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Modeling and simulation lie at the heart of the design process of any nuclear application. An accurate representation of the radiation environment ensures not only the feasibility of new technologies, but it also aids in operation, maintenance, and even decommissioning. With increasingly complex designs, high-fidelity models have become a necessity for design maturity. McCAD has been under development for many years at Karlsruhe Institute of Technology (KIT) to facilitate the process of generating suitable models for nuclear analyses. In this paper, an overview of the major advances in the new version of the code is presented. A novel conversion algorithm has proven to be robust in significantly reducing the processing time to generate radiation transport models, making it easier to iterate on design details. A first-of-a-kind capability to generate hierarchical void cells is also discussed with preliminary analysis showing performance gains for particle tracking.

ITER is an international collaborative effort towards the realization of fusion energy via the magnetic confinement concept. Two of the equatorial ports in the facility are dedicated to the testing of tritium breeding concepts, which is essential for the tritium self-sufficiency of future fusion reactors. The concerned Test Blanket System (TBS) consists of a Test Blanket Module (TBM) residing inside the TBM–Port Plug (TBM-PP) and its associated ancillary systems in the Tokamak facility. In this paper, the results of a full suite of nuclear analyses concerning the shielding performance of the Pipe Forest (PF) and Bioshield Plug (BP), to reflect on the evolution of their designs, are discussed. On the BP side, the design of the peripheral part has been reviewed considering the ventilation openings and butterfly doors, to assure the design compliance with the Radiation Map (RadMap) requirements for the neutron flux in the Port Cell (PC), behind the BP. On the PF side, the pipes routing and maintenance corridor door have been redesigned, by taking into account results from previously concluded nuclear analyses. The neutronics model was developed from CAD and was used to perform transport simulations in two plasma modes: on and off. For plasma-on mode, the plasma neutron field in the Port Interspace (PI) as well as behind the BP was assessed and few shielding options were explored. The responses due to decay neutrons from 17N in activated cooling water were also considered. For the plasma-off mode, the focus was shifted to further refine the ShutDown Dose Rate (SDDR) maps, which is of importance for maintenance operations that are foreseen to take place at various stages of ITER operation, in particular following the FPO-1, FPO-2, and Short operation scenarios. In addition, detailed activation analyses were carried out to provide a provisional waste classification.

One of the main aims of the European fusion programme is the design of a DEMOnstration fusion power plant (DEMO). The related work is conducted by the EUROfusion consortium and includes a strong su...

The first mirrors of optical diagnostics in ITER are exposed to high radiation and fluxes of particles which escape the plasma, in the order of 1020 m−2s−1. At the position of the mirror, the flux may still reach about 1018 m−2s−1. First mirrors are thus the most vulnerable in-vessel optical components, being subject to erosion, esp. by fast charge-exchange neutrals, or to deposition of impurities at flux rates which can reach 0.05 nm/s. The material selected for the reflecting surface must combine a high optical reflectivity in a wide spectral range and a sufficient resistance to physical sputtering during normal operation and during mirror cleaning discharges, if any is installed. Rhodium (103Rh) was identified early as a possible or even promising candidate. It combines several attractive properties, for instance a mass which leads in most cases to low sputtering yields together with an optical reflectance ( R Rh ≈ 75 % ) which is much higher than of some other options. R Rh is insensitive to large temperature changes. Rhodium is fairly inert and its low oxidation is an appreciable advantage in case of steam ingress events. The core-plasma CXRS diagnostic in ITER (UPP 3) have now turned to Rh as a baseline. The aim is to procure monocrystalline rhodium (SC-Rh) to mitigate the increase of the diffuse reflection with the damage due to physical sputtering.

For the ITER core CXRS diagnostic, a shutter is mandatory to protect the first mirror when no measurement takes place. With the continuing design development of the diagnostic and the knowledge obtained from previous shutter prototypes, the frictionless fast shutter did undergo some design changes and manufacturing improvements without degrading the performance (opening/closing 1 s). In this paper, the major modifications of the shutter design are presented. An adjustment system was developed to allow compensation of manufacturing deviations and precise positioning of the protective blade according to the final aperture position, which is not known before the final assembly with the as-built dimensions of the port plug and diagnostic shield module (DSM). The shutter components and their adjustment systems are implemented into the shutter‘s actively cooled base structure, which in turn simplifies the remote handling procedure. For the helium circuit an activity study is presented, which allows to access the nuclear pressure equipment classification and to determine the number of required purges during the ITER lifetime. With the heat loads obtained from the known heat flow on the blade and from the neutronics analysis, a thermal FE-analysis was performed on the full shutter model by which the expected heat flow at the mechanical interfaces can be given.

Several experimental activities have been conducted within the 'ACT' sub-project under the EUROfusion WPJET3 programme with the purpose to ultimately irradiate ITER materials within the JET D-T neutron environment. The latest results of these activities are presented in this work. The ITER materials include: Poloidal field coil jacket samples and toroidal field coil radial closure plate steels, EUROFER 97-2 steel, W and CuCrZr materials from the divertor, Inconel 718, CuCrZr and 316L stainless steel for blanket modules and vacuum vessel forging samples. The experimental results presented here include gamma spectrometry measurements and analysis obtained from post-irradiated samples following the 2019 C38 (D-D) campaign, where a total of 97 samples were irradiated in a newly prepared long-term irradiation station assembly, comprising 27 ITER material samples, 70 dosimetry foils and two combination-foil detectors, known as 'VERDI' detectors. Measurements using a range of dosimetry reactions that are present in the IRDFF-II nuclear data file have also been used in preparatory work to characterise the irradiation locations. The latest ITER sample measurement results are presented along with initial comparisons with corresponding neutron transport and activation calculation predictions.

