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The radial neutron camera (RNC) is a key ITER diagnostic system designed to measure the uncollided 14- and 2.5-MeV neutrons from deuterium-tritium (DT) and deuterium-deuterium (DD) fusion reactions, through an array of detectors covering a full poloidal plasma section along collimated lines of sight (LoS). Its main objective is the assessment of the neutron emissivity/alpha source profile and the total neutron source strength, providing spatially resolved measurements of several parameters needed for fusion power estimation, plasma control, and plasma physics studies. The present RNC layout is composed of two fan-shaped collimating structures viewing the plasma radially through vertical slots in the diagnostic shielding module (DSM) of ITER Equatorial Port 1 (EP01): the ex-port subsystem and the in-port one. The ex-port subsystem, devoted to the plasma core coverage, extends from the Port Interspace to the Bioshield Plug: it consists of a massive shielding unit hosting two sets of collimators lying on different toroidal planes, leading to a total of 16 interleaved LoS. The in-port system consists of a cassette, integrated inside the port plug DSM, containing two detectors per each of the six LoS looking at the plasma edges. The in-port system must guarantee the required measurement performances in critical operating conditions in terms of high radiation levels, given its proximity to the plasma neutron source. This article presents an updated neutronic analysis based on the latest design of the in-port system and port plug. It has been performed by means of the Monte Carlo MCNP code and provides nuclear loads on the in-port RNC during normal operating conditions (NOC) and inputs for the measurement performance analysis.

In the framework of the ITER Physics Department of the EUROfusion Consortium, JET mission is focused on preparing for the next step devices by developing high performance scenarios and testing reactor relevant techniques and technologies. In terms of scenario development, a complete scan in isotopic composition has already started and, according to the present schedule, should include full T operation and culminate in a 50/50 DT campaign by 2020. In DD, significant increases in input power and various adjustments in fueling have allowed reaching the IPBp8(y,2) scaling up to 3 MA. A complete set of diagnostic upgrades is being implemented to support operation and to guarantee adequate scientific exploitation of the experiments. With regard to the technology, the main enhancements and specific tests are focused on maximizing the returns from the unprecedented 14 MeV neutron field and the tritium fuel cycle.

The solution of the problem of heat exhaust has been pointed out as one of the main challenge towards the realization of magnetic confinement fusion. In the last years, two concepts have been proposed in alternative to the conventional divertor solution adopted for ITER: modification of the magnetic topology in the divertor region and liquid metal as plasma facing component. The role of the Divertor Tokamak Test facility (DTT) in the power exhaust implementation strategy is discussed. The evolution of the project, since the original proposal in 2015 to the present design, is shown. The DTT facility is well integrated in the European strategy and the final decision on the divertor configuration will be made, within 2022-23, on the basis of the indication of the Power Exhaust Group constituted by the EUROfusion Consortium. Finally, the main milestones and the timeline of the project are illustrated. © 2019 Elsevier B.V.

Neutronics experiments are carried out at the Joint European Torus (JET) in preparation for the forthcoming high performance Deuterium-Tritium (D-T) campaign. Among others, activation foil measurements are performed in order to evaluate neutron streaming in various positions within the torus hall far from the plasma source. In the present study, the neutron fluence measurements performed using activation foils during the JET Deuterium-Deuterium (D-D) campaign are discussed. The activation foil results in terms of reaction rates are compared against Monte Carlo simulations and a satisfactory agreement is observed. Moreover, the pre-analysis of the experiments to be performed in the forthcoming JET Tritium-Tritium (T-T) and Deuterium-Tritium (D-T) campaigns is presented. The results of the study provide important data and support the preparation of the experimental activities in view of the planned JET D-T operation, allowing the exploitation of the unique 14 MeV neutron yields anticipated. Furthermore, they contribute to the validation of the tools employed in nuclear analyses, which are fundamental for the design and safety of ITER and future fusion power plants.

An experimental nuclear fusion device could be seen as a step toward the development of the future nuclear fusion power plant. If compared with other possible solutions to the energy problem, nuclear fusion has advantages that ensure sustainability and security. In particular, considering the radioactivity and the radioactive waste produced in a nuclear fusion plant, the component materials for the plant could be selected in order to limit the decay period, making recycling possible in a new reactor after about 100yr from the beginning of decommissioning. To achieve this and other pertinent goals, many experimental machines have been developed and operated worldwide in the last decades, underlining that radiation protection and worker exposure are critical aspects of these facilities due to the high-flux, high-energy neutrons produced in the fusion reactions. Direct radiation, material activation, tritium diffusion, and other related issues pose a real challenge to demonstrating that these devices are safer than nuclear fission facilities. In Italy, for the past 30yr, a limited number of fusion facilities have been constructed and operated, mainly at the ENEA Frascati Center, where a new one, the Italian Divertor Tokamak Test Facility (DTT), is now under development. The radiation protection approach, addressed by national licensing requirements, shows that respecting the constraints for worker exposure to ionizing radiation is not always straightforward. In the current analysis the main radiation protection issues encountered in the Italian fusion facilities are considered and discussed, and the technical and legal requirements are described. The licensing process for this kind of device is outlined and compared with that of other European countries. The following aspects are considered throughout the current study: description of the installation, plant, and systems; suitability of the area; buildings and structures; radioprotection structures and organization; exposure of personnel; accident analysis and relevant radiological consequences; and radioactive waste assessment and management. In conclusion, the analysis points out the need for special attention to the radiological exposure of workers in order to demonstrate at least the same level of safety as that reached at nuclear fission facilities.

As a demonstration fusion power plant, EU DEMO has to prove the maturity of fusion technology and its viability for electricity production. The central requirements for DEMO rest on its capability to generate significant net electric power to the grid (300 MW to 500 MW) safely and consistently. Plant availability and lifetime will approach that of a commercial fusion power plant. Operating at such regimes presents many complex challenges, of which one is plasma exhaust. To mitigate the risk that the implementation in preceding experimental devices, namely ITER, does not extrapolate to the requirement of DEMO, alternative solutions must be sought. The investigation of alternative divertor configurations was born out of this motive, seeking to resolve a ‘critical’ challenge for the realisation of DEMO. In this paper, we study the neutronics performance of three concepts: Single Null (SN), Super-X (SX) and X-divertor (XD). This is the first time a preliminary analysis of alternative configurations to the SN baseline has been performed. The shielding proposals and design recommendations presented herein should be integrated with other engineering and physics constraints in future iterations of the chosen divertor concept.

