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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.

Two broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format, dedicated to LWR shielding and pressure vessel dosimetry applications, were generated following the methodology recommended by the US ANSI/ANS-6.1.2-1999 (R2009) standard. These libraries, named BUGJEFF311.BOLIB and BUGENDF70.BOLIB, are respectively based on JEFF-3.1.1 and ENDF/B-VII.0 nuclear data and adopt the same broad-group energy structure (47 n + 20 γ) of the ORNL BUGLE-96 similar library. They were respectively obtained from the ENEA-Bologna VITJEFF311.BOLIB and VITENDF70.BOLIB libraries in AMPX format for nuclear fission applications through problem-dependent cross section collapsing with the ENEA-Bologna 2007 revision of the ORNL SCAMPI nuclear data processing system. Both previous libraries are based on the Bondarenko self-shielding factor method and have the same AMPX format and fine-group energy structure (199 n + 42 γ) as the ORNL VITAMIN-B6 similar library from which BUGLE-96 was obtained at ORNL. A synthesis of a preliminary validation of the cited BUGLE-type libraries, performed through 3D fixed source transport calculations with the ORNL TORT-3.2 SN code, is included. The calculations were dedicated to the PCA-Replica 12/13 and VENUS-3 engineering neutron shielding benchmark experiments, specifically conceived to test the accuracy of nuclear data and transport codes in LWR shielding and radiation damage analyses. © 2016 Owned by the authors, published by EDP Sciences.

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.

The validation effort related to the comparison between the predictions of the new ANITA-IEAF code package and the activity measurements obtained from the Karlsruhe Isochronous Cyclotron is presented and discussed in this paper. The code, able to handle the numerous reaction channels for neutron energies higher than 20 MeV, has been set up for activation calculations in the frame of the IFMIF facility. The validation was possible thanks to the availability of new experimental data obtained with neutrons of energy greater than 20 MeV. In this experiment, a thick target of natural lithium was irradiated with 40 MeV deuterons. Samples of two different steels, SS-316 and F82H, pure vanadium and a vanadium alloy were irradiated in the resulting neutron spectrum and the specific activities of many radionuclides were measured at various cooling times. The results of the comparison are presented and discussed.

This paper presents the summary of the results of the comparison between the calculated and measured photons and electrons decay heat and activations of material samples irradiated at the Fusion Neutron Generator of the ENEA Research Centre of Frascati that were used for activation code validation purpose. Samples material considered are: Mo, Cu, Hf, Mg, Ni, Cd, Sn, Re, Ti, W, Ag, and Al. Material decay heat (and activity) at various times after the end of the samples irradiation have been evaluated using the ANITA-2000 activation code, by considering three different neutron activation libraries (EAF-99, FENDL/A-2, and EAF-2003) and the results have been compared with the experimental ones. The ratios between calculated and experimental values have been obtained and they are discussed in this paper.