The assessment of nuclear loads to the ITER Vacuum Vessel is critical in ensuring the safe and reliable operation of the machine. Radiation transport simulations were performed to compute energy deposition rates and other nuclear responses in the Vacuum Vessel due to plasma neutrons by means of MCNP6.1. Weight windows generated by ADVANTG code were used in order to accelerate convergence of nuclear heat response tallies at the required locations, to optimize computer resources and to achieve meaningful statistical errors within a reasonable timescale. The basis of the modelling was the most detailed and up-to-date ITER reference neutronics model of regular sector, so-called “C-model”, provided with specific VV regular sector components, such as the Triangular Support. Best estimates of integral heating were obtained, and corrections were applied to account for systematic errors of modelling and random uncertainties. Consideration of other contributions to the integral heat was also made, leading to an upper bound estimate of integral heating of 20.9 MW ± 2.3 MW. A complete and detailed set of 3D nuclear heat deposition maps were also produced to be used as input to subsequent thermal-hydraulics and thermal-mechanics analyses.

The role of nuclear data for neutronics analyses of Fusion Technology (FT) facilities is presented in this paper. Nuclear data evaluations of high quality are required for transport simulations, uncertainty assessments, activation and radiation damage calculations affecting the design and performance of the FT facilities, as well as safety, licensing, waste management and decommissioning issues. A corresponding programme on the nuclear data development (NDD) and qualification is conducted within the Power Plant Physics and Technology (PPPT) programme of EUROfusion supporting the development of the DEMO fusion power plant and the IFMIF-DONES neutron source. The programme builds on the achievements of pre-ceding NDD activities within the European fusion programme as addressed in the paper. Further needs for design, shielding, activation and radiation dose calculations are discussed, and recommendations are given to further improve and qualify the nuclear data base for PPPT nuclear analyses. This includes dedicated experimental activities tailored to the needs of DEMO and IFMIF-DONES.

The ITER Organization maintains neutronics reference models for nuclear analysis. The “lite” series of tokamak sector models has progressed over many years to the C-lite V1 model released at the end of 2013. Since then several model updates have been continuously implemented in various C-lite V2 models. The latest release, C-Model represents the 40° regular sector of the tokamak up to the bioshield with central upper, equatorial and lower ports, i.e. an even-port configuration. C-Model, as previous C-lite models, is based on a modular concept of model envelopes which fill all space of the torus sector. This envelope block structure has been adopted to facilitate model update management on the basis of single standalone models, which can be easily replaced in the model assembly. C-Model has been significantly improved with respect to C-lite in terms of specific system model design status and level of detail, focused on in-vessel and vessel components. The development of those updated system models was conducted according to the quality standards of ITER reference models. Modelling instructions, documentation and verification formed an essential part of the related activities.

The ITER core charge exchange recombination spectroscopy, which contains five in-port optical mirrors, is intended to transfer the visible light emitted by interaction of the plasma with the diagnostic neutral beam to the corresponding spectrometers. The first mirror (M1) is a key and the most vulnerable component of the diagnostics. In order to provide the required mirror lifetime, maintainability and structural integrity, M1 is composed of special materials, i.e. it is made of a thin 1 mm single-crystal rhodium (ScRh) plate diffusion bonded to a specially matched tungsten-copper substrate. A rhodium nanocrystalline coating (NcRh) can be an option. The paper presents the evolution of the M1 design developed for the optical layout of 2018. M1 is adjustable and cleanable with about 100÷500 procedures of 60 MHz plasma discharge. The mirror design is supported by multifield thermal, electromagnetic and structural analyses and uses experimental data of R&Ds made by Forschungszentrum Juelich, Germany. The study confirms the working capacity of the developed mirror solution.

In ITER [1] , it is foreseen to have two Test Blanket Systems (TBS) installed in two dedicated Equatorial Ports (EP) for the purpose of testing and validating different concepts of tritium breeding blankets [2] . Each TBS consists of a Test Blanket Module Port Plug (TBM PP) and its associated systems; the Pipe Forest (PF) in the Port Interspace (PI) area and the corresponding systems in the Port Cell (PC) area. While analyses have been conducted before [3] , the design has since gone through major evolution with a redesign of the PF, a new BP based on the butterfly doors concept, and an optional activated water delay tank. In this paper, the results of a comprehensive nuclear analyses that aim at assessing the neutronics performance of the revised PF and BP, will be discussed. These include calculations of neutron flux distributions and Shutdown Dose Rate (SDDR), following SA2 [4] scenario, in a baseline configuration and exploring the option of adding additional shielding plates in the PF. It has been found that the additional shielding plates reduces the neutron flux in the PI by ∼40–60% and the SDDR in the PI, due to activated components, by ∼50%.

Activities under the EURO fusion work package JET3 programme have been established to enable the technological exploitation of the planned JET experiments over the next few years, which culminates in a D-T experimental campaign, DTE-2. In the areas of nuclear technology and nuclear safety the programme offers a unique opportunity to provide experimental data that is relevant to ITER. The key purpose of the collected data will be to support bench marking and validation activities relating to neutronics and activation codes, and associated nuclear data, that are used to predict the nuclear behavior of ITER component and materials, during and after operations. This paper details the status and key issues of the ongoing ACT sub-project under work package JET3, which aims to take advantage of the large 14 MeV neutron fluence expected during JET DTE-2 to irradiate samples of real ITER materials used in the manufacturing of the main in-vessel tokamak components. The materials considered, with specified minor elemental impurity levels, include: Nb3Sn, SS316L steels from a range of manufacturers, SS304B, Alloy 660, W, CuCrZr, OF-Cu, XM-19, Al bronze, NbTi and EUROFER. The activities include provision for measurement of nuclide activities for each material and comparison against the predicted quantities through calculation with the FISPACT-II inventory code. Included here are key pre-analysis results for the selected ITER irradiation samples, and corresponding optimization of dosimetry foils (Ti, Mn, Co, Ni, Y, Fe, Co, Sc, Ta) that will be irradiated at selected positions inside JET irradiation stations in order to determine the neutron spectrum. Preliminary experimental activation results through recent JET D–D operations are discussed.

