neutron-induced activation predicted by calculations for structural materials developed for fusion application. Up to now, experiments have been carried out at neutron generators producing mono-energetic 14 MeV neutrons and using experimental assemblies simulating reactor-like neutron flux spectra. Since in October 2003 a Tritium Trace Experiment (TTE) was performed at JET, two samples of EUROFER-97 were exposed to the real DT neutron source produced in the tokamak. The goal was to measure the activity induced by neutrons in this material and to compare it to that calculated using the European Activation System EASY-2003 with FENDL-2/A neutron activation cross-section library. This was the first attempt to compare experimental data obtained in a real DT neutron spectrum with calculation results. The conclusion of this work is that the EASY code and the FENDL-2/A can evaluate the induced activity of medium half-life isotopes with an acceptable uncertainty, within the experimental total error. (c) 2005 Elsevier B.V. All rights reserved.

Research paper thumbnail of Experimental validation of shut down dose rates calculations inside ITER cryostat

Fusion Engineering and Design, 2001

A neutronics experiment has been performed at the 14 MeV Frascati Neutron Generator (FNG) to vali... more A neutronics experiment has been performed at the 14 MeV Frascati Neutron Generator (FNG) to validate the calculations of shut down dose rates inside the ITER cryostat. A proper experimental set-up, in which a neutron spectrum is generated similar to that occurring in the ITER vacuum vessel, has been irradiated for sufficiently long time to create a level of radioactivity

Research paper thumbnail of Validation of shutdown dose rate Monte Carlo calculations through a benchmark experiment at JET

Fusion Engineering and Design, 2008

A benchmark experiment has been performed during the 2005-2007 campaigns of JET in order to valid... more A benchmark experiment has been performed during the 2005-2007 campaigns of JET in order to validate the rigorous two-step (R2S) and direct one-step (D1S) methods for the prediction of shutdown dose rates in full 3D geometry. Dose rate levels calculated using D1S and R2S methods have been compared with experimental data collected before and during off-operational periods and at the end of the last JET campaigns. Measurements have been carried out in two irradiation positions: in-vessel with high sensitivity thermoluminescent detectors and ex-vessel with an active detector of Geiger-Mueller type. Satisfying agreement between calculations and measurements has been obtained in in-vessel position, whereas both methods underestimate the experimental quantity in the external one.

Research paper thumbnail of Status of JT-60SA tokamak under the EU-JA Broader Approach Agreement

Fusion Engineering and Design, 2008

JT-60SA is a fully superconducting coil tokamak to be built under the framework of the EU-JA Broa... more JT-60SA is a fully superconducting coil tokamak to be built under the framework of the EU-JA Broader Approach Agreement, and it aims to contribute to the complement of the ITER experiments and to the DEMO reactor design by the study of steady-state high-beta plasma experiments. The conceptual design of the JT-60SA tokamak and the peripheral systems has been carried out in close collaboration of EU and Japan from middle of 2006, and the JT-60SA Conceptual Design Report which involves an analysis of plasma physics and engineering design is summarized in May 2007. This paper intends to present an overall view of the JT-60SA project, and the latest design status of the key engineering issues such as superconducting magnets, divertor target, plasma heating devices, and magnet power supplies.

Research paper thumbnail of State-of-the-art 3-D radiation transport methods for fusion energy systems

Fusion Engineering and Design, 2008

Recent advances in radiation transport simulation tools enable an increased fidelity and accuracy... more Recent advances in radiation transport simulation tools enable an increased fidelity and accuracy in modeling complex geometries in fusion systems. Future neutronics calculations will increasingly be based directly on these 3-D CAD-based geometries, allowing enhanced model complexity and improved quality assurance. Improvements have been made in both stochastic and deterministic radiation transport methodologies and their new capabilities will be compared briefly. A code comparison benchmark exercise has been specified based on a 40 • sector of the ITER machine and the analysis results show good agreement. Additional analyses will be discussed, with particular attention to how these new capabilities provide new insights for engineering design of ITER components.