A conceptual design and performance studies for a collimated neutron flux monitor and neutron spectrometer for the Divertor Tokamak Test (DTT) facility are presented. This study is based on the single-null divertor configuration and for “Half Power” and “Full power” scenarios with 15 MW of negative-ion NBI, 29 MW of ECH and 3 MW of ICRF heating with a maximum neutron yield of 1.5 × 10 17 s − 1 . Fast ion distributions (both from auxiliary heating systems and fusion born) have been simulated in TRANSP/NUBEAM and the corresponding neutron energy spectra have been calculated using DRESS. Synthetic diagnostics have been implemented to determine the neutron fluxes and spectra at the detector location. Neutron emissivity profiles, plasma position, core ion temperature and the ratio of thermal and non-thermal D ion populations can be obtained with good accuracy and time resolution.

The Divertor Tokamak Test facility (DTT) will be a superconductive machine that aims at studying the power exhaust issue in a DEMO relevant environment. A significant power across the separatrix (compared to the radius of the machine), PSep/R ≅15 MW/m is of primary importance to fulfill this requirement and it can be obtained by injecting to the plasma up to 45 MW of additional power. This amount of power will be reached with a mix of Heating and Current Drive (HCD) systems: Electron Cyclotron Resonance Heating (ECRH), Ion Cyclotron Resonance Heating (ICRH) and Negative Neutral Beam Injection (NNBI). Power supply and wave generators of HCD will be hosted in remote-control buildings around the DTT Tokamak hall, and thus several ducts have been designed to connect the systems to the machine. The evacuated transmission lines will pass through dedicated holes in the main building making almost empty and big penetrations which cause a large neutron streaming through the walls. This work is devoted to a three-dimensional neutronics study, in support of the DTT licensing procedure, addressing the neutron and gamma streaming through the ECRH penetrations in order to optimize and determine the effectiveness of several proposed shielding options required to protect workers, population and auxiliary systems. Neutron and gamma transport simulations were carried out with the MCNP5 Monte Carlo code. Surface sources of neutrons and gammas have been used to propagate the radiation outside the penetrations. The ADVANTG hybrid transport code has been also used for variance reduction techniques in order to obtain statistically significant results in regions located several meters away from the plasma source. The results of the calculations are presented and discussed as well as the effectiveness of the proposed options to mitigate the radiation streaming and to reduce the doses outside the tokamak hall.

In a demonstrational fusion power plant (DEMO), divertor is supposed to protect vacuum vessel and superconducting magnets against neutron flux in the bottom region of the vessel. The vessel is subject to a strict design limit in irradiation damage dose and the magnets in nuclear heating power, respectively. Thus, the DEMO divertor must have the capability to protect sufficiently the vessel and the magnets against neutron flux being substantially stronger than in ITER. In this paper, a first systematic neutronics study for the European DEMO divertor is reported. Results of the extensive assessment of key nuclear loading features (nuclear heating, irradiation damage & helium production) are presented for two optional concepts, namely, dome and shielding liner including minor geometrical variants. The shielding performance of the two competing design options is discussed together with the case of a bare cassette (no shielding), particularly in terms of damage dose compared with the design limits specified for the European DEMO. It was found that both the dome and shielding liner were able to significantly reduce the nuclear loads in the cassette body and the vessel. The maximum damage dose at the end of the lifetime remained subcritical for the cassette body for both cases whereas it exceeded the limit for the vessel under the dome, but only locally on the surface underneath the pumping duct. But, the damage could be reduced below the limit for the vessel by increasing the size of the dome or by deploying the shielding liner. The most critical feature was the excessive damage occurring in the own body of the shielding components where the maximum damage dose in the steel heat sink of the dome and the shielding liner far exceeded the design limit at the end of the lifetime.

The joint evaluated fission and fusion nuclear data library 3.3 is described. New evaluations for neutron-induced interactions with the major actinides U , U and Pu , on Am and Na , Ni , Cr, Cu, Zr, Cd, Hf, W, Au, Pb and Bi are presented. It includes new fission yields, prompt fission neutron spectra and average number of neutrons per fission. In addition, new data for radioactive decay, thermal neutron scattering, gamma-ray emission, neutron activation, delayed neutrons and displacement damage are presented. JEFF-3.3 was complemented by files from the TENDL project. The libraries for photon, proton, deuteron, triton, helion and alpha-particle induced reactions are from TENDL-2017. The demands for uncertainty quantification in modeling led to many new covariance data for the evaluations. A comparison between results from model calculations using the JEFF-3.3 library and those from benchmark experiments for criticality, delayed neutron yields, shielding and decay heat, reveals that JEFF-3.3 performes very well for a wide range of nuclear technology applications, in particular nuclear energy., Acknowledgements The authors would like to thank all contributors to the JEFF project that shared their insights and provided their specific input for the present JEFF-3.3 evaluation and its immediate predecessors JEFF-3.2 and JEFF-3.1.2.

The nuclear heat loads due to gamma rays emitted from the decay of 16N and delayed neutrons from17N, generated by the activation of water in cooling circuits, are critical for ITER design. The assessment of nuclear heating from activated water is complex; it requires temporal and spatial dependent transport and activation calculations taking into account variation of irradiation, water flow conditions and cooling circuits’ parameters. A water activation experiment has been recently conducted at the14 MeV Frascati Neutron Generator (FNG) in order to validate the methodology for water activation assessment used for ITER and to reduce the safety factors applied to the calculation results, which have a large impact on the schedule, commissioning and licensing. Water circulating inside an ITER First Wall (FW) mock-up was irradiated with 14 MeV neutrons and then measured using a large CsI scintillator detector. The system consists of a closed water loop where the cooling water, transiting through an ITER FW mock-up, is irradiated by FNG. The induced 16N activity via 14 MeV neutrons interactions with 16O via the 16O(n,p)16N reaction is measured in a dedicated counting station via an expansion volume. The water then passes to a much larger holding delay tank, and after several 16N half-lives decay time, it is then recirculated and exposed again to neutrons in the ITER First Wall (FW) mock-up. The measured 16N activity is obtained measuring the emitted characteristic 6.13 and 7.12 MeV gamma-rays. Calculations were performed in an accurate model of the FW mock-up using the MCNP Monte Carlo code and FENDL-3.1 nuclear data library to obtain the predicted flux impinging on the water. The EASY-2007 inventory code was used to predict the 16N activity. In this work, a comparison between measurements and calculations is reported together with associated uncertainty analysis.