The ITER ECRH system consists of 24 gyrotrons with up to 24 MW installed millimeter wave heating power at 170 GHz, power supplies, control system, transmission lines, one Equatorial and the four Upper Launchers. With its high frequency and small beam focus the ECRH has the unique capability of driving locally current. While the Equatorial Launcher mainly acts for central heating and current profile shaping, the Upper Launchers aim on suppressing MHD instabilities, especially neoclassical tearing modes (NTM) triggering plasma disruptions. The Upper Launchers inject millimeter waves through a quasi-optical section consisting of three fixed and the front steering mirror set. The eight overlapping beams have focal points optimized for suppression of the q = 3/2 and q = 2/1 NTMs., Several project change requests required the redesign of the Upper Launchers and the connected ex-vessel system. This redesign includes a new boundary geometry of the launchers as well as a newly designed cooling system for the Blanket Shield Module (BSM), a modified flange of the BSM to the structural main frame and a refined optical design. Additionally shield blocks with integrated in-vessel waveguides were added and the closure plate with waveguide and supply line feedthroughs was adapted. Further changes, not all caused by PCRs, include newly designed ex-vessel waveguide components with a reduced aperture and redesigned ultra low-loss CVD diamond windows. Finally several components originally foreseen as off-the-shelf components have become part of the design scope. The new launcher design status is presented with selected results on numerical design validation.

The radiation protection issues linked to the activated cooling water coming from the in-vessel components of a fusion reactor to the ex-vessel regions are important design and safety constraints in the building layout definition, as demonstrated at ITER. To tackle these issues more rigorously already in the pre-conceptual design of DEMO some relevant activation and neutronic studies have been performed for the re-definition of a preliminary architecture of the tokamak complex. The design conditions of the Water Cooled Lithium Lead (WCLL) Breeding Blanket (BB) concept have been considered as well as those of the Helium Cooled Pebble Bed (HCPB). The radiation environment in those areas of the building hosting the pipes carrying activated water devoted to the cooling of the First Wall/limiter and divertor has been characterized in terms of decay photons and secondary neutron emissions, which are emitted by N16 and N17 radionuclides. In particular, the feasibility of placing radiation sensitive equipment near flowing activated water is evaluated by calculating accumulated dose in silicon and neutron flux. Viability for manual maintenance activities is assessed by calculating biological dose rate for workers due to Activated Corrosion Products (ACP). The values achieved are then compared to applicable limits. Identification of problematic areas stimulated the examination of some design proposals to improve shielding materials and components.

The electron cyclotron-heating upper launcher (ECHUL) will be installed in four upper ports of the ITER tokamak thermonuclear fusion reactor. Each ECHUL is able to deposit 8 MW power into the plasma for plasma mode stabilization via microwave beam lines. An essential part of these beam lines are the diamond windows. They are located in the upper port cell behind the bioshield to reduce the radiation levels to a minimum. The paper describes the first detailed neutronic modelling of the ECHUL port cell with ECHUL equipment. The bioshield plug is modelled including passageways for the microwave beam lines, piping and cables looms as well as rails and openings for ventilation. The port cell is equipped with the beam lines including the diamond windows, the beam lines mounting box, conduit boxes and rails. The neutrons are transported into the port cell starting from a surface source in front of the bioshield. Neutronic results are obtained for radiation levels in the port cell at different positions, mainly focusing on the diamond windows position. It is shown that the radiation level is below the limit for maintenance in the port cell. The radiation level at the diamond window is very low and should not influence its performance.

Nuclear shielding of the ITER tokamak encompasses several systems and interfaces in a complex radiation environment. Therefore any shielding design has to involve a series of structures, systems and components in an integrated approach. This is evident for the complex ex-vessel radiation environment with streaming and leakage of plasma neutrons and subsequent activation of ex-vessel structures which give raise to excessive shutdown dose rates in accessible areas of the cryostat. The paper reviews recent nuclear analyses related to the performance of primary shields and highlights challenges toward an integrated nuclear shielding design. The general need of propagation of shielding requirements is highlighted in the context of radiation cross talk due to penetrations. Radiation streaming through gaps and penetrations is a key problem in any efficient shield design. The impact on the evolving radiation environment due to several design options along streaming paths such as port gaps, as well as their modeling for nuclear analysis, is presented. Implications regarding design integration and compliance with integrated shielding requirements and ALARA dose are finally given.

The High Fidelity Monte Carlo for fusion neutronics project HIFIMC aims at providing the computational resources for the development, testing and application of advanced modeling and simulation techniques as required for the design and optimization of upcoming fusion reactors like ITER, DEMO, HELIAS and related research facilities like IFMIF or DONES. Large-scale simulations are required to provide high fidelity results to assess the performance of such facilities. These include numerous time consuming Monte Carlo simulations for describing the transport of particles (neutrons and gammas) through the complex and heterogeneous geometry. The main computational tool for such application is the MCNP Monte Carlo code, developed at LANL, US, and the coupled transport-activation tool R2Smesh, developed at KIT. Several concurrent research topics have been conducted and are reported here. Further development of R2Smesh on enhancing performance and studies on convergence issues in meshing of the activation responses are required to qualify the tools for many large nuclear performance and radiation shielding applications in fusion devices. Verification and validation of alternative codes, like GEANT 4, developed at CERN, CH, are supporting these efforts. Applications to port systems in ITER (next step fusion reactor under construction in France), breeding blankets in DEMO (Demonstration fusion reactor), full reactor of HELIAS (Helical-Axis Advanced Stellarator) and irradiation test systems of IFMIF/DONES (International Fusion Material Irradiation Facility/DEMO-oriented neutron source) are presented. It is shown that the tools on high-performance computing platforms are capable to tackle the challenging problems of radiation shielding and activation in complex geometries involving both deep penetration and radiation streaming.

To achieve the validation and testing of tritium breeding blanket concepts, mock-ups of breeding blankets, called Test Blanket Modules (TBMs), are tested in three equatorial ports of the ITER tokamak. The TBM Port Plugs consist of Frame and TBM-Sets, which will be tested sequentially, each one tailored to the specific plasma operation scenario and test objective. In case a TBM-Set is not available, it can be replaced by a Dummy TBM at any time. TBM Frame and Dummy TBM are made of stainless steel and also need to meet ITER requirements on shielding, heat removal and safety, including activation. The maintenance operations for the TBM Port Plug include various activities in the Port Cell as well in the Hot Cell. In order to assess the requirements, a detailed MCNP model of the TBM Port Plug (PP), the Pipe Forest (PF) inside the Port Interspace (PI) and the Bioshield Plug (BP) was developed and integrated into the ITER neutronics “C-Model”. This analysis concerns the neutronic behaviour of the TBM PP with two Dummy TBMs by means of activation and shut-down dose rate (SDDR) calculations. The analysis includes the evaluation of the radiation fields around the TBM PP inside the ITER tokamak at several cooling times. Two configuration scenarios have been considered, one with a PF and PF structure inside the PI and one without this structure. The results show that the contribution to the SDDR from the TBM PP at 12 days after reactor shut down is small compared to the contribution originating from the PF structure.

