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The validation of the key technologies relevant for a DEMO Breeding Blanket is one of the main objectives of the design and operation of the Test Blanket Systems (TBS) in ITER. In compliance with the main features and technical requirements of the parent breeding blanket concepts, the European TBM Project is developing the HCLL (Helium Cooled Lithium Lead) and HCPB (Helium Cooled Pebble Bed)-TBS, focusing in this phase on the design life cycle and on R&D activities in support of the design. The TBS ancillary systems are mainly circuits devoted to the removal of thermal power and to the extraction and recovery of the tritium generated in the Test Blanket Modules. They are: • The Helium Cooling System (HCS); • The Coolant Purification System (CPS); • The Tritium Extraction System (HCLL-TRS, HCPB-TES); • The Lead Lithium Loop. Their conceptual design was deeply analyzed during the ITER Conceptual Design Review (CDR) in 2015. The assessment of the CDR was overall positive and recommendations for improvements were made. The present design takes into account the main CDR recommendations as well as the implementation of the requirements related to ITER operation, safety principles application and physical space constrains.

In the framework of the EU fusion roadmap implementing activities, an accelerator-based Li(d,xn) neutron source called DONES (Demo-Oriented early NEutron Source) is being designed within the EUROfusion Work Package Early Neutron Source as an essential irradiation facility for testing candidate materials for DEMO reactor and future fusion power plants. DONES will employ a high speed liquid lithium jet struck by a 125 mA, 40 MeV deuteron beam to generate the intense neutron flux used to irradiate the material samples up to the desired level of displacement damage (˜10 dpa/fpy for iron in 0.3 l) and He production rates (˜10-13 appm He/dpa). In order to rapidly achieve a sound and stable design, a new configuration of the DONES target system based on the so-called integral concept has been proposed as reference solution in place of the former baseline design that envisaged a target assembly endowed with a replaceable back-plate, being the latter solution not yet fully qualified and thus not readily implementable. Moreover, following the outcomes of detailed dedicated analyses taking into account different aspects, the design of several components (Quench Tank, Lithium inlet pipe, Vacuum Chamber, etc.) has been changed. In this paper, a brief description of the current design of the DONES Target System is presented including all the above-mentioned aspects, showing the capability of the system to fulfil the prescribed requirements.

In the framework of the current IFMIF Engineering Validation and Engineering Design Activities (IFMIF/EVEDA) phase, ENEA is responsible for the design of the European concept of the IFMIF lithium target system which foresees the possibility to periodically replace only the most irradiated and thus critical component (i.e., the backplate) while continuing to operate the rest of the target for a longer period (bayonet backplate concept). With the objective of evaluating the performances of the system in terms of temperature, stress and displacement fields evolution during start-up and shutdown phases, an uncoupled thermomechanical transient analysis has been performed in close collaboration with the University of Palermo by means of a qualified finite element (FE) thermomechanical code. The calculations employed a realistic 3D time-dependent FE model which takes into account all the mechanical and thermal loads including the nuclear heating due to neutron and prompt gamma fields during start-up and decay power of activated products during shutdown. The nuclear data have been calculated by ENEA as part of a parallel extensive neutronic analysis carried out through the MCNP transport code and the EASY-2010 activation code package and then passed as input to the thermomechanical FE model. In this paper, the results of the above transient analyses are reported, highlighting the relevant indications obtained with respect to the fulfillment of the design requirements and possible hints for improving the system design. In particular, it is found that ITER design rules taking into account secondary loads are not always fulfilled during the transient, whereas the predicted displacements allow to exclude any contact of the target system with the surrounding structures. © 2014 Elsevier B.V.

ENEA is developing an accelerator-driven 14 MeV neutron source exploiting the deuterium–tritium fusion reaction to produce 99Mo medical radioisotope as an alternative production route not based on fission reactors. It is expected that, during normal operation, a number of radionuclides, generated by means of neutron irradiation on the raw material (natural Molybdenum), will be produced and managed. The present manuscript, as foreseen by national law, discusses a hypothetical scenario to test the environmental screening models, in turn evaluating the mechanisms and parameters which affect and control the path of liquid effluents potentially released during normal operation of the facility. The aim is to estimate the amount of radioactivity to be operated and the fraction potentially discharged in this hypothetical scenario, so as to ensure that the radioactive material can be managed without any risk for the population and the environment, according to national regulations and thoroughly fulfilling the international guidelines.