Research paper thumbnail of ITER Driver Blanket, European Community design

Fusion Engineering and Design, 1993

Depending on the final decision on the operation time of ITER (International Thermonuclear Experi... more Depending on the final decision on the operation time of ITER (International Thermonuclear Experimental Reactor), the Driver Blanket might become a basic component of the machine with the main function of producing a significant fraction (close to 0.8) of the tritium required for the ITER operation, the remaining fraction being available from external supplies. The Driver Blanket is not required to provide reactor relevant performance in terms of tritium self-sufficiency. However, reactor relevant reliability and safety are mandatory requirements for this component in order not to significantly affect the overall plant availability and to allow the ITER experimental program to be safely and successfully carried out. Within the framework of the ITER Conceptual Design Activities (CDA, 1988-1990), a conceptual design of the ITER Driver Blanket has been carried out by ENEA Fusion Dept., in collaboration with ANSALDO S.p.A. and SRS S.r.l., and in close consultation with the NET Team and CFFTP (Canadian Fusion Fuels Technology Project). Such a design has been selected as EC (European Community) reference design for the ITER Driver Blanket. The status of the design at the end of CDA is reported in the present paper.

Research paper thumbnail of Design of plasma facing components for the ITER feat divertor

Fusion Engineering and Design, 2002

A comprehensive design of the ITER divertor has been developed within the EU R&am... more A comprehensive design of the ITER divertor has been developed within the EU R&D for ITER. It consists of plasma facing components (PFCs) and cassettes body (CB). The PFCs are actively cooled thermal shields while the CB are massive supports for the PFCs providing also a neutronic shield. The present paper gives a detailed design of the PFCs and the

Research paper thumbnail of Neutronics and nuclear data issues in ITER and their validation

Fusion Engineering and Design, 2008

Neutron benchmark experiments carried out at 14 MeV neutron generators on ITER relevant materials... more Neutron benchmark experiments carried out at 14 MeV neutron generators on ITER relevant materials and components have provided validation of nuclear data used in shielding and activation calculations for ITER, as well as an assessment of the related uncertainties. Further validation activity is still needed which could be performed in ITER, especially for activation cross sections, for dose rate calculations and for tritium production rates. In particular, concerning the nuclear performance of breeder blankets to be tested in ITER, the paper discusses the extensive preparatory work, to be carried out in a coordinated way among the participating Parties, required to finalize the design of neutronics Test Blanket Modules (TBMs), so that these tests can provide a complete and comparable, therefore useful, picture of the different concepts. Neutronics experiments being carried out at neutron generators on TBM mock-ups are providing important results on the quality of relevant neutron cross sections and for the preparation of the measurement techniques.

Research paper thumbnail of Design optimisation and measuring techniques for the neutronics experiment on a HCLL–TBM mock-up

Fusion Engineering and Design, 2009

A neutronics experiment on a mock-up of the EU Test Blanket Module (TBM), helium cooled lithium l... more A neutronics experiment on a mock-up of the EU Test Blanket Module (TBM), helium cooled lithium lead concept, is in preparation with the objective to validate the capability of the neutronics codes and nuclear data to predict nuclear responses, such as the tritium production rate (TPR), with qualified uncertainties. Three independent measurements of the TPR will be performed using Li 2 CO 3 pellets. Other measurement techniques have been developed using thermoluminescence detectors, and diamond detectors covered with 6 LiF. Neutron flux spectra will also be measured from fast energies down to thermal energies, relevant for TPR. Comparison of measured quantities (E) with the same calculated quantities (C) will be provided, together with the related uncertainties. The paper presents the results of development of the measurement techniques and their relevance for tritium measurements in TBM in ITER. It presents also the pre-analyses conducted to optimise the mock-up configuration so that the neutron spectra are as similar as possible to those in the TBM in ITER. Sensitivity/uncertainty assessments of the TPR show that the calculation uncertainty due to the uncertainties of the neutron cross sections amounts to a few %, depending on position. The largest uncertainties are due to the elastic scattering (n,2n), and (n,3n) reactions on Pb.

Research paper thumbnail of Three-dimensional neutronic analysis of the European ceramic BIT DEMO blanket

Fusion Engineering and Design, 1991

The present study concerns the 3-D geometry evaluations of the tritium breeding ratio of the bree... more The present study concerns the 3-D geometry evaluations of the tritium breeding ratio of the breeder-in-tube helium-cooled ceramic blanket performed at CEA and ENEA in the framework of the recently released DEMONET specifications. For the calculations the nuclear data have all been derived by the European Fusion File (EFF-1), but different Monte Carlo codes, and different design variants and geometrical models have been used. A benchmark calculation has been defined in order to better analyse the results; such an exercise has shown that TBR differences of a few percent could be due to the calculation methods. The determined global TBR ranges from 1.06 in the CEA results to 1.15 in the ENEA results, the expected difference being mainly due to the greater neutron coverage of the ENEA variant of the design. Heat deposition density and He production in beryllium are also given.