The Ion Cyclotron Heating and Current Drive (ICH&CD) system will reside in ITER equatorial port plugs 13 and 15. Shutdown dose rates (SDDR) within the port interspace are required to be less than 100 μSv/h at 10 6 s cooling. A significant contribution to the SDDR results from neutrons streaming down gaps around the port frame, and the mitigation of this streaming is the main subject of these analyses. An updated MCNP model of the antenna was created and integrated into an ITER reference model. Shielding plates were defined in the port gaps, and scoping studies conducted to assess their effectiveness in several configurations, based on which a front dog-leg arrangement was selected for high resolution 3-D activation analysis using MCR2S. It was concluded that the selected configuration reduced the SDDR from ∼500 μSv/h to 220 μSv/h but were still in excess of dose rate requirements. Approximately 30% of this was due to cross-talk from neighbouring ports. In addition, increased dose rates were observed in the port interspace along the lines of sight of the removable vacuum transmission lines. Design optimisation is continuing, however an integrated approach is needed with regard to ITER port plug design and the shielding of surrounding systems.

The development of high quality nuclear data, radiation detectors and instrumentation techniques for fusion technology applications in Europe is supported by Fusion for Energy (F4E) and conducted in a joint and collaborative effort by several European research associations (ENEA, KIT, JSI, NPI, AGH, and CCFE) joined to form the "Consortium on Nuclear Data Studies/Experiments in Support of TBM Activities". This paper presents the neutronics activities carried out by the Consortium. A selection of available results are presented. Among then a benchmark experiment on a pure copper block to study the Cu cross sections at neutron energies relevant to fusion, the fabrication of prototype neutron detectors able to withstand harsh environment and temperature >200 °C (artificial diamond and self-powered detectors) developed for operating in ITER-TBM as well as measurement of relevant activation and integral gas production cross-sections. The latter measured at neutron energies relevant to IFMIF (>14 MeV) and the development of innovative experimental techniques for tritium measurement in TBM. © 2015 Elsevier B.V. All rights reserved.

For the calculations of highly reliable shutdown dose rate (SDR) maps in fusion devices like a DEMO plant, the Rigorous-2-step (R2S) method is nowadays routinely applied using high-resolution decay gamma sources from initial high-resolution neutron flux meshes activating all materials in the system. This approach has been utilized in the present paper with the objective to provide SDR results relevant for RH systems of a conceptual DEMO design developed in the EU. The primary objective was to assess specific locations of interest for RH equipment inside the vessel and along the extension of maintenance ports. To this end, a provisional DEMO MCNP model has been used, featuring HCLL-type blankets, tungsten/copper divertor, manifolds, vacuum vessel with ports and toroidal field coils. The operational scenario assumed 2.1 GW fusion power and a life-time of 20 years with plant availability of 30%, where removable parts will be extracted after 5.2 years. Results of absorbed dose rate distributions for several relevant materials are presented and discussed in terms of the different contributions from the various activated components.

The methodological approach employed for the neutronics in the PPPT (Power Plant Physics and Technology) programme of EUROfusion is presented. It encompasses development works on advanced computational tools and activities related to the nuclear design and performance evaluation of the DEMO power plant including safety, maintenance, and waste management issues. Development work is conducted on Monte Carlo codes, on the CAD geometry conversion for Monte Carlo simulations, and on coupled radiation transport and activation computation systems. The role of nuclear data for reliable DEMO neutronics design analyses and uncertainty assessments is also addressed. Specific examples of nuclear analyses are presented including breeder blanket and shielding analyses for the different DEMO blanket concepts as well as related activation, decay heat and shut-down dose rate analyses.

The European approach for the development of nuclear data for fusion technology applications is presented. Related R&D activities are conducted by the Consortium on Nuclear Data Development and Analysis for Fusion to satisfy the nuclear data needs of the major projects including ITER, the Early Neutron Source (ENS) and DEMO. Recent achievements are presented in the area of nuclear data evaluations, benchmarking and validation, nuclear model improvements, and uncertainty assessments.