The Water Cooled Lithium Lead (WCLL) is one of the breeding blanket (BB) concept candidate for the future European Demonstration Fusion Power Reactor (DEMO). In the frame of the EUROfusion Consortium, a WCLL BB mock-up neutronics experiment will be performed at the 14 MeV Frascati Neutron Generator (FNG) irradiation facility, devoted to the validation of the Tritium Breeding Ratio (TBR) and the shielding capability predicted by neutronics design calculations. It will include the assessment of the uncertainties on the calculated Tritium Production Rate (TPR) and the neutron/gamma attenuation which will be useful to evaluate design margins required to guarantee tritium self-sufficiency. In preparation of the future experiment, a detailed pre-analysis has been performed with the MCNP Monte Carlo code for the definition of the mock-up assembly and for the optimisation of the experimental setup. The WCLL mock-up has been designed in such a way that the essential nuclear features of the WCLL BB in DEMO can be represented in the experiment and the relevant nuclear quantities can be measurable with sufficient accuracy under FNG irradiation conditions, with several experimental techniques. This study reports the results of the pre-analysis: the optimised experimental layout and its representativeness of WCLL DEMO blanket.

Three-dimensional neutronics, activation and shutdown dose rate analyses were performed with MCNP5 Monte Carlo code, FISPACT-II inventory code and Advanced D1S dynamic tool for the design and licensing of Divertor Tokamak Test facility (DTT). Advanced shielding concepts and mitigation strategies have been studied to guarantee sufficient protection of the superconducting coils and to reduce the streaming and the neutron-induced radioactivity. The present nuclear design study provides main outcomes for the loads assessment, shielding and materials requirements and on maintenance strategy and storage of activated components.

In the frame of the EUROfusion roadmap, the development of a conceptual design for the Demonstration Fusion Power Reactor (DEMO), beyond ITER, is a key issue. The DEMO reactor shall guarantee the tritium self-sufficiency, generate electricity and operate as a test facility for the fusion power plant relevant technologies, such as the breeding blanket (BB). The Water Cooled Lithium Lead (WCLL) concept has been chosen as a candidate for the DEMO BB: it relies on liquid Lithium Lead as breeder and neutron multiplier, Eurofer as structural material and pressurized water as coolant. A detailed MCNP model of the latest WCLL BB layout has been generated and integrated in the DEMO MCNP generic model suitably designed for neutronic analyses. Three-dimensional neutron and gamma transport simulations have been carried out by means the MCNP Monte Carlo code and JEFF nuclear data libraries in order to assess the WCLL-DEMO performances in terms of tritium self-sufficiency and shielding effectiveness to protect the vacuum vessel and the toroidal field coils. Moreover, the impact on the Tritium production of the water content in the first wall (FW) and the effect of its pressure/temperature has been addressed. The outcomes of the present study provide guidelines for the optimization of the WCLL DEMO reactor nuclear performances through the assessment of the loads on sensitive components and the estimation of its potential tritium generation capabilities.

The EUROfusion Work Package JET3 programme, established to enable the technological exploitation of the future Deuterium-Tritium (DT) operations at JET over the next years, includes, within the NEXP subproject, a novel Shutdown Dose Rate (SDR) benchmark experiment. The measurement of the SDR due to neutron activation in a fusion machine operating with Deuterium and Tritium is of primary importance for planning its operation in respect of dose limits for the external radiation exposure. The next high-performance DT campaign at JET (DTE2) which will follow a Tritium-Tritium (TT) operation phase, is a unique opportunity to validate the numerical tools for ITER shutdown dose rate analysis, through the comparison between numerical predictions and measured quantities in terms of C/E (Calculation/Experiment) ratios. Within this framework, this work is a pre-analysis of the impact of the future campaigns on the shutdown dose rate. Three-dimensional simulations are performed with the R2Smesh (Rigorous 2-step) approach at the end of DTE2 to predict dose rate levels at different cooling times. The results are presented and discussed in the paper with the major objective to contribute to the optimization of the planned SDR benchmark experiments.

Water activated by 14 MeV neutrons emits both high energy gammas from 16N and delayed neutrons from 17N. The isotopes 16,17N are produced through the 16O(n,p)16N (T1/2 = 7.13 s) and 17O(n,p)17N (T1/2 = 4.1 s) reactions, respectively. These sources of radiation represent one of the main issues for ITER having large impact on the design of the machine and on the qualification of components. To validate the calculation methodology used for ITER and to reduce the safety factors applied to the calculation results a water activation experiment was commissioned by F4E to ENEA and performed at the 14 MeV Frascati neutron generator (FNG). The experimental set-up consisted of a closed water loop, about 40 m long, where the cooling water, while transiting through an ITER FW mock-up, was irradiated by the D-T neutrons produced at FNG. 6.73 MeV gamma-rays from 16N and delayed (fast) neutrons from 17N were measured using a large CsI gamma-ray detector and an array of absolutely calibrated 3He detectors, respectively. This paper reports on the measurements of the delayed neutrons emission from 17N. The calibration of the 3He detector array and the experimental set-up are first discussed. Then the experimental results (E) for the delayed neutrons as well as the comparison with the calculation (C) performed using the MCNP Monte Carlo code (coupled to the FENDL3.1 data library) and the EASY-2007 inventory code are presented. The C/E ratio results close to unity, however it is affected by a very large error (> ±60 %) to be ascribed mainly to the uncertainty on the nuclear data used for calculation.