Research paper thumbnail of SEAFP, Safety and Environmental Assessment of Fusion Power, Final Report

The Safety and Environmental Assessment of Fusion Power (SEAFP) was undertaken for the Commission... more The Safety and Environmental Assessment of Fusion Power (SEAFP) was undertaken for the Commission of the European Union in the framework of the Fusion Programme 1990-94. Its terms of reference were in accordance with the programme decision of the Council of Ministers which followed a request by the European Parliament and a recommendation of the Fusion Programme Evaluation Board. The SEAFP is part of an ongoing effort to consider the safety and environmental aspects of fusion power. SEAFP was carried out by the NET Team, the Euratom/UKAEA Association, and by Industry, with contributions from other Associated Laboratories, the Joint Research Centre and the Canadian fusion programme.

Research paper thumbnail of Activation analysis and characteristics of the European community water cooled ceramic breeder blanket design proposal for ITER

The European Community (EC) Home Team has proposed various alternative blanket designs to the bas... more The European Community (EC) Home Team has proposed various alternative blanket designs to the basic concept (essentially integrated first wall, cooled by liquid metal, with structures made by vanadium alloys). One of the EC proposal is the Water Cooled Ceramic Blanket developed on the basis of a common action between NET and ENEA. It is based on a more conservative approach, but involving well proven technologies and qualified materials: SS-316L as structural material, Li{sub 2}ZrO{sub 3} as first breeder material choice (50% Li{sup 6} enrichment) and low temperature water coolant (160/200{degrees}C). Beryllium has been chosen as multiplying material. The nominal performance are: 1 MW/m{sup 2} as average neutron wall load, corresponding to 1.5 GW fusion power, 1 MW-y/m{sup 2} beneath it has been proved to withstand power excursion till 5 GW. The proposed blanket concept is based on a Breeder Inside Tube (BIT) type technology, with poloidal breeding elements, each one consisting of t...

Research paper thumbnail of EFD-P(13)38 Response of LaBr3(Ce) Scintillators to 2.5MeV Fusion Neutrons

Measurements of the response function of LaBr3(Ce) to 2.5MeV neutrons have been carried out at th... more Measurements of the response function of LaBr3(Ce) to 2.5MeV neutrons have been carried out at the Frascati Neutron Generator and at tokamak facilities with deuterium plasmas. The observed spectrum has been interpreted by means of a MCNP model. It is found that the main contributor to the measured response is neutron inelastic scattering on 79Br, 81Br and 139La. An extrapolation of the count rate response to 14MeV neutrons from deuterium-tritium plasmas is also presented. The results are of relevance for the design of gamma-ray diagnostics of fusion burning plasmas.

Research paper thumbnail of Three-Dimensional Neutronics and Shielding Analyses for the ITER Divertor

Fusion Technology, 1996

3-D neutronics and shielding analyses have been performed for the divertor region of the ITER int... more 3-D neutronics and shielding analyses have been performed for the divertor region of the ITER interim design. The peak neutron wall loading in the divertor region is 0.6 MW/m² at the divertor cassette dome. The total nuclear heating in the 60 divertor cassettes is 102.4 MW. The peak helium production in the VV behind the pumping ducts is 0.5 He

Research paper thumbnail of European Reference Design of the Water-Cooled Lithium-Lead Blanket for a Demonstration Reactor

Fusion Technology, 1994

The water-cooled Pb-17Li blankets represent one of the blanket lines selected within the European... more The water-cooled Pb-17Li blankets represent one of the blanket lines selected within the European Union for DEMO-relevant design and R&D activities. This paper gives a presentation of the reference conceptual design for water-cooled Pb-17Li DEMO blankets and an overview on the results of its performance assessments. Moreover, a critical discussion about the technical aspects requiring further improvements and/or modifications is performed taking into account the present status of the associated R&D. This concept appears to be a very promising candidate for a DEMO reactor breeding blanket.

Research paper thumbnail of The ITER Radial Neutron Camera Detection System

AIP Conference Proceedings, 2008

Research paper thumbnail of Nuclear analysis in support of ITER design

Proceedings of 16th International Symposium on Fusion Engineering

This paper summarizes the main results of nuclear analysis calculations performed during the Inte... more This paper summarizes the main results of nuclear analysis calculations performed during the International Thermonuclear Experimental Reactor (ITER) Engineering Design Activity (EDA). Major efforts were devoted to fulfilling the General Design Requirements to minimize the nuclear heating rate in the superconducting magnets and ensuring that radiation conditions at the cryostat are suitable for hands-on-maintenance after reactor shutdown .