The 2014–2016 JET results are reviewed in the light of their significance for optimising the ITER research plan for the active and non-active operation. More than 60 h of plasma operation with ITER first wall materials successfully took place since its installation in 2011. New multi-machine scaling of the type I-ELM divertor energy flux density to ITER is supported by first principle modelling. ITER relevant disruption experiments and first principle modelling are reported with a set of three disruption mitigation valves mimicking the ITER setup. Insights of the L–H power threshold in Deuterium and Hydrogen are given, stressing the importance of the magnetic configurations and the recent measurements of fine-scale structures in the edge radial electric. Dimensionless scans of the core and pedestal confinement provide new information to elucidate the importance of the first wall material on the fusion performance. H-mode plasmas at ITER triangularity (H = 1 at ßN ~ 1.8 and n/nGW ~ 0.6) have been sustained at 2 MA during 5 s. The ITER neutronics codes have been validated on high performance experiments. Prospects for the coming D–T campaign and 14 MeV neutron calibration strategy are reviewed. This work has been carried out within the framework of the EUROfusion Consortium and has received funding from the Euratom research and training programme 2014–2018 under grant agreement No 633053. Peer Reviewed Article signat per 1.173 autors/es: X. Litaudon35, S. Abduallev39, M. Abhangi46, P. Abreu53, M. Afzal7, K.M. Aggarwal29, T. Ahlgren101, J.H. Ahn8, L. Aho-Mantila112, N. Aiba69, M. Airila112, R. Albanese105, V. Aldred7, D. Alegre93, E. Alessi45, P. Aleynikov55, A. Alfier12, A. Alkseev72, M. Allinson7, B. Alper7, E. Alves53, G. Ambrosino105, R. Ambrosino106, L. Amicucci90, V. Amosov88, E. Andersson Sundén22, M. Angelone90, M. Anghel85, C. Angioni62, L. Appel7, C. Appelbee7, P. Arena30, M. Ariola106, H. Arnichand8, S. Arshad41, A. Ash7, N. Ashikawa68, V. Aslanyan64, O. Asunta1, F. Auriemma12, Y. Austin7, L. Avotina103, M.D. Axton7, C. Ayres7, M. Bacharis24, A. Baciero57, D. Baião53, S. Bailey7, A. Baker7, I. Balboa7, M. Balden62, N. Balshaw7, R. Bament7, J.W. Banks7, Y.F. Baranov7, M.A. Barnard7, D. Barnes7, M. Barnes27, R. Barnsley55, A. Baron Wiechec7, L. Barrera Orte34, M. Baruzzo12, V. Basiuk8, M. Bassan55, R. Bastow7, A. Batista53, P. Batistoni90, R. Baughan7, B. Bauvir55, L. Baylor73, B. Bazylev56, J. Beal110, P.S. Beaumont7, M. Beckers39, B. Beckett7, A. Becoulet8, N. Bekris35, M. Beldishevski7, K. Bell7, F. Belli90, M. Bellinger7, É. Belonohy62, N. Ben Ayed7, N.A. Benterman7, H. Bergsåker42, J. Bernardo53, M. Bernert62, M. Berry7, L. Bertalot55, C. Besliu7, M. Beurskens63, B. Bieg61, J. Bielecki47, T. Biewer73, M. Bigi12, P. Bílková50, F. Binda22, A. Bisoffi31, J.P.S. Bizarro53, C. Björkas101, J. Blackburn7, K. Blackman7, T.R. Blackman7, P. Blanchard33, P. Blatchford7, V. Bobkov62, A. Boboc7, G. Bodnár113, O. Bogar18, I. Bolshakova60, T. Bolzonella12, N. Bonanomi97, F. Bonelli56, J. Boom62, J. Booth7, D. Borba35,53, D. Borodin39, I. Borodkina39, A. Botrugno90, C. Bottereau8, P. Boulting7, C. Bourdelle8, M. Bowden7, C. Bower7, C. Bowman110, T. Boyce7, C. Boyd7, H.J. Boyer7, J.M.A. Bradshaw7, V. Braic87, R. Bravanec40, B. Breizman107, S. Bremond8, P.D. Brennan7, S. Breton8, A. Brett7, S. Brezinsek39, M.D.J. Bright7, M. Brix7, W. Broeckx78, M. Brombin12, A. Brosławski65, D.P.D. Brown7, M. Brown7, E. Bruno55, J. Bucalossi8, J. Buch46, J. Buchanan7, M.A. Buckley7, R. Budny76, H. Bufferand8, M. Bulman7, N. Bulmer7, P. Bunting7, P. Buratti90, A. Burckhart62, A. Buscarino30, A. Busse7, N.K. Butler7, I. Bykov42, J. Byrne7, P. Cahyna50, G. Calabrò90, I. Calvo57, Y. Camenen4, P. Camp7, D.C. Campling7, J. Cane7, B. Cannas17, A.J. Capel7, P.J. Card7, A. Cardinali90, P. Carman7, M. Carr7, D. Carralero62, L. Carraro12, B.B. Carvalho53, I. Carvalho53, P. Carvalho53, F.J. Casson7, C. Castaldo90, N. Catarino53, J. Caumont7, F. Causa90, R. Cavazzana12, K. Cave-Ayland7, M. Cavinato12, M. Cecconello22, S. Ceccuzzi90, E. Cecil76, A. Cenedese12, R. Cesario90, C.D. Challis7, M. Chandler7, D. Chandra46, C.S. Chang76, A. Chankin62, I.T. Chapman7, S.C. Chapman28, M. Chernyshova49, G. Chitarin12, G. Ciraolo8, D. Ciric7, J. Citrin38, F. Clairet8, E. Clark7, M. Clark7, R. Clarkson7, D. Clatworthy7, C. Clements7, M. Cleverly7, J.P. Coad7, P.A. Coates7, A. Cobalt7, V. Coccorese105, V. Cocilovo90, S. Coda33, R. Coelho53, J.W. Coenen39, I. Coffey29, L. Colas8, S. Collins7, D. Conka103, S. Conroy22, N. Conway7, D. Coombs7, D. Cooper7, S.R. Cooper7, C. Corradino30, Y. Corre8, G. Corrigan7, S. Cortes53, D. Coster62, A.S. Couchman7, M.P. Cox7, T. Craciunescu86, S. Cramp7, R. Craven7, F. Crisanti90, G. Croci97, D. Croft7, K. Crombé15, R. Crowe7, N. Cruz53, G. Cseh113, A. Cufar81, A. Cullen7, M. Curuia85, A. Czarnecka49, H. Dabirikhah7, P. 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Trimble7, M. Tripsky58, M. Tsalas38, P. Tsavalas71, D. Tskhakaya jun102, I. Turner7, M.M. Turner32, M. Turnyanskiy34, G. Tvalashvili7, S.G.J. Tyrrell7, A. Uccello45, Z. Ul-Abidin7, J. Uljanovs1, D. Ulyatt7, H. Urano69, I. Uytdenhouwen78, A.P. Vadgama7, D. Valcarcel7, M. Valentinuzzi8, M. Valisa12, P. Vallejos Olivares42, M. Valovic7, M. Van De Mortel7, D. Van Eester58, W. Van Renterghem78, G.J. van Rooij38, J. Varje1, S. Varoutis56, S. Vartanian8, K. Vasava46, T. Vasilopoulou71, J. Vega57, G. Verdoolaege58, R. Verhoeven7, C. Verona95, G. Verona Rinati95, E. Veshchev55, N. Vianello45, J. Vicente53, E. Viezzer62,92, S. Villari90, F. Villone100, P. Vincenzi12, I. Vinyar74, B. Viola90, A. Vitins103, Z. Vizvary7, M. Vlad86, I. Voitsekhovitch34, P. Vondráček50, N. Vora7, T. Vu8, W.W. Pires de Sa52, B. Wakeling7, C.W.F. Waldon7, N. Walkden7, M. Walker7, R. Walker7, M. Walsh55, E. Wang39, N. Wang39, S. Warder7, R.J. Warren7, J. Waterhouse7, N.W. Watkins28, C. Watts55, T. Wauters58, A. Weckmann42, J. Weiland23, H. Weisen33, M. Weiszflog22, C. Wellstood7, A.T. West7, M.R. Wheatley7, S. Whetham7, A.M. Whitehead7, B.D. Whitehead7, A.M. Widdowson7, S. Wiesen39, J. Wilkinson7, J. Williams7, M. Williams7, A.R. Wilson7, D.J. Wilson7, H.R. Wilson110, J. Wilson7, M. Wischmeier62, G. Withenshaw7, A. Withycombe7, D.M. Witts7, D. Wood7, R. Wood7, C. Woodley7, S. Wray7, J. Wright7, J.C. Wright64, J. Wu89, S. Wukitch64, A. Wynn110, T. Xu7, D. Yadikin16, W. Yanling39, L. Yao89, V. Yavorskij102, M.G. Yoo80, C. Young7, D. Young7, I.D. Young7, R. Young7, J. Zacks7, R. Zagorski49, F.S. Zaitsev18, R. Zanino75, A. Zarins103, K.D. Zastrow7, M. Zerbini90, W. Zhang62, Y. Zhou42, E. Zilli12, V. Zoita86, S. Zoletnik113, I. Zychor65 and JET Contributorsa // EUROfusion Consortium JET, Culham Science Centre, Abingdon, OX14 3DB, United Kingdom / 1 Aalto University, PO Box 14100, FIN-00076 Aalto, Finland / 2 Aix Marseille Université, CNRS, Centrale Marseille, M2P2 UMR 7340, 13451, Marseille, France / 3 Aix-Marseille Université, CNRS, IUSTI UMR 7343, 13013 Marseille, France / 4 Aix-Marseille Université, CNRS, PIIM, UMR 7345, 13013 Marseille, France / 5 Arizona State University, Tempe, AZ, United States of America / 6 Barcelona Supercomputing Center, Barcelona, Spain / 7 CCFE, Culham Science Centre, Abingdon, Oxon, OX14 3DB, United Kingdom / 8 CEA, IRFM, F-13108 Saint Paul Lez Durance, France / 9 Center for Energy Research, University of California at San Diego, La Jolla, CA 92093, United States of America / 10 Centro Brasileiro de Pesquisas Fisicas, Rua Xavier Sigaud, 160, Rio de Janeiro CEP 22290-180, Brazil / 11 Consorzio CREATE, Via Claudio 21, 80125 Napoli, Italy / 12 Consorzio RFX, corso Stati Uniti 4, 35127 Padova, Italy / 13 Daegu University, Jillyang, Gyeongsan, Gyeongbuk 712-174, Republic of Korea / 14 Departamento de Física, Universidad Carlos III de Madrid, 28911 Leganés, Madrid, Spain / 15 Department of Applied Physics UG (Ghent University) St-Pietersnieuwstraat 41 B-9000 Ghent, Belgium / 16 Department of Earth and Space Sciences, Chalmers University of Technology, SE-41296 Gothenburg, Sweden / 17 Department of Electrical and Electronic Engineering, University of Cagliari, Piazza d’Armi 09123, Cagliari, Italy / 18 Department of Experimental