In the framework of the Work Package Early Neutron Source (WPENS) of the EUROfusion Consortium, the engineering design of the IFMIF-DONES (International Fusion Material Irradiation Facility - Demo Oriented Neutron Source) lithium target system has been accomplished. The development of the design required that new neutronic calculations had to be performed in ENEA, in order to update those already achieved in the past for the previous geometrical layout. The aim of these evaluations is to provide quantities, such as power deposition, dpa and gas production, useful for the thermo-mechanical analysis that is required for assessing the structural behavior of the system subjected to irradiation. Coupled neutron-gamma transport calculations have been carried out by using the MCNP6.2 code integrated with the McDeLicious-17 neutron source. The geometrical input for MCNP used in the calculations was provided by KIT. Neutron activation calculations have been also performed by means of the FISPACT-II activation code package in order to provide radioactive quantities, such as decay heat and contact dose rates, in various parts of the system, useful for thermo-mechanical analysis and safety purposes, like waste management considerations. This paper presents the main results obtained from the above analyses.

The methodological approach applied for the nuclear analyses of the IFMIF-DONES facility, developed within the Early Neutron Source (ENS) project of EUROfusion, is presented. It utilizes dedicated tools, developed and adapted to the specific needs and peculiarities of an accelerator based d-Li neutron source. Major elements relate to the simulation of deuteron interactions with the Li target and the accelerator structures, the generation and transport of neutrons and photons, and the production of radio-active nuclides with the subsequent emission and transport of decay gamma radiation. With such an approach, all neutronics data can be provided which are required to design, optimize and evaluate the systems/sub-systems of the facility with regard to their nuclear and shielding performance. This is shown on the example of nuclear analyses conducted for the latest design of IFMIF-DONES.

ANITA-IEAF is an activation package (code and libraries) developed in the past in ENEA-Bologna in order to assess the activation of materials exposed to neutrons with energies greater than 20 MeV. An updated version of the ANITA-IEAF activation code package has been developed. It is suitable to be applied to the study of the irradiation effects on materials in facilities like the International Fusion Materials Irradiation Facility (IFMIF) and the DEMO Oriented Neutron Source (DONES), in which a considerable amount of neutrons with energies above 20 MeV is produced. The present paper summarizes the main characteristics of the updated version of ANITA-IEAF, able to use decay and cross section data based on more recent evaluated nuclear data libraries, i.e. the JEFF-3.1.1 Radioactive Decay Data Library and the EAF-2010 neutron activation cross section library. In this paper the validation effort related to the comparison between the code predictions and the activity measurements obtained from the Karlsruhe Isochronous Cyclotron is presented. In this integral experiment samples of two different steels, SS-316 and F82H, pure vanadium and a vanadium alloy, structural materials of interest in fusion technology, were activated in a neutron spectrum similar to the IFMIF neutron field. © The Authors, published by EDP Sciences, 2017.

The availability of a high flux neutron source for testing candidate materials under irradiation conditions, which will be typically encountered in future fusion power reactors (ITER, DEMO, FPR), is a fundamental step towards the development of fusion energy. To this purpose, the International Fusion Materials Irradiation Facility (IFMIF) represents the reference option to provide the fusion community with a DEMO-relevant neutron source capable of irradiating samples at a damage rate of up to 20 dpa/fpy (in steel) in a volume of 0.5 l. In the framework of the engineering design activities of IFMIF, ENEA is committed in the design of the lithium target assembly (TA) with removable (bayonet) backplate (BP) whose development has recently progressed up to a well advanced stage. However, an optimization of the system is still to be accomplished. In particular, the BP design needs to be revised in order to totally satisfy the ITER SDC-IC reference design criteria and fulfil the requirements on its lifetime which is limited by the neutron-induced swelling effects. In this work, a full thermo-mechanical analysis of the whole TA including a pseudo-transient simulation of the swelling effects in the BP over one year of full power operation has been performed by means of a 3D finite element (FE) model implemented through the quoted ABAQUS v. 6.14 code. A detailed neutronic analysis has been also performed by ENEA using the MCNP code to obtain the prompt nuclear responses, in terms of DPA and volumetric density of deposited nuclear heat power, to be used as input for the thermo-mechanical calculations. A new BP design capable to verify the SDC-IC design criteria and ensure its required swelling lifetime is proposed and described on the basis of the results of the performed analysis.