Research paper thumbnail of A Helium-Cooled Ceramic Breeder Blanket Concept with Lead/Graphite as Multiplier/Moderator

Fusion Technology 1990, 1991

Helium-cooled ceramic breeder blanket designs are commonly based on the use of beryllium as both ... more Helium-cooled ceramic breeder blanket designs are commonly based on the use of beryllium as both multiplying and moderating material. The possibility of using lead as multiplier and graphite as moderator, instead of beryllium, is investigated by means of a proper optimization code, aiming at maximizing the tritium breeding ratio (TBR) vs the material composition under the constraints of the thermal-hydraulic dimensioning. For a fixed value of the volume percentage of voids (25%, including helium coolant), and for upper and lower limits of ceramic breeder ( 10%), respectively, an optimal blanket configuration is obtained with a 1-D TBR of 1.40 (lithium silicate) and 1.30 (lithium alluminate), with a 75% Li6 enrichment of the breeder material. Results of the optimization code together with their design implications are discussed in the paper. A conceptual design of the optimal blanket configuration is developed, starting from a poloidal breeder-in-tube scheme based on the ongoing ENEA helium-cooled blanket design for DEMO. The main operating conditions and features of the DEMO-relevant ceramic breeder lead/graphite blanket design are max. neutron wall load = 2.6 MW/m2; helium coolant inlet/outlet temperature = 250/500°C; max. structure temperature = 550°C; cooling pumping power percentage = 4%; and helium purge circuit entirely separated from the coolant circuit.

Research paper thumbnail of Plasma Facing Component Activation in Net-II Initial D-T Pulsed Operations

Fusion Technology 1992, 1993

This paper presents the main results of the NET-II/ITER first wall stainless steel activation and... more This paper presents the main results of the NET-II/ITER first wall stainless steel activation and dose evaluation performed using different activation codes, nuclear cross section data, and data processing techniques. Basic results have been obtained by ANITA activation code. A benchmark exercise has been set up in order to compare nuclear cross section data libraries (AMPX, VITAMIN-C, and VITAMIN-J) and activation codes (ANITA, ORIGEN, and FISPACT-2). The update of the ORIGEN neutron Data Library was obtained by collapsing the 100-group GREAC-ECN-5 Activation Library with the flux-weighted spectrum provided by XSDRNPM-S code; this method allows the radioactivity inventory and the decay heat power evaluations to be done by the ORIGEN-S code. XSDOSE has been used in conjunction with a fixed-source XSDRNPM to evaluate dose rates outside the shield. The ORIGEN-S provides the gamma and neutron source strength and spectra. Various neutron power loads and fluences have been considered. The dose assessment methodology has been checked by comparison of calculated and experimental gamma-dose rate values following non-continuous irradiation of standard fuel elements (U, UO2 natural or slightly enriched) used over a period of 15 years in RB-3 research critical facility, at Montecuccolino, Bologna.

Research paper thumbnail of Thermal and hydraulic analysis of the cooling system for the ITER equatorial port plugs

Fusion Engineering and Design, 2009

Thermo-hydraulic analysis of the equatorial port plug (EPP) is currently being performed by CIEMA... more Thermo-hydraulic analysis of the equatorial port plug (EPP) is currently being performed by CIEMAT within the framework of the engineering activities related to the ITER port plug diagnostic integration launched by EFDA. This study is focussed on calculating the main hydraulic parameters for the reference cooling circuit, and analyzing different piping configurations, in order to achieve a balanced circuit, this being a key point for obtaining an homogeneous refrigeration of the port plug. It also analyzes the cooling requirements for the Diagnostic Shield Modules (DSM), considering not only the reference circuit design, where the piping cooling circuits run through those parts that need to be cooled, but also a new design proposal which entails a port plug cooling system based on the thermal contact between the DSM and the port plug itself. The overall study represents not only an improvement in the equatorial port plug cooling system design, but also a better understanding of the whole system so as to foresee the weak points that will need to be taken into account in the subsequent detailed design.