The Divertor Tokamak Test facility is an Italian experimental facility under design and construction at ENEA C.R. Frascati. The main goal of DTT is to provide an integrated environment, relevant to DEMO, where testing possible solution to the power exhaust problem in a tokamak (like for example: i) Plasma facing components technology ; ii) Plasma and divertor shape; iii) impurity seeding to increase radiation). In this respect, DTT has been designed to be flexible and adopting technologies relevant to DEMO. After its initial inception in 2015, concluded with the publication of the "DTT project proposal", a complete re-baseline has been provided oncluded with the publication of the "DTT Interim Design Report" in 2019, aimed at accommodating the request of flexibility coming from the international fusion community. During 2019, the engineering integration activity has started and the first construction contracts have been signed. This paper provides an overview of the integrated design activity towards the realization of the facility within 2025.

Abstarct The Divertor Tokamak Test facility (DTT) will operate with d -D plasmas producing pulses lasting about 60 s and producing a neutron yield up to 1.5 × 1017 n/s. Due to the high neutron yield, a significant activation of the structures and residual radioactivity after shut-down will be present. Consequently, DTT will need a robust radiation protection system to prevent undue worker and population exposure. The objectives of this system must be met during normal operation, maintenance, decommissioning and in emergency situations, thus several safety features must be routinely guarantee, among which the radiation monitoring system is probably the most crucial. This paper presents the general requirements of this monitoring system, giving some details about the preferred characteristics for the involved devices and monitors. The monitoring system shall be composed by portable instruments, passive ambient dosimeters and a limited number of fixed active measuring stations, suitable to detect and record the appropriate dosimetric quantities both for neutrons and photons. The active measuring stations and the passive dosimeters will constitute two monitoring networks designed in such a way to cover the area of interest to radiological survey around DTT.

In the frame of the EUROfusion Consortium programme, the Water Cooled Lithium Lead (WCLL) option has been chosen as a candidate for the breeding blanket (BB) of the European fusion power demonstration plant (DEMO) conceptual design. Neutronic analyses play a fundamental role in the development of the WCLL blanket, providing guidelines for its design based on the evaluation of the nuclear performances. A detailed three-dimensional MCNP model of the latest WCLL layout has been generated and integrated in a DEMO MCNP generic model suitably designed for neutronic analyses. Three-dimensional neutron and gamma transport simulations have been performed using the MCNP5v1.6 Monte Carlo code and JEFF 3.2 nuclear data libraries, in order to assess the WCLL-DEMO tritium self-sufficiency and the shielding capabilities of the breeding blanket/manifold system to protect the vacuum vessel and toroidal field coils. Furthermore, radial profiles of the neutron flux, nuclear heating, neutron damage and he-production have been assessed in the inboard and outboard equatorial planes. The outcome of the present study highlights the potential and suitability of the WCLL breeding blanket for the application to DEMO, both in terms of tritium production and shielding performances.

The ITER In-Vessel Viewing System (IVVS) consists of six identical units located at the B1 level of the Tokamak complex, at lower ports 3, 5, 9, 11, 15 and 17. It can be used for visual inspection of the plasma chamber in between plasma pulses. This work is devoted to the nuclear analyses performed for the latest design of the IVVS, to assess the shutdown dose rate (SDDR) in the port cell (PC) and the gallery at 1 day after shutdown in order to verify the design limit in the maintenance area. This analysis was performed through a combined use of the R2S method, for the activation of components from the vacuum vessel VV up to the bioshield, and Advanced D1S method for activation of components from the bioshield up to the gallery. Several MCNP calculations have been performed to transport neutrons and decay gammas up to the bioshield, then, by means of secondary sources, the radiation field has been extended in the PC up to the gallery. This strategy allowed to provide high resolution (voxel size 10 × 10 × 10 cm3) maps of shutdown dose rate in the PC with a statistical uncertainty within ±1% and in the gallery with a statistical uncertainty within ±20%. The consistency of the SDDR results obtained by the two codes at the interface highlighted the reliability of the coupled techniques used in the present analyses. Moreover it is assessed that the design target of 10 μSv/h in the PC is satisfied for the IVVS port.

Fast neutron induced water activation will result in significant additional nuclear heating loads to cryogenic components in the ITER fusion reactor during operation, which are considered as an important issue in the ITER design. A water activation experiment has been proposed at the Joint European Torus (JET) which would enable measurements of radiation dose rates and 16 N concentrations in water, activated in the most representative fusion environment achievable, and extrapolability to ITER conditions. This paper presents a study aimed at establishing the experiment feasibility, in terms of the identification and selection of an experimental location at JET, and the definition of an experimental setup enabling accurate measurements of the quantities of interest. A suitable experimental location in JET was identified in the basement of the JET Torus Hall. The expected 16 O ( n , p ) 16 N activation rates in the cooling water in the JET tokamak were determined through Monte Carlo particle transport calculations and the induced 16 N activity in the water was propagated to the experimental location. Subsequent calculations were performed to determine the expected detector response to 16 N γ rays for three different JET pulse scenarios. Two scintillator γ detectors were calibrated with standard γ calibration sources and tested with a 244 Cm/ 13 C neutron source, also emitting 6.13 MeV γ rays, corresponding to the 16 N γ ray energy. The presented work demonstrates the feasibility of a water activation experiment at JET and provides valuable information for the experiment design.