Physics, Faculty of Mathematics, Physics and Informatics Comenius University Mlynska dolina F2, 84248 Bratislava, Slovakia / 19 Department of Materials Science, Warsaw University of Technology, PL-01-152 Warsaw, Poland / 20 Department of Nuclear and Quantum Engineering, KAIST, Daejeon 34141, Korea / 21 Department of Physics and Applied Physics, University of Strathclyde, Glasgow, G4 ONG, United Kingdom / 22 Department of Physics and Astronomy, Uppsala University, SE-75120 Uppsala, Sweden / 23 Department of Physics, Chalmers University of Technology, SE-41296 Gothenburg, Sweden / 24 Department of Physics, Imperial College London, London, SW7 2AZ, United Kingdom / 25 Department of Physics, SCI, KTH, SE-10691 Stockholm, Sweden / 26 Department of Physics, University of Basel, Basel, Switzerland / 27 Department of Physics, University of Oxford, Oxford, OX1 2JD, United Kingdom / 28 Department of Physics, University of Warwick, Coventry, CV4 7AL, United Kingdom / 29 Department of Pure and Applied Physics, Queens University, Belfast, BT7 1NN, United Kingdom / 30 Dipartimento di Ingegneria Elettrica Elettronica e Informatica, Università degli Studi di Catania, 95125 Catania, Italy / 31 Dipartimento di Ingegneria Industriale, University of Trento, Trento, Italy / 32 Dublin City University (DCU), Dublin, Ireland / 33 Ecole Polytechnique Fédérale de Lausanne (EPFL), Swiss Plasma Center (SPC), CH-1015 Lausanne, Switzerland / 34 EUROfusion Programme Management Unit, Boltzmannstr. 2, 85748 Garching, Germany / 35 EUROfusion Programme Management Unit, Culham Science Centre, Culham, OX14 3DB, United Kingdom / 36 European Commission, B-1049 Brussels, Belgium / 37 Fluid and Plasma Dynamics, ULB—Campus Plaine—CP 231 Boulevard du Triomphe, 1050 Bruxelles, Belgium / 38 FOM Institute DIFFER, Eindhoven, Netherlands / 39 Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung—Plasmaphysik, 52425 Jülich, Germany / 40 Fourth State Research, 503 Lockhart Dr, Austin, TX, United States of America / 41 Fusion for Energy Joint Undertaking, Josep Pl. 2, Torres Diagonal Litoral B3, 08019, Barcelona, Spain / 42 Fusion Plasma Physics, EES, KTH, SE-10044 Stockholm, Sweden / 43 General Atomics, PO Box 85608, San Diego, CA 92186-5608, United States of America / 44 HRS Fusion, West Orange, NJ, United States of America / 45 IFP-CNR, via R. Cozzi 53, 20125 Milano, Italy / 46 Institute for Plasma Research, Bhat, Gandhinagar-382 428, Gujarat State, India / 47 Institute of Nuclear Physics, Radzikowskiego 152, 31-342 Kraków, Poland / 48 Institute of Physics, Opole University, Oleska 48, 45-052 Opole, Poland / 49 Institute of Plasma Physics and Laser Microfusion, Hery 23, 01-497 Warsaw, Poland / 50 Institute of Plasma Physics AS CR, Za Slovankou 1782/3, 182 00 Praha 8, Czechia / 51 Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031, People’s Republic of China / 52 Instituto de Física, Universidade de São Paulo, Rua do Matão Travessa R Nr.187 CEP 05508-090 Cidade Universitária, São Paulo, Brasil / 53 Instituto de Plasmas e Fusão Nuclear, Instituto Superior Técnico, Universidade de Lisboa, Lisboa, Portugal / 54 Ioffe Physico-Technical Institute, 26 Politekhnicheskaya, St Petersburg 194021, Russian Federation / 55 ITER Organization, Route de Vinon, CS 90 046, 13067 Saint Paul Lez Durance, France / 56 Karlsruhe Institute of Technology, PO Box 3640, D-76021 Karlsruhe, Germany / 57 Laboratorio Nacional de Fusión, CIEMAT, Madrid, Spain / 58 Laboratory for Plasma Physics Koninklijke Militaire School—Ecole Royale Militaire, Renaissancelaan 30 Avenue de la Renaissance B-1000, Brussels, Belgium / 59 Lithuanian energy institute, Breslaujos g. 3, LT-44403, Kaunas, Lithuania / 60 Magnetic Sensor Laboratory, Lviv Polytechnic National University, Lviv, Ukraine / 61 Maritime University of Szczecin, Waly Chrobrego 1-2, 70-500 Szczecin, Poland / 62 Max-Planck-Institut für Plasmaphysik, D-85748 Garching, Germany / 63 Max-Planck-Institut für Plasmaphysik, Teilinsitut Greifswald, D-17491 Greifswald, Germany / 64 MIT Plasma Science and Fusion Centre, Cambridge, MA 02139, United States of America / 65 National Centre for Nuclear Research (NCBJ), 05-400 Otwock-Świerk, Poland / 66 National Fusion Research Institute (NFRI), 169-148 Gwahak-ro, Yuseong-gu, Daejeon 305-806, Republic of Korea / 67 National Institute for Fusion Science, Oroshi, Toki, Gifu 509-5292, Japan / 68 National Institute for Fusion Science, Toki, 509-5292, Japan / 69 National Institutes for Quantum and Radiological Science and Technology, Naka, Ibaraki 311-0193, Japan / 70 National Technical University of Athens, Iroon Politechniou 9, 157 73 Zografou, Athens, Greece / 71 NCSR ‘Demokritos’, 153 10, Agia Paraskevi Attikis, Greece / 72 NRC Kurchatov Institute, 1 Kurchatov Square, Moscow 123182, Russian Federation / 73 Oak Ridge National Laboratory, Oak Ridge, TN 37831-6169, United States of America / 74 PELIN LLC, 27a, Gzhatskaya Ulitsa, Saint Petersburg, 195220, Russian Federation / 75 Politecnico di Torino, Corso Duca degli Abruzzi 24, I-10129 Torino, Italy / 76 Princeton Plasma Physics Laboratory, James Forrestal Campus, Princeton, NJ 08543, United States of America / 77 Purdue University, 610 Purdue Mall, West Lafayette, IN 47907, United States of America / 78 SCK-CEN, Nuclear Research Centre, 2400 Mol, Belgium / 79 Second University of Napoli, Consorzio CREATE, Via Claudio 21, 80125 Napoli, Italy / 80 Seoul National University, Shilim-Dong, Gwanak-Gu, Republic of Korea / 81 Slovenian Fusion Association (SFA), Jozef Stefan Institute, Jamova 39, SI-1000 Ljubljana, Slovenia / 82 Space and Plasma Physics, EES, KTH SE-100 44 Stockholm, Sweden / 83 Technical University of Denmark, Department of Physics, Bldg 309, DK-2800 Kgs Lyngby, Denmark / 84 The ‘Horia Hulubei’ National Institute for Physics and Nuclear Engineering, Magurele-Bucharest, Romania / 85 The National Institute for Cryogenics and Isotopic Technology, Ramnicu Valcea, Romania / 86 The National Institute for Laser, Plasma and Radiation Physics, Magurele-Bucharest, Romania / 87 The National Institute for Optoelectronics, Magurele-Bucharest, Romania / 88 Troitsk Insitute of Innovating and Thermonuclear Research (TRINITI), Troitsk 142190, Moscow Region, Russian Federation / 89 University of Electronic Science and Technology of China, Chengdu, People’s Republic of China / 90 Unità Tecnica Fusione, ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati (Roma), Italy / 91 Universidad Complutense de Madrid, Madrid, Spain / 92 Universidad de Sevilla, Sevilla, Spain / 93 Universidad Nacional de Educación a Distancia, Madrid, Spain / 94 Universidad Politécnica de Madrid, Grupo I2A2, Madrid, Spain / 95 Università di Roma Tor Vergata, Via del Politecnico 1, Roma, Italy / 96 University College Cork (UCC), Ireland / 97 University Milano-Bicocca, piazza della Scienza 3, 20126 Milano, Italy / 98 University of Basilicata, Consorzio CREATE, Via Claudio 21, 80125 Napoli, Italy / 99 University of California, 1111 Franklin St., Oakland, CA 94607, United States of America / 100 University of Cassino, Consorzio CREATE, Via Claudio 21, 80125 Napoli, Italy / 101 University of Helsinki, PO Box 43, FI-00014 University of Helsinki, Finland / 102 University of Innsbruck, Fusion@Österreichische Akademie der Wissenschaften (ÖAW), Innsbruck, Austria / 103 University of Latvia, 19 Raina Blvd., Riga, LV 1586, Latvia / 104 University of Lorraine, CNRS, UMR7198, YIJL, Nancy, France / 105 University of Napoli ‘Federico II’, Consorzio CREATE, Via Claudio 21, 80125 Napoli, Italy / 106 University of Napoli Parthenope, Consorzio CREATE, Via Claudio 21, 80125 Napoli, Italy / 107 University of Texas at Austin, Institute for Fusion Studies, Austin, TX 78712, United States of America / 108 University of Toyama, Toyama, 930-8555, Japan / 109 University of Tuscia, DEIM, Via del Paradiso 47, 01100 Viterbo, Italy / 110 University of York, Heslington, York YO10 5DD, United Kingdom / 111 Vienna University of Technology, Fusion@Österreichische Akademie der Wissenschaften (ÖAW), Austria / 112 VTT Technical Research Centre of Finland, PO Box 1000, FIN-02044 VTT, Finland / 113 Wigner Research Centre for Physics, PO Box 49, H-1525 Budapest, Hungary