Research paper thumbnail of Neutronics experiment for the validation of activation properties of DEMO materials using real DT neutron spectrum at JET

Fusion Engineering and Design, 2006

ABSTRACT A key issue in the long-term programme of European Fusion Development Agreement (EFDA) i... more ABSTRACT A key issue in the long-term programme of European Fusion Development Agreement (EFDA) is the experimental validation of the neutron-induced activation predicted by calculations for structural materials developed for fusion application. Up to now, experiments have been carried out at neutron generators producing mono-energetic 14 MeV neutrons and using experimental assemblies simulating reactor-like neutron flux spectra. Since in October 2003 a Tritium Trace Experiment (TTE) was performed at JET, two samples of EUROFER-97 were exposed to the real DT neutron source produced in the tokamak. The goal was to measure the activity induced by neutrons in this material and to compare it to that calculated using the European Activation System EASY-2003 with FENDL-2/A neutron activation cross-section library. This was the first attempt to compare experimental data obtained in a real DT neutron spectrum with calculation results. The conclusion of this work is that the EASY code and the FENDL-2/A can evaluate the induced activity of medium half-life isotopes with an acceptable uncertainty, within the experimental total error. (c) 2005 Elsevier B.V. All rights reserved.

Research paper thumbnail of Experimental validation of shut down dose rates calculations inside ITER cryostat

Fusion Engineering and Design, 2001

A neutronics experiment has been performed at the 14 MeV Frascati Neutron Generator (FNG) to vali... more A neutronics experiment has been performed at the 14 MeV Frascati Neutron Generator (FNG) to validate the calculations of shut down dose rates inside the ITER cryostat. A proper experimental set-up, in which a neutron spectrum is generated similar to that occurring in the ITER vacuum vessel, has been irradiated for sufficiently long time to create a level of radioactivity

Research paper thumbnail of Validation of shutdown dose rate Monte Carlo calculations through a benchmark experiment at JET

Fusion Engineering and Design, 2008

A benchmark experiment has been performed during the 2005-2007 campaigns of JET in order to valid... more A benchmark experiment has been performed during the 2005-2007 campaigns of JET in order to validate the rigorous two-step (R2S) and direct one-step (D1S) methods for the prediction of shutdown dose rates in full 3D geometry. Dose rate levels calculated using D1S and R2S methods have been compared with experimental data collected before and during off-operational periods and at the end of the last JET campaigns. Measurements have been carried out in two irradiation positions: in-vessel with high sensitivity thermoluminescent detectors and ex-vessel with an active detector of Geiger-Mueller type. Satisfying agreement between calculations and measurements has been obtained in in-vessel position, whereas both methods underestimate the experimental quantity in the external one.

Research paper thumbnail of Status of JT-60SA tokamak under the EU-JA Broader Approach Agreement

Fusion Engineering and Design, 2008

JT-60SA is a fully superconducting coil tokamak to be built under the framework of the EU-JA Broa... more JT-60SA is a fully superconducting coil tokamak to be built under the framework of the EU-JA Broader Approach Agreement, and it aims to contribute to the complement of the ITER experiments and to the DEMO reactor design by the study of steady-state high-beta plasma experiments. The conceptual design of the JT-60SA tokamak and the peripheral systems has been carried out in close collaboration of EU and Japan from middle of 2006, and the JT-60SA Conceptual Design Report which involves an analysis of plasma physics and engineering design is summarized in May 2007. This paper intends to present an overall view of the JT-60SA project, and the latest design status of the key engineering issues such as superconducting magnets, divertor target, plasma heating devices, and magnet power supplies.

Research paper thumbnail of State-of-the-art 3-D radiation transport methods for fusion energy systems

Fusion Engineering and Design, 2008

Recent advances in radiation transport simulation tools enable an increased fidelity and accuracy... more Recent advances in radiation transport simulation tools enable an increased fidelity and accuracy in modeling complex geometries in fusion systems. Future neutronics calculations will increasingly be based directly on these 3-D CAD-based geometries, allowing enhanced model complexity and improved quality assurance. Improvements have been made in both stochastic and deterministic radiation transport methodologies and their new capabilities will be compared briefly. A code comparison benchmark exercise has been specified based on a 40 • sector of the ITER machine and the analysis results show good agreement. Additional analyses will be discussed, with particular attention to how these new capabilities provide new insights for engineering design of ITER components.