The Divertor Plasma Facing Components (PFCs) of a fusion reactor play a key role in the removal of plasma power exhaust and impurity particles, being exposed to very intense thermal load and high neutron irradiation, which can severely compromise their functioning capability. Therefore, the choice of the plasma facing materials (PFMs) is a fundamental aspect in the design of these components. Tungsten is presently the main candidate being characterised by several advantages from a thermo-physical point of view, but also some drawbacks (e.g. significant activation under neutron irradiation). Neutronics assessment plays a key role in the design of these critical components and in the choice of the most suitable PFMs. Detailed nuclear analyses have been performed by implementing, for the first time, heterogeneous models of the actual ITER-Like PFCs geometry [ 1 , 2 ], aimed to assess the impact of neutronics and activation issues on design, lifetime, operations and safety of the PFCs of DEMO divertor and to provide outcomes useful in the PFCs selection concept, under study within the EUROfusion WPDIV-2 project. In particular way, the effects of two different blankets (WCLL and HCPB) on W activation and the impact of three different W compositions and chromium, have been compared. Three-dimensional neutronic analyses have been performed with the MCNP5 Monte Carlo [ 3 ] using the reference JEFF 3.3 nuclear data libraries [ 4 ] to calculate neutron spectra in PFCs subcomponents. Activation analysis has been performed with FISPACT II inventory codes [ 5 ] to assess the specific activity, decay heat and contact dose at the end of DEMO-1 operations. The calculations have been carried-out on ENEA HPC CRESCO6 cluster (Ponti et al., 2014).

The development of a conceptual design for the Demonstration Fusion Power Reactor (DEMO) is a key issue within the EUROfusion roadmap. The DEMO reactor is designed to produce a fusion power of about 2 GW and generate a substantial amount of electricity, relying on a closed tritium fuel cycle: it implies that the breeding blanket (BB) shall guarantee a suitable tritium production to enable a continuous operation without any external supply. The Water-Cooled Lithium Lead (WCLL) concept is a candidate for the DEMO BB: it uses liquid Lithium Lead as breeder and neutron multiplier and water in PWR condition as coolant. The neutronics analyses carried out in the past have been performed using a semi-heterogeneous representation of the BB, since the complexity of its structure makes the generation of a detailed MCNP model a very demanding and challenging task. Results highlighted good performances for the WCLL BB, both in terms of shielding effectiveness and tritium self-sufficiency. A recently updated assessment of the tritium breeding ratio (TBR) requirement for DEMO, considering margins for calculation uncertainties and incomplete models of the whole machine, led to the definition of a tentative 1.15 value for the TBR. Moreover, the implementation of an accurate BB neutronics model, consistent with the engineering design, is recommended for the evaluation of the tritium self-sufficiency. In order to tackle these issues, an MCNP model of the DEMO tokamak, integrating a fully heterogeneous WCLL BB has been developed for the first time, including an accurate description of the FW water channels, as well as a comprehensive definition of the breeding zone inner structure. A complete assessment of the WCLL BB nuclear performances, through 3D neutron and gamma transport simulations, has been carried out by means the MCNP Monte Carlo code and JEFF nuclear libraries.

The ITER In-Vessel Viewing System (IVVS) consists of six identical units located at the B1 level of the Tokamak complex. In this work relevant nuclear quantities in the Port Cell (PC) have been calculated by means of the MCNP Monte Carlo code in a full 3-D geometry, including the IVVS and its shielding blocks. A comprehensive MCNP model of the PC has been developed including a detailed description of the Bioshield plug, pipes, penetrations, cask rails and PC door. The neutron and gamma sources needed to perform the nuclear analyses include both contributions from the radiation streaming through the Lower Port and activated water. Monte Carlo calculations have been performed to assess the radiation field inside the PC through maps of neutron and gamma fluxes. Absorbed dose on sensitive components over the ITER lifetime has been calculated in the PC area, in order to estimate the nuclear loads that the installed equipment has to withstand. Moreover, the shutdown dose rate distribution at 1 day after shutdown has been calculated.

Water-cooled lithium-lead breeding blanket is considered a candidate option for European DEMO nuclear fusion reactor. ENEA and the linked third parties have proposed and are developing a multi-module blanket segment concept based on DEMO 2015 specifications. The layout of the module is based on horizontal (i.e. radial-toroidal) water-cooling tubes in the breeding zone, and on lithium lead flowing in radial-poloidal direction. This design choice is driven by the rationale to have a modular design, where a basic geometry is repeated along the poloidal direction. The modules are connected with a back supporting structure, designed to withstand thermal and mechanical loads due to normal operation and selected postulated accidents. Water and lithium lead manifolds are designed and integrated with a consistent primary heat transport system, based on a reliable pressurized water reactor operating experience, and the lithium lead system. Rationale and features of current status of water-cooled lithium-lead breeding blanket design are discussed and supported by thermo-mechanics, thermo-hydraulics and neutronics analyses. Open issues and areas of research and development needs are finally pointed out.

The ITER Radial Neutron Camera (RNC) is a multichannel detection system hosted in the Equatorial Port Plug 1 (EPP 1) designed to provide information on the neutron source total strength and emissivity profiles. It consists of two sub-systems: the ex-port line-of-sights (LOSs), covering the plasma core, embedded in a massive shielding block located in the Port Interspace, and the in-port LOSs distributed in two removable cassettes integrated inside the Port Plug. Presently, the RNC layout development process is undergoing a System Level Design phase: several preliminary architectural options based on a System Engineering work have been defined: a detailed nuclear analysis of these options has been performed through radiation transport calculations with the MCNP Monte Carlo code. The radiation environment at the detectors positions has been fully characterized through the evaluation of the expected neutron spectra and the secondary gamma background and the analysis of the 3D radiation maps. Moreover, the impact of a reduced ex-port shielding block on the neutron and gamma spectra has been investigated. The results of the present study provide guidelines for the development of the RNC final design and the necessary data for the measurement performance analysis.

The design and operation of ITER experimental fusion reactor requires the development of neutron measurement techniques and numerical tools to derive the fusion power and the radiation field in the device and in the surrounding areas. Nuclear analyses provide essential input to the conceptual design, optimisation, engineering and safety case in ITER and power plant studies. The required radiation transport calculations are extremely challenging because of the large physical extent of the reactor plant, the complexity of the geometry, and the combination of deep penetration and streaming paths. This article reports the experimental activities which are carried-out at JET to validate the neutronics measurements methods and numerical tools used in ITER and power plant design. A new deuterium-tritium campaign is proposed in 2019 at JET: the unique 14 MeV neutron yields produced will be exploited as much as possible to validate measurement techniques, codes, procedures and data currently used in ITER design thus reducing the related uncertainties and the associated risks in the machine operation.

