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... 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.
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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... 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...
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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... 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.
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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... 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
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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... 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.
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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... 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.
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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... 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.
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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... 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.
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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... 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
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Research Interests: Mechanical Engineering, Computer Science, Monte Carlo, Engineering Design, Quality Assurance, and 10 moreFusion Engineering, Energy System, Three Dimensional, Interdisciplinary Engineering, Radiation Transport, Model Complexity, Simulation Tool, Fusion Power, neutron transport, and Discrete Ordinates
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A comprehensive design of the ITER divertor has been developed within the EU R&D for ITER. It consists of plasma facing... 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
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A blanket option featured by helium cooling and lithiated ceramics inside tubes is under development at CEA/ENEA within the framework of the European Test Blanket Program. A first conceptual design...
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Abstract The ITER main vacuum system consists of six torus exhaust pumps integrated in the Cryostat through dedicated housings at the building B1 level. The Port Cell area outside the cryopump plug is affected by neutrons streaming... more
Abstract The ITER main vacuum system consists of six torus exhaust pumps integrated in the Cryostat through dedicated housings at the building B1 level. The Port Cell area outside the cryopump plug is affected by neutrons streaming through the housing structure and diagnostics penetrations. The aim of the study presented in this paper is to perform a complete assessment of the nuclear responses in the Cryopump Port Cell #12 by means of the MCNP-5 Monte Carlo code in a full 3-D geometry. The results of the neutronic analyses provide guidelines for the design and maintenance of the embedded components, ensuring their structural integrity and proper operation. Radiation transport calculations have been carried out to determine the radiation field inside the Port Cell through 3-D neutrons and gamma maps. Nuclear heating induced by neutrons and photons and the absorbed dose during the ITER lifetime on steel and silicon have been estimated in the Port Cell area, in order to assess the nuclear loads that the diagnostics and related electronic components have to withstand. Furthermore, the impact of the gamma-rays emitted by neutron-activated water circulating in the Primary Heat Transfer System have been evaluated on the Port Cell environment: 3-D maps of the gamma flux, nuclear loads on steel/silicon during plasma operation are provided.
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Abstract The ITER machine will be equipped with 6 torus Cryopumps (TCP) that are positioned in their housings (TCPH) and integrated into the cryostat walls at B1 level in the port cells. A comprehensive nuclear analysis of the Cryopump... more
Abstract The ITER machine will be equipped with 6 torus Cryopumps (TCP) that are positioned in their housings (TCPH) and integrated into the cryostat walls at B1 level in the port cells. A comprehensive nuclear analysis of the Cryopump Ports #4 and #12 has been carried out by means of the MCNP-5 Monte Carlo code in a full 3-D geometry, providing guidelines for the design of the embedded components. Radiation transport calculations have been performed in order to determine the radiation field inside the Lower Ports, up the Port Cell: 3-D neutrons and gamma maps have been provided in order to evaluate the shielding effectiveness of the TCPHs. Nuclear heating induced by neutron and photons have been estimated on the TCP and TCPH to assess the nuclear loads during plasma operations. The shutdown dose rate in the maintenance area of the Lower Ports has been assessed with the Advanced D1S method to verify the design limits.
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In the ITER (International Thermonuclear Experimental Reactor, [1]) project calculations of the dose rate after shutdown are very important and their results are critical for the machine design. A new method has been proposed [2] which... more
In the ITER (International Thermonuclear Experimental Reactor, [1]) project calculations of the dose rate after shutdown are very important and their results are critical for the machine design. A new method has been proposed [2] which makes use of MCNP [3] also for decay gamma-ray transport calculations. The objective is to have an easy tool giving results affected by low uncertainty due to the modeling or simplifications in the flux shape assumptions. Further improvements to this method are here presented.
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Abstract In the frame of ITER (International Tokamak Experimental Reactor) design program, 3D neutronics calculations have been carried out to assess the system shielding performances in the basic machine configuration by means of Monte... more
Abstract In the frame of ITER (International Tokamak Experimental Reactor) design program, 3D neutronics calculations have been carried out to assess the system shielding performances in the basic machine configuration by means of Monte Carlo Neutron Photon (MCNP) code (3-B version). The main issue concerns the estimation of the nuclear heat and radiation loads on the toroidal field superconducting coils. “Self generated weight windows” (w.w.) and source biasing technique have been used to treat the deep penetration through the bulk shield and the streaming through the system gaps and openings. The main results are reported together with a discussion of the computing methods, especially of the variance reduction techniques adopted.
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ABSTRACT Nuclear analyses have been performed for the ITER Test Blanket Module Port Plug (TBM PP) using the MCNP-5 Monte Carlo Code. A detailed 3D model of the TBM Port Plug with dummy TBM has been integrated into the ITER MCNP model... more
ABSTRACT Nuclear analyses have been performed for the ITER Test Blanket Module Port Plug (TBM PP) using the MCNP-5 Monte Carlo Code. A detailed 3D model of the TBM Port Plug with dummy TBM has been integrated into the ITER MCNP model (B-lite v.3). Neutron fluxes, nuclear heating, helium production and neutron damage have been calculated in all the TBM PP components. Global shutdown dose rate calculations have also been performed with Advanced D1S method for different configurations of the TBM PP system. This paper presents the results of these analyses and discusses potential design improvements aiming to further reduce the shutdown dose rate in the port interspace.