Europe is currently developing two reference breeder blankets concepts for DEMO reactor specifications that will be tested in ITER under the form of Test Blanket Modules (TBMs): the Helium-Cooled Lithium-Lead (HCLL) concept which uses the eutectic Pb-16Li as both breeder and neutron multiplier; the Helium-Cooled Pebble-Bed (HCPB) concept which features lithiated ceramic pebbles as breeder and beryllium pebbles as neutron multiplier. Each TBM is associated with several sub-systems required for their operation; together they form the Test Blanket System (TBS). This paper presents the state of HCLL and HCPB TBS instrumentation design. The discussion is based on the systems functional analysis, from which three main categories of instrumentation are defined: those relevant to safety functions; those relevant to interlock functions; those designed for the control and scientific exploitation of the devices based on the TBM program objectives.

The European Fusion Development Agreement (EFDA) recently launched a programme on Power Plant Physics and Technology (PPPT) with the aim to develop a conceptual design of a fusion demonstration reactor (DEMO) addressing key technology and physics issues. A dedicated part of the PPPT programme is devoted to the neutronics which, among others, has to define and verify requirements and boundary conditions for the DEMO systems. The quality of the provided data depends on the capabilities and the reliability of the computational tools. Accordingly, the PPPT activities in the area of neutronics include both DEMO nuclear analyses and development efforts on neutronic tools including their verification and validation. This paper reports on first neutronics studies performed for DEMO, and on the evaluation and further development of neutronic tools. � 2014 Karlsruhe Institute of Technology.