Summary The water-cooled lithium–lead breeding blanket is a candidate option for the European Demonstration Power Plant (DEMO) nuclear fusion reactor. This breeding blanket concept relies on the liquid lithium–lead as breeder–multiplier, pressurized water as coolant, and EUROFER as structural material. The current design is based on DEMO 2015 specifications and represents the follow-up of the design developed in 2015. The single-module-segment approach is employed. This is constituted by a basic geometry repeated along the poloidal direction. The power is removed by means of radial–toroidal (i.e., horizontal) water cooling tubes in the breeding zone. The lithium–lead flows in a radial–poloidal direction. On the back of the segment, a 100-mm-thick plate is in charge of withstanding the loads due to normal operation and selected postulated initiating events. Water and lithium–lead manifolds are designed and integrated with a consistent primary heat transport system, based on a reliable pressurized water reactor operating experience, and the lithium–lead system. Rationale and features of the single-module-segment water-cooled lithium–lead breeding blanket design are discussed and supported by thermo-mechanic, thermo-hydraulic, and neutronic analyses. Preliminary integration with the primary heat transfer system, the energy storage system, and the balance of plant is briefly discussed. Open issues, areas of research, and development needs are finally pointed out. @EUROfusion Consortium*, 2017. *For more details see http://www.euro-fusionscipub.org/disclaimer-copyright

This paper presents the engineering aspects of the design review of the Italian DTT (Divertor Tokamak Test facility), illustrating the rationale for the design choices and focusing on the main differences with respect to the original proposal.

In the framework of EUROfusion Work Package International Cooperation R&D activities, a close collaboration has been started among University of Palermo, ENEA Brasimone and ENEA Frascati for the development of the Water Cooled Lithium Lead (WCLL) Breeding Blanket (BB) concept. In this context, a research campaign has been carried out at the University of Palermo in order to investigate the influence of First Wall (FW) cooling water configuration on the nuclear response of the WCLL BB under irradiation in EU-DEMO, in order to gain useful indications for the WCLL BB pre-conceptual designs. To this end, three-dimensional nuclear analyses have been performed by MCNP5 v. 1.6 Monte Carlo code. A semi-heterogeneous MCNP model of the WCLL BB “single module segment” concept has been used and its nuclear response has been assessed for different FW configurations, centring the attention, mainly, on global quantities as nuclear power deposition and tritium breeding ratio. The results obtained provided some information on the nuclear features of the WCLL BB concept, useful for its design optimisation. The outcomes of this parametric study are herein presented and critically discussed.

The Divertor Tokamak Test (DTT) facility is a project proposed by an Italian consortium aimed to test the physics and technology of various alternative divertor concepts in order to design a heat and power exhaust system able to withstand the large loads expected in the divertor of a DEMO fusion power plant. The DTT machine is expected to produce up to 1.0 × 1017 2.5-MeV neutrons per second through deuterium-deuterium (DD) reactions with a significant 14 MeV neutron production (1% of the total yield) due to triton burn-up during the high-performance phase. This work is devoted to a preliminary three-dimensional shielding study to optimize the DTT building for licensing purposes. Neutron and gamma transport simulations were carried-out with MCNP5 Monte Carlo code using weight windows generated with ADVANTG hybrid code. Three-dimensional model of the DTT machine and building has been developed to assess the spatial distributions of the neutron and gamma fluxes, spectra and effective doses inside and outside the tokamak hall. The wall thickness (in the range 150–250 cm) has no impact on satisfying dose limits for the public, but is important when satisfying limits for workers.

One relevant experiment selected for the DT phase at JET is the irradiation of the HCPB-TBM mock-up of ITER which will be located in front of the horizontal port of Octant-8. The neutronics performances of the HCPB-TBM will be studied through independent measurement techniques and the experimental quantities compared to the results of Monte Carlo simulation. Of paramount importance is the measurement of the tritium production rate (TPR) to be performed by using a single crystal diamond detector covered with a thin layer of 6LiF (LiDia detector). This requires the knowledge of the 6Li mass. In this work the 6Li mass calibration of the selected LiDia detector was first performed in a well characterized thermal neutron flux spectrum. Furthermore, a performance test of the detector was carried out by locating it inside a polyethylene phantom irradiated with 14 MeV neutrons at the Frascati Neutron generator. The experimental set-up was accurately simulated by using the MCNP5 code and the TPR compared to the measured one. The goal was to point out and assess the issues with the measurement method.

The latest design of the Water Cooled Lithium Lead blanket for the European DEMO reactor is based on a Single Module Segment (SMS) concept for the blanket segments: the removal of the gaps between the modules (present in the former Multi Module structure design), with the consequent reduction of the neutron streaming, improves the tritium production and the shielding capabilities. A nuclear analysis has been carried out with the Monte Carlo N-Particle transport code version 5 (MCNP5) and the Joint Evaluated Fission and Fusion nuclear data libraries version 3.2 (JEFF 3.2). A detailed three-dimensional MCNP DEMO model with the Single Module layout has been generated. Three-dimensional neutron and gamma transport simulations have been carried out in order to assess the tritium production, nuclear power deposition in each subcomponent of the blanket and nuclear heating density, neutron flux, neutron damage (dpa) and Helium production, relevant for the design of the system. The results confirm the fulfilment of tritium self-sufficiency and shielding requirements of DEMO. Furthermore, a parametric analysis has been carried out by varying the thickness of the first wall tungsten layer to evaluate how the blanket performances are influenced by the tungsten layer thickness.

The water-cooled lithium-lead breeding blanket is in the pre-conceptual design phase. It is a candidate option for European DEMO nuclear fusion reactor. This breeding blanket concept relies on the liquid lithium-lead as breeder-multiplier, pressurized water as coolant and EUROFER as structural material. Current design is based on DEMO 2017 specifications. Two separate water systems are in charge of cooling the first wall and the breeding zone: thermo-dynamic cycle is 295–328 °C at 15.5 MPa. The breeder enters and exits from the breeding zone at 330 °C. Cornerstones of the design are the single module segment approach and the water manifold between the breeding blanket box and the back supporting structure. This plate with a thickness of 100 mm supports the breeding blanket and is attached to the vacuum vessel. It is in charge to withstand the loads due to normal operation and selected postulated initiating events. Rationale and progresses of the design are presented and substantiated by engineering evaluations and analyses. Water and lithium lead manifolds are designed and integrated with the two consistent primary heat transport systems, based on a reliable pressurized water reactor operating experience, and six lithium lead systems. Open issues, areas of research and development needs are finally pointed out. © 2019 Elsevier B.V.