A thermo-hydraulic analysis of high-speed free surface Li flow over a concave plate (IFMIF geometry) is performed.Simulations are done for bulk velocities between 10m/s and 20m/s using ANSYS Fluent. A pre-computed heat sourcewas imposed at the center of the curved section to simulate the interaction of a dual deuteron beam with the Li jet.LES and k- models were used for turbulence modeling and Volume of Fluid and Level Set methods were used tomodel the free surface flow. Results reported are the variation of temperature, pressure and velocities across the Lijet at various locations along the curved region. Safety margins before Li starts boiling are also predicted. All casespredicted smooth surfaces without any waves.Keywords:IFMIF, variable curvature backplate, LES, k- , RANS, thermohydraulic1. IntroductionThe purpose of the International Fusion Materials Ir-radiation Facility (IFMIF) is to test and qualify mate-rials that can be operated at high temperatures and un-der intense (neutron) radiation, conditions typical of nu-clear fusion devices. In IFMIF, material samples to beanalyzed will be exposed to an intense neutron flux cre-ated by bombarding a dual deuteron beam on a free sur-face liquid lithium (Li) flow over a concave backplate.A high speed flow is required to rapidly and continu-ously evacuate the heat deposited by the two deuteronbeams. The curvature of the backplate is necessary toprevent Li from boiling due to the high energy deposi-tion by the deuteron beams. To test the conceptual de-sign of IFMIF various mock up experiments and simu-lations on geometries similar to IFMIF were performed.In particular, Gordeev et al. [1, 2], studied the e ect ofunsteady structures generated by nozzle and straighten-ers on the free surface and preferred the use of entirelycurved back-surface for the facility. Gordeev et al. [3]studied the e ects of vortices generated in the flow dueto turbulence and relaxation of the flow bounded by noz-zle walls to a free surface flow on the stability on the Lifree surface and found good match with experiments.IPPE (Institute for Physics and Power Engineering) testcase [4] is a similar setup where the influence of the ge-ometrical design and inflow parameters were tested fortheir e ects on the surface stability of the flow.The hydrodynamic stability of the lithium jet in thetarget section is an important problem in the IFMIF tar-get creation [4]. Through the years, various design im-provements (like the use of a two-stage SHIMA nozzleinstead of a single stage one, confirmation of a stablefree surface jet flow in experiments like IPPE and con-firmation through analysis that heated Li will not boilthanks to the high pressure on the concave backplate)have now shifted the focus of the research on the sta-bility of the Li jet free surface [5]. Most of the pub-lished experiments [4, 6, 7, 8] have shown the presenceof surface waves in the bulk velocity range of 10m/sto 20m/s. Various numerical studies on the surfacestability have also demonstrated the presence of thesewaves [1, 3, 4, 9]. Gordeev et al. outline four types ofinstabilities that can a ect the free surface [1]. Basedon this classification, the reported surface perturbationscan be grouped in three categories - a) due to the insta-bility of the shear layer (Kelvin-Helmholtz instability orsudden relaxation to free surface from a wall boundedflow), b) due to impurities / solidified lithium stuck atthe nozzle exit or due to the side walls (Newton waves)and c) due to the geometry of the structure (changein backplate curvature or divergence of the nozzle andopen channel). Turbulence generated in the nozzle andat the backplate along with the possible generation of

This paper presents a synthesis of the ENEA-Bologna Nuclear Data Group programme dedicated to generate and validate group-wise cross section libraries for shielding and radiation damage deterministic calculations in nuclear fission reactors, following the data processing methodology recommended in the ANSI/ANS-6.1.2-1999 (R2009) American Standard. The VITJEFF311.BOLIB and VITENDF70.BOLIB finegroup coupled n-γ (199 n + 42 γ-VITAMIN-B6 structure) multi-purpose cross section libraries, based on the Bondarenko method for neutron resonance self-shielding and respectively on JEFF-3.1.1 and ENDF/B-VII.0 evaluated nuclear data, were produced in AMPX format using the NJOY-99.259 and the ENEA-Bologna 2007 Revision of the SCAMPI nuclear data processing systems. Two derived broad-group coupled n-γ (47 n + 20 γ-BUGLE-96 structure) working cross section libraries in FIDO-ANISN format for LWR shielding and pressure vessel dosimetry calculations, named BUGJEFF311.BOLIB and BUGENDF70.BOLIB, were generated by the revised version of SCAMPI, through problem-dependent cross section collapsing and self-shielding from the cited fine-group libraries. The validation results on the criticality safety benchmark experiments for the fine-group libraries and the preliminary validation results for the broad-group working libraries on the PCA-Replica and VENUS-3 engineering neutron shielding benchmark experiments are reported in synthesis. © Owned by the authors, published by EDP Sciences, 2013.

The PCA-Replica 12/13 (H2O/Fe) neutron shielding benchmark experiment was analysed using the TORT-3.2 3D SN code. PCA-Replica reproduces a PWR ex-core radial geometry with alternate layers of water and steel including a pressure vessel simulator. Three broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format with the same energy group structure (47 n + 20 γ) and based on different nuclear data were alternatively used: the ENEA BUGJEFF311.BOLIB (JEFF-3.1.1) and UGENDF70.BOLIB (ENDF/B-VII.0) libraries and the ORNL BUGLE-B7 (ENDF/B-VII.0) library. Dosimeter cross sections derived from the IAEA IRDF-2002 dosimetry file were employed. The calculated reaction rates for the Rh-103(n,n′)Rh-103m, In-115(n,n′)In-115m and S-32(n,p)P-32 threshold activation dosimeters and the calculated neutron spectra are compared with the corresponding experimental results. © Owned by the authors, published by EDP Sciences, 2016.