This paper presents an overview of the activities of the European Consortium on Nuclear Data Development and Analysis for Fusion. The Consortium combines available European expertise to provide services for the generation, maintenance, and validation of nuclear data evaluations and data files relevant for ITER, IFMIF and DEMO, as well as codes and software tools required for related nuclear calculations.

This work presents neutronic analyses to support the design of the test facility of the International Fusion Materials Irradiation Facility (IFMIF) in the framework of the Broader Approach activities. In order to evaluate the necessary biological shielding thickness and the reliability of remote handling tools for activated components inside the Test Cell, a detailed activation analysis has been carried out to obtain a detailed residual dose map resulting from the entire High Flux Test Module and Test Cell by utilizing the McDeLicious-11 Monte Carlo code and the mesh-tally based rigorous 2-Step approach (R2Smesh) developed at KIT. The absorbed dose rate expected at the working place of remote handling tools inside the Test Cell is approximately 100 Gy/h. A wall thickness of 150 cm for heavy concrete can be used as the design basis for hot cells in the test facility.

Arkady Serikov , Ulrich Fischer , Mark Henderson , Dieter Leichtle , Charles Spencer Pitcher , Peter Spaeh , Dirk Strauss , Alejandro Suarez b and Bastian Weinhorst a a Karlsruhe Institute of Technology (KIT), Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen, Germany; b ITER Organization, Route de Vinon sur Verdon, 13115, St. Paul lez Durance, France

The McCad geometry conversion tool has been developed at KIT to enable the automatic conversion of CAD models into the semi-algebraic geometry representation as utilized in Monte Carlo particle transport simulations. McCad is entirely based on open source software, it is running under the Linux operating system and utilizes the Open Cascade CAD kernel with the Qt4 libraries for the graphical user interface (GUI). The converted geometry models can be output in the syntax of the Monte Carlo codes MCNP and TRIPOLI. Related visualization capabilities are based on the coupling of McCad with the ParaView software and allow to overlay mesh tally distributions to the CAD geometry. This enables perspective 3D representations or animations on the CAD geometry. The paper presents the current status of the McCad approach and its implementation, and discusses its capabilities, limitations as well as future development needs. The use of McCad for fusion neutronics applications is illustrated on the examples of the MCNP model generation for ITER Test Blanket Modules (TBM) and the test cell facility of the IFMIF neutron source including Monte Carlo shielding calculations using the converted models.

Specific neutronics features of the ITER equatorial and upper ports are discussed in this paper as related to the design development of Diagnostic Equatorial Port Plug (EPP) and Electron Cyclotron Heating Upper Launcher (ECHUL). The focus is on the EPP analysis. Neutronics analyses are based on new calculation results of neutron/photon fluxes, nuclear heating, neutron damage and shutdown dose rates. The results were obtained with the radiation transport code MCNP5, the activation code FISPACT-2007, and the Rigorous 2 Step mesh-based (R2Smesh) interface code which allows to provide shutdown dose rate distributions with high spatial resolution. It was revealed that neutron flux levels in the port access areas, the resulting activation of the port materials and the equivalent radiation doses imposed to work personnel after ITER shutdown are strongly dependent on the openings in the port shielding blocks. For the ECHUL in the upper port, neutron fluxes have been calculated for an updated ITER model and compared with previous results. This work reveals substantial radiation cross-talk effects between the lower, the equatorial and the upper ITER ports and provides fundamental results which can be utilized for further investigations of suitable design options of ITER ports.

Several methodologies have been developed for the calculation of shut-down dose rates based on the use of the Monte Carlo (MC) technique for particle transport simulations including the rigorous tw...

The well validated Rigorous-2-Step (R2S) approach for the calculation of decay gamma shutdown dose rates has been extended utilizing mesh tally capabilities of the MCNP5 code. Contrary to other similar approaches, the new R2