An application of the ADVANTG (AutomateD VAriaNce reducTion Generator) code, which combines the Denovo deterministic transport solver with the MCNP (Monte Carlo N-Particle) Monte Carlo code, to the JET3-NEXP (Joint European Torus work package 3) streaming benchmark experiment is presented in this paper. An ADVANTG input parameter variation analysis was performed in order to find optimal input parameters for the hybrid two-step workflow. ADVANTG-accelerated calculations from three different institutions (JSI (Jožef Stefan Institute), ORNL (Oak Ridge National Laboratory), and CCFE (Culham Centre for Fusion Energy)) were compared to analog MCNP simulations confirming no bias is introduced due to the use of ADVANTG. Additionally, ADVANTG-accelerated MCNP numerical simulations of the neutron fluence were compared to experimental results performed in 2016 at JET using thermo-luminescence detectors. Calculation/Experiment values from 0.7 to 13 were calculated for the experimental positions in the SW (southwest) labyrinth and SE (southeast) chimney. Using ADVANTG-generated variance reduction parameters, speed-up factors of up to 1100 relative to analog calculations were achieved. The MCNP statistical tests on track length estimator volume averaged tallies were consistently passed for all experimental locations with ADVANTG-generated variance reduction parameters. These results demonstrate that ADVANTG is capable of accelerating tally convergence at locations far from the source in streaming-dominated transport simulations of complex fusion facility models.

The conceptual design activity of the Demonstration Fusion Power Reactor (DEMO) is in progress in the Power Plant Physics and Technology (PPPT) programme, within the EUROfusion Consortium. In this work neutronics studies, fundamental for the nuclear design of DEMO, are presented for a novel design of the lower port (LP). Two possible configurations of the LP have been investigated: the vacuum pumping port, with the pumping unit located inside the port, and an empty port designed for remote handling. For both configurations 3-dimensional Monte Carlo calculations have been performed with MCNP5 to assess the neutron flux inside and around the port and the nuclear heating in sensitive components, such as the toroidal field coil conductor, the vacuum pumps, the shielding elements and the port closure plate. Different shielding configurations have been considered, by adding shielding blocks at the lower port entrance. Single and double wall port walls and closure plates with different thickness have been studied to reduce nuclear loads and neutron flux. Nuclear quantities under analysis were found to be within the limit for all the components, with the exception of the nuclear heating on the toroidal field coils. The absence of any shield of the divertor cassette pumping duct, as in the DEMO 2017 baseline configuration, is responsible for a huge amount of radiation streaming inside the pumping duct that can cause the excess of heating on the coil conductor. The use of a liner, or of an equivalent shielding component, is proposed, but further improvements are needed to keep the nuclear loads on the coil conductor within the limit.

Water activated by fast neutrons emits both high energy gammas from 16N and delayed neutrons from 17N. The radioisotopes 16,17 N are produced through the 16O(n,p)16N and 17O(n,p)17N reactions (thresholds about 10 MeV) and decay through the reactions 16N ( T 1∕2 = 7 . 13 ± 2 s) → 16 O + γ (67% @ 6.13 MeV, 5% @ 7.12 MeV) and 17N ( T 1∕2 = 4 . 173 ± 4 s) → 17 O → 16 O + n (37.7% @ 0.386 MeV, 0.6% @ 0.886 MeV, 49.8% @ 1.16 MeV, 6.9% @ 1.69 MeV), γ (3.7% @ 0.87 MeV), respectively. Activated water, flowing throughout a nuclear plant distributes the radioactivity to many critical components where personnel and/or instrumentation can be located thus depositing additional radiation dose which requires a proper shielding. This source of radiation represents an issue for a D-T fusion device like ITER which uses water as the main cooling fluid for components such as e.g. first wall and divertor. Owing to the scarce experimental information available for the cross sections of the 16O(n,p)16N and 17O(n,p)17N reactions and few benchmark validations, large uncertainties are affecting the calculations of the water activation induced by DT neutrons in a tokamak. In the attempt to improve the available nuclear data Fusion for Energy (F4E) tasked ENEA to carry on a benchmark experiment to validate the water activation calculations presently performed for ITER. This benchmark experiment was carried out in the year 2019 irradiating with the 14 MeV neutrons produced by the Frascati Neutron generator (FNG) a small size ITER First Wall mock-up. Demineralized water was circulating inside the mock-up and downstream of the irradiation zone the 16N gamma-rays decay and the 17N neutrons decay were measured using proper detectors. The experiment was simulated using the MCNP and the FISPACT codes. The obtained C/E values are in a range close to unit for both 16,17 N isotopes. However, due to the large uncertainty on the 16O(n,p)16N and 17O(n,p)17N reactions cross-sections, for some of the used cross-section data base, e.g. ENDF/B-VIII.0, the C/E is affected by large uncertainty. This is especially true for the case of 17N which production cross-section is affected by ± 60% uncertainty. In the attempt to measure and validate the cross-section data for 16,17 N production, the water activation experiment was repeated at FNG so to measure directly the 16,17 O(n,p) 16,17 N reaction cross-sections at almost monochromatic energies. The results of this new experiment are presented in this paper and compared to the data currently used by ITER project for calculating the water activation.

In the frame of the WPJET3-DT Technology project within the EUROfusion Consortium program, neutronics experiments are in preparation for the future deuterium–tritium campaign on JET (DTE2). The experiments will be conducted with the purpose to validate the neutronics codes and tools used in ITER, thus reducing the related uncertainties and the associated risks in the machine operation. This paper summarizes the status of previous shutdown dose rate benchmarks experiments and analyses performed at JET and focuses on the computational and experimental efforts conducted in preparation of the future DTE2 experiments. In particular, preliminary calculations and studies to select detectors and positions aimed to reduce uncertainties in the shutdown dose rate experiment are presented and discussed.