Research paper thumbnail of ITER Driver Blanket, European Community design

Fusion Engineering and Design, 1993

Depending on the final decision on the operation time of ITER (International Thermonuclear Experi... more Depending on the final decision on the operation time of ITER (International Thermonuclear Experimental Reactor), the Driver Blanket might become a basic component of the machine with the main function of producing a significant fraction (close to 0.8) of the tritium required for the ITER operation, the remaining fraction being available from external supplies. The Driver Blanket is not required to provide reactor relevant performance in terms of tritium self-sufficiency. However, reactor relevant reliability and safety are mandatory requirements for this component in order not to significantly affect the overall plant availability and to allow the ITER experimental program to be safely and successfully carried out. Within the framework of the ITER Conceptual Design Activities (CDA, 1988-1990), a conceptual design of the ITER Driver Blanket has been carried out by ENEA Fusion Dept., in collaboration with ANSALDO S.p.A. and SRS S.r.l., and in close consultation with the NET Team and CFFTP (Canadian Fusion Fuels Technology Project). Such a design has been selected as EC (European Community) reference design for the ITER Driver Blanket. The status of the design at the end of CDA is reported in the present paper.

Research paper thumbnail of Design of plasma facing components for the ITER feat divertor

Fusion Engineering and Design, 2002

A comprehensive design of the ITER divertor has been developed within the EU R&am... more A comprehensive design of the ITER divertor has been developed within the EU R&D for ITER. It consists of plasma facing components (PFCs) and cassettes body (CB). The PFCs are actively cooled thermal shields while the CB are massive supports for the PFCs providing also a neutronic shield. The present paper gives a detailed design of the PFCs and the

Research paper thumbnail of Neutronics and nuclear data issues in ITER and their validation

Fusion Engineering and Design, 2008

Neutron benchmark experiments carried out at 14 MeV neutron generators on ITER relevant materials... more Neutron benchmark experiments carried out at 14 MeV neutron generators on ITER relevant materials and components have provided validation of nuclear data used in shielding and activation calculations for ITER, as well as an assessment of the related uncertainties. Further validation activity is still needed which could be performed in ITER, especially for activation cross sections, for dose rate calculations and for tritium production rates. In particular, concerning the nuclear performance of breeder blankets to be tested in ITER, the paper discusses the extensive preparatory work, to be carried out in a coordinated way among the participating Parties, required to finalize the design of neutronics Test Blanket Modules (TBMs), so that these tests can provide a complete and comparable, therefore useful, picture of the different concepts. Neutronics experiments being carried out at neutron generators on TBM mock-ups are providing important results on the quality of relevant neutron cross sections and for the preparation of the measurement techniques.

Research paper thumbnail of Design optimisation and measuring techniques for the neutronics experiment on a HCLL–TBM mock-up

Fusion Engineering and Design, 2009

A neutronics experiment on a mock-up of the EU Test Blanket Module (TBM), helium cooled lithium l... more A neutronics experiment on a mock-up of the EU Test Blanket Module (TBM), helium cooled lithium lead concept, is in preparation with the objective to validate the capability of the neutronics codes and nuclear data to predict nuclear responses, such as the tritium production rate (TPR), with qualified uncertainties. Three independent measurements of the TPR will be performed using Li 2 CO 3 pellets. Other measurement techniques have been developed using thermoluminescence detectors, and diamond detectors covered with 6 LiF. Neutron flux spectra will also be measured from fast energies down to thermal energies, relevant for TPR. Comparison of measured quantities (E) with the same calculated quantities (C) will be provided, together with the related uncertainties. The paper presents the results of development of the measurement techniques and their relevance for tritium measurements in TBM in ITER. It presents also the pre-analyses conducted to optimise the mock-up configuration so that the neutron spectra are as similar as possible to those in the TBM in ITER. Sensitivity/uncertainty assessments of the TPR show that the calculation uncertainty due to the uncertainties of the neutron cross sections amounts to a few %, depending on position. The largest uncertainties are due to the elastic scattering (n,2n), and (n,3n) reactions on Pb.

Research paper thumbnail of Three-dimensional neutronic analysis of the European ceramic BIT DEMO blanket

Fusion Engineering and Design, 1991

The present study concerns the 3-D geometry evaluations of the tritium breeding ratio of the bree... more The present study concerns the 3-D geometry evaluations of the tritium breeding ratio of the breeder-in-tube helium-cooled ceramic blanket performed at CEA and ENEA in the framework of the recently released DEMONET specifications. For the calculations the nuclear data have all been derived by the European Fusion File (EFF-1), but different Monte Carlo codes, and different design variants and geometrical models have been used. A benchmark calculation has been defined in order to better analyse the results; such an exercise has shown that TBR differences of a few percent could be due to the calculation methods. The determined global TBR ranges from 1.06 in the CEA results to 1.15 in the ENEA results, the expected difference being mainly due to the greater neutron coverage of the ENEA variant of the design. Heat deposition density and He production in beryllium are also given.