The PCA-Replica 12/13 (H2O/Fe) neutron shielding benchmark experiment was analysed using the ORNL TORT-3.2 3D SN code. PCA-Replica, specifically conceived to test the accuracy of nuclear data and transport codes employed in LWR shielding and radiation damage calculations, reproduces a PWR ex-core radial geometry with alternate layers of water and steel including a PWR pressure vessel simulator. Three broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format with the same energy group structure (47 n + 20 γ) and based on different nuclear data were alternatively used: the ENEA BUGJEFF311.BOLIB (JEFF-3.1.1) and BUGENDF70.BOLIB (ENDF/B-VII.0) libraries and the ORNL BUGLE-96 (ENDF/B-VI.3) library. Dosimeter cross sections derived from the IAEA IRDF-2002 dosimetry file were employed. The calculated reaction rates for the Rh-103(n,n′)Rh-103 m, In-115(n,n′)In-115m and S-32(n,p)P-32 threshold activation dosimeters and the calculated neutron spectra are compared with the corresponding experimental results. © 2016 Owned by the authors, published by EDP Sciences.

Nel XVIII secolo l'attenzione alla resa del sembiante umano si stempera dalla ricerca di riconoscibilità del ritratto alla definizione del carattere, della psicologia attraverso lo studio di soggetti anonimi, personaggi indagati dal vero. Questa testa di vecchia e' paradigmatica delle capacità del Gandolfi di restituire, anche con crudezza, la realtà del vivere. La qualità della sua pittura emerge anche dal confronto tra opere degli esordi e della tarda maturità. L'artista infatti nella prima produzione è in grado di confrontarsi con lo stile tiepolesco, nella freschezza della stesura e velocità di scrittura, mantenendo, nell'attenzione al vero, fedeltà alla tradizione di scuola, per poi giungere all'espressione della devozione cattolica in termini di comprensione dell'Aufklarung cattolica e di partecipazione alle richieste del nuovo gusto di fine secolo.

This paper presents the ANITA-IEAF code package for the activation characterisation of materials exposed to neutrons with energies up to 150 MeV. It computes the radioactive inventories of materials exposed to neutron irradiation, continuous or stepwise. The activity, isotopic nuclide density, decay heat, biological hazard, clearance index and gamma ray source spectra are calculated at shutdown and at different cooling times. The code package is provided with a complete database that includes neutron activation data library, decay, hazard and clearance data library, and gamma library. The paper also presents an application of the ANITA-IEAF code package to the neutron exposure characterisation for the AISI 316 liner of the Test Cell area of the International Fusion Materials Irradiation Facility.

The paper presents the main features of the Sn calculation sequence SCALENEA-1 that has been extensively used for radioactive inventories, source terms, wastes and dose rates calculation for ITER, SEAFP and IFMIF fusion machines. The sequence includes the following steps: 1. nuclear data processing (from Evaluated Nuclear Data File), to produce master and working data libraries (problem dependent); 2. radiation (n, g) transport analysis, to produce neutron and gamma spectra; 3. radiation transport results post-processing, such as: a) radioactive inventories (activation calculations with activation code packages, e.g. the ANITA-2000), b) radiation damage, c) dose rates, and d) nuclear heating. As part of the validation process for the code package and for the calculation sequence, experimental-calculation comparison has been performed using the information and the measured isotope radioactive inventories and dose rate gathered from drum containers of a “low-level waste repository”. The study was developed through the following steps: a) collection and processing of measured data (radioactivity content and dose rate), from the cemented containers of the repository (data from 35 drums were analyzed); b) decay gamma source calculation by the ANITA-2000 code package (the input data for the calculations are the measured isotope activities for each container); c) decay gamma transport calculation by the Scalenea-1 shielding Sn sequence approach (Bonami-Nitawl-Xsdrnpm-Xsdose modules of the Scale 4.4a code system, using the Vitenea-J library, based on FENDL/E-2 data) to obtain dose rates on the surfaces and at various points outside the containers ; d) comparison experimental-calculated dose rates, taking into account also the measurement uncertainties. In order to guarantee a complete Quality Assurance for codes and calculation scheme, a simulation of the radioactive containers to evaluate the dose rates was done also by using the Monte Carlo MCNP-4c code. The following conclusions can be outlined from the result analysis: - Agreement with the (gamma dose rates) experimental data: discrepancies (C-E)/E lower than 50%. - Sn (Scalenea-1) and Monte Carlo (MCNP 4c) comparison: discrepancies lower than 1%. The analyses done provided a contribution to the Validation process of the ANITA-2000 activation code package and, more extensively, of the overall calculation sequence used to perform activation calculations for ITER GSSR. As a general conclusion, the Scalenea-1 calculation sequence is able to provide reliable results for various ITER safety analyses purposes.