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.

A neutronics benchmark experiment on a pure Copper block (dimensions 60 × 70 × 60 cm3), aimed at testing and validating the recent nuclear data libraries for fusion applications, was performed at the 14-MeV Frascati Neutron Generator (FNG) as part of a F4E specific grant (F4E_FPA-395_01) assigned to the European Consortium on Nuclear Data and Experimental Techniques. The relevant neutronics quantities (e.g., reaction rates, neutron flux spectra, doses, etc.) were measured using different experimental techniques and the results were compared to the calculated quantities using fusion relevant nuclear data libraries. This paper focuses on the analyses carried-out by ENEA through the activation foils techniques. 197Au(n,γ)198Au, 186W(n,γ)187W, 115In(n,n′)115In, 58Ni(n,p)58Co, 27Al(n,α)24Na, 93Nb(n,2n)92Nbm activation reactions were used. The foils were placed at eight different positions along the Cu block and irradiated with 14 MeV neutrons. Activation measurements were performed by means of High Purity Germanium (HPGe) detector. Detailed simulation of the experiment was carried-out using MCNP5 Monte Carlo code and the European JEFF-3.1.1 and 3.2 nuclear cross-sections data files for neutron transport and IRDFF_v1.05 library for the reaction rates in activation foils. The calculated reaction rates (C) were compared to the experimental quantities (E) and the C/E ratio with relative uncertainties was assessed.

In the frame of the EUROfusion Consortium activities were launched in 2014 to develop a concept of a DEMO reactor including a large R&D program and the integrated design of the tokamak systems. The integration of the in-vessel components (IVCs) must accommodate numerous constraints imposed by their operating environment, the requirements for precise alignment, high performance, reliability, and remote maintainability. This makes the development of any feasible design a major challenge. Although DEMO is defined to be a one-of-a-kind device there needs to be in addition to the development of the IVC design solutions a remarkable emphasis on the optimization of these solutions already at the conceptual level. Their design has a significant impact on the machine layout, complexity, and performance. This paper identifies design and technology limitations of IVCs, their consequences on the integration principles, and introduces strategies currently considered in the DEMO tokamak design approach.

In the frame of the EU DEMO development program within the EUROfusion Consortium, the integration of the in-vessel components is crucial even at an early stage of the design process. The auxiliary, heating and fueling systems have to be integrated into the Breeding Blanket and, thus, they will undergo to a harsh nuclear environment during operation, and have a significant impact on the Tritium Breeding capability, loads and shielding performances. This work presents the neutronics and thermal analyses performed to support the integration studies of the pellet injection systems. This study is mainly devoted to optimize the design of a fueling line option consisting in a protrusion of the Vacuum Vessel (VV) in the inboard side supporting a guiding tube to drive the pellets along the ideal trajectory as close as possible to the First Wall. An alternative concept with a free-flight injector through the upper port, has also been studied evaluating the feasibility and the shielding needs. The three-dimensional neutronics simulations were carried-out with the MCNP5 Monte-Carlo code using a DEMO model with the integrated systems to calculate neutron fluxes, damage and nuclear heating distribution. Thermal analyses were carried-out using ABAQUS. Results of neutronics and thermal analyses on the fueling systems are presented and discussed.

The Advanced Direct 1-Step (AdD1S) is one of the most validated tools for the evaluation of Shutdown Dose Rate in complex fusion tokamak machines. The present work is built on the experience in analyzing tokamaks using prior versions of the code, where the activation analysis was limited to the first-step reactions. In the light of the neutron irradiation scenarios foreseen for fusion power reactors and the increasing importance of safety requirements, the possibility to treat the multi-step reactions introduced in the present version represents a novel and relevant development in the field. Moreover, a new Python library, PyD1S, has been developed to optimize the MCNP-FISPACT interface and data analysis. In this paper, the recent developments and the first applications are presented.

The Divertor Tokamak Test (DTT) facility, whose design phase is currently under finalization, is an Italian project aimed to investigate alternative power exhaust solutions for DEMO. It is designed to operate with significant power loads and enough flexibility to test innovative divertor configurations, different plasma edge and bulk conditions approaching, as much as possible, those planned for DEMO. Among the neutron diagnostics, a multi-channel neutron camera, most likely equipped with the liquid scintillators NE213, is foreseen to provide spatially resolved measurements of several plasma parameters needed for fusion power estimation, plasma control and plasma physics studies. This paper presents a preliminary study performed in support of the DTT neutron camera design. A detailed MCNP model representing a 20° sector of the machine integrating its main components and detectors assemblies has been developed and used for this study. Three-dimensional neutron transport simulations have been carried out by means of the MCNP Monte Carlo code coupled with the FENDL nuclear data libraries. The diagnostic design was optimized starting from the assessment of the expected detector performances obtained by using the calculated neutron fluxes and spectra and the NE213 response functions. The outcomes of this analysis provide the detectors requirements and guidelines for the development of the above-mentioned diagnostics, investigating its feasibility and suitability with the neutron emissivity foreseen for the DTT operational scenarios.

The EUROfusion Work Package JET3 programme, established to enable the technological exploitation of the JET experiments over the next years, includes, within the NEXP subproject, a novel Shutdown Dose Rate (SDR) experiment. Considering its ITER-relevance, SDR experiments at JET represent a unique opportunity to validate the numerical tools for ITER nuclear analysis, through the comparison between numerical predictions and measured quantities (C/E). Within this framework, two active gamma dosimeters based on spherical air-vented ionization chambers (ICs) have been installed in ex-vessel positions close to the horizontal ports of the tokamak in Octants 1 and 2. The first JET campaign exploited in the novel SDR experiment is the latest 5-week Deuterium-Deuterium campaign (c36b), which achieved the best results in recent years in terms of high power operation. The present work is dedicated to the analysis of dose rate measurements carried out during this campaign and after shutdown. Proper correction factors are evaluated and applied to the instrument reading, while influence quantities and error sources are analyzed in order to calculate the overall experimental uncertainty.