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ABSTRACT The Neutral Beam (NB) devices for the International Thermonuclear Experimental Reactor (ITER) are medium energy light particle accelerators. Accelerated particles are negative deuterium ions with maximum energy of 1 MeV. The... more
ABSTRACT The Neutral Beam (NB) devices for the International Thermonuclear Experimental Reactor (ITER) are medium energy light particle accelerators. Accelerated particles are negative deuterium ions with maximum energy of 1 MeV. The deuteron beam with a pulse current of 40 A and a duration time of 20 s is stopped on a water-copper alloy calorimeter. The main features of the NBs will be tested in a Test Facility (NBTF) that will be built in a different site from that of ITER. In the work presented here the assessment of the radiological safety for the maintenance of the calorimeter of the NBTF is described and the related results are discussed. The activation analysis of NBTF components is performed using MCNP Monte Carlo Code and FISPACT inventory Code in order to predict shutdown dose rates. Starting from a refined description of the D-D neutron source, the induced activation of the NBTF components is calculated, for two realistic operational scenarios. Normal maintenance activities for the calorimeter are identified and the related operator positions are defined. Finally individual and collective doses are assessed for the whole maintenance operation.
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In the frame of the studies developed for DEMO breeding blankets, a type of blanket with Pb17Li as breeder and water as coolant has been developed at the JRC-Ispra. It deals with breeder circular modules, internally cooled by water in... more
In the frame of the studies developed for DEMO breeding blankets, a type of blanket with Pb17Li as breeder and water as coolant has been developed at the JRC-Ispra. It deals with breeder circular modules, internally cooled by water in U-type tubes and arranged poloidally according to rows inside boxes, named segments. A 3-D neutronic analysis of the DEMO reactor with this blanket configuration has been performed, to verify if tritium breeding self-sufficiency could be achieved without the need of additional multipliers. The DEMONET specifications have been applied, and the MCNP-3B Monte-Carlo code has been used. A careful modelling of the complicated blanket layout has been carried out, avoiding where possible homogenization procedures. A total Tritium Breeding Ratio (TBR) of 1.152 is obtained as a final result of the MCNP run. If the presence of the ports is considered, a net TBR of 1.09 is achieved. It is therefore verified that this blanket permits to achieve self-sufficiency without any additional multiplier. The results of this analysis are critically compared with previous ones, which used 1-D models and approximated expressions to estimate the net TBR. A good agreement between the two studies is found out.
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AbstractAs part of the Fusion Advanced Studies Torus (FAST) project, a neutronic analysis has been performed, aimed to design optimization and radiological safety assessment. The neutron emissivity source foreseen for various FAST... more
AbstractAs part of the Fusion Advanced Studies Torus (FAST) project, a neutronic analysis has been performed, aimed to design optimization and radiological safety assessment. The neutron emissivity source foreseen for various FAST scenarios has been calculated and used as ...
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Silicon carbide (SiC) in the form of ceramic matrix is a low activation structural material proposed for fusion reactors. Its development is pursued in the European Fusion Technology Program. A SiC block (457×457×711 mm3), borrowed from... more
Silicon carbide (SiC) in the form of ceramic matrix is a low activation structural material proposed for fusion reactors. Its development is pursued in the European Fusion Technology Program. A SiC block (457×457×711 mm3), borrowed from JAERI, was irradiated with 14 MeV neutrons at the FNG facility of ENEA Frascati. Activation reaction rates, neutron fluxes and spectra, as well as
Research Interests: Mechanical Engineering, Nuclear Engineering, Materials Science, Nuclear Physics, Monte Carlo, and 11 moreSilicon Carbide, Uncertainty analysis, Cross Section, Interdisciplinary Engineering, Neutron Flux, Nuclear Data, Neutron, Reaction Rate, Experimental Analysis, neutron transport, and Discrete Ordinates
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ABSTRACT The reliable assessment of dose rates is one of the key issues in operating nuclear machines. For fusion technology applications, two different computational methods have been recently developed to this end, the so-called R2S and... more
ABSTRACT The reliable assessment of dose rates is one of the key issues in operating nuclear machines. For fusion technology applications, two different computational methods have been recently developed to this end, the so-called R2S and the D1S approach. First experimental and computational benchmarks showed good results and promising prospective in application to ITER. The application to the biggest European fusion machine JET, showed some deficiencies, especially in the outer regions of the machine, where doses are lower, but predictability importance higher. This has been attributed to the very complex geometry and the poor accuracy of both the measurements and the model used in the calculation for the outer regions. A new benchmark experiment has been proposed for the next campaign of JET. Preliminary calculation results of the dose rates achievable in such experiments are reported. Those are compatible with the sensibility of the detectors that will be used even if the residual background gives a consistent contribution. In the outer part of the machine the critical issue will be the ability to describe in detail the components close to the detector in the MCNP model.