The ENEA-Bologna Nuclear Data Group produced the VITJEFF32.BOLIB multi-group coupled neutron/photon (199 n + 42 γ) cross section library in AMPX format, based on the OECD-NEA Data Bank JEFF-3.2 evaluated nuclear data library. VITJEFF32.BOLIB was conceived for nuclear fission applications as European counterpart of the ORNL VITAMIN-B7 similar library (ENDF/B-VII.0 data). VITJEFF32.BOLIB has the same neutron and photon energy group structure as the former ORNL VITAMIN-B6 reference library (ENDF/B-VI.3 data) and was produced using similar data processing methodologies, based on the LANL NJOY-2012.53 nuclear data processing system for the generation of the nuclide cross section data files in GENDF format. Then the ENEA-Bologna 2007 Revision of the ORNL SCAMPI nuclear data processing system was used for the conversion into the AMPX format. VITJEFF32.BOLIB contains processed cross section data files for 190 nuclides, obtained through the Bondarenko (f-factor) method for the treatment of neutron resonance self-shielding and temperature effects. Collapsed working libraries of self-shielded cross sections in FIDO-ANISN format, used by the deterministic transport codes of the ORNL DOORS system, can be generated from VITJEFF32.BOLIB through the cited SCAMPI version. This paper describes the methodology and specifications of the data processing performed and presents some results of the VITJEFF32.BOLIB validation. © The Authors, published by EDP Sciences, 2017.

The importance of the design for the realization of an intense 14 MeV neutron facility devoted to test and validate materials su itable for harsh neutron environments, such as a fusion reactor, is well established. The "New Sorgentina" Fusion Source (NSFS) is a project that proposes an intense D-T 14 MeV neutron source achievable with T and D ion beams impinging on 2 m radius rotating targets. NSFS may produce about 10 15 n/s at the target and has to be intended as an European facility that maybe realized in a few years, once provided a preliminary technological program devoted to the operation of the ion source in continuous mode, target heat loading/removal, target and tritium handling, inventor as well as site licensing. In this contribution, the main characteristics of NSFS project will be presented and its possible use as a multipurpose facility outlined.

ANITA-2000 is a code package for the activation characterization of materials exposed to neutron irradiation released by ENEA to OECD-NEADB and ORNL-RSICC. The main component of the package is the activation code ANITA-4M that computes the radioactive inventory of a material exposed to neutron irradiation. The code requires the decay data library (file fl1) containing the quantities describing the decay properties of the unstable nuclides and the library (file fl2) containing the gamma ray spectra emitted by the radioactive nuclei. The fl1 and fl2 files of the ANITA-2000 code package, originally based on the evaluated nuclear data library FENDL/D-2.0, were recently updated on the basis of the JEFF-3.1.1 Radioactive Decay Data Library. This paper presents the results of the validation of the new fl1 decay data library through the comparison of the ANITA-4M calculated values with the measured electron and photon decay heats and activities of fusion material samples irradiated at the 14 MeV Frascati Neutron Generator (FNG) of the NEA-Frascati Research Centre. Twelve material samples were considered, namely: Mo, Cu, Hf, Mg, Ni, Cd, Sn, Re, Ti, W, Ag and Al. The ratios between calculated and experimental values (C/E) are shown and discussed in this paper. © Owned by the authors, published by EDP Sciences, 2016.