We address the task of evaluating the basic performances of neutron time-of-flight spectrometers, by means of an appropriate figure of merit. Calculations of neutron flux, energy resolution, and figure of merit are presented with a critical discussion on their definitions. The results of Monte Carlo simulations for two presently operating facilities, CERN n_TOF and GELINA, are commented and compared.

Antigliadin antibodies have been widely used in the screening of celiac disease. Using this test, candidates for jejunal biopsy were selected from 328 first-degree relatives of 128 adult celiac patients. All relatives geographically available were willing to participate in the study. Twenty-one turned out to be positive for antigliadin antibodies, and in 13 jejunal histology was consistent with celiac disease. In the remaining 8, the morphometric evaluation of jejunal biopsy specimens showed a mean surface to volume ratio that, although in the normal range, was significantly lower than that of other 10 relatives negative for antigliadin antibodies and 16 biopsied controls. It was concluded that antigliadin antibody testing is a valuable method for the screening of celiac disease among family members and that relatives with genetically predisposed gliadin sensitivity, without gross histological lesions but with minor morphometric abnormalities of the jejunal mucosa, may be regarded as subjects with latent celiac disease.

In the framework of the 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 (the so-called bayonet backplate concept). In this work, the results of the steady state thermomechanical analysis of the IFMIF bayonet target assembly under two different design loading scenarios (a "hot" scenario and a "cold" scenario) are briefly reported highlighting the relevant indications obtained with respect to the fulfillment of the design requirements. In particular, the analyses have shown that in the hot scenario the temperatures reached in the target assembly are within the material acceptable limits while in the cold scenario transition below the ductile to brittle transition temperature (DBTT) cannot be excluded. Moreover, results indicate that the contact between backplate and high flux test module is avoided and that the overall structural integrity of the system is assured in both scenarios. However, stress linearization analysis reveals that ITER Structural Design Criteria for In-vessel Components (SDC-IC) design rules are not always met along the selected paths at backplate middle plane section in the hot scenario, thus suggesting the need of a revision of the backplate design or a change of the operating conditions. © 2015 Elsevier B.V.

The International Fusion Materials Irradiation Facility (IFMIF), presently in its Engineering Validation and Engineering Design Activities (EVEDA) phase under the frame of the Broader Approach Agreement between Europe and Japan, accomplished in summer 2013, on schedule, its EDA phase with the release of the engineering design report of the IFMIF plant, which is here described. Many improvements of the design from former phases are implemented, particularly a reduction of beam losses and operational costs thanks to the superconducting accelerator concept, the re-location of the quench tank outside the test cell (TC) with a reduction of tritium inventory and a simplification on its replacement in case of failure, the separation of the irradiation modules from the shielding block gaining irradiation flexibility and enhancement of the remote handling equipment reliability and cost reduction, and the water cooling of the liner and biological shielding of the TC, enhancing the efficiency and economy of the related sub-systems. In addition, the maintenance strategy has been modified to allow a shorter yearly stop of the irradiation operations and a more careful management of the irradiated samples. The design of the IFMIF plant is intimately linked with the EVA phase carried out since the entry into force of IFMIF/EVEDA in June 2007. These last activities and their on-going accomplishment have been thoroughly described elsewhere (Knaster J et al [19]), which, combined with the present paper, allows a clear understanding of the maturity of the European-Japanese international efforts. This released IFMIF Intermediate Engineering Design Report (IIEDR), which could be complemented if required concurrently with the outcome of the on-going EVA, will allow decision making on its construction and/or serve as the basis for the definition of the next step, aligned with the evolving needs of our fusion community. © 2015 IAEA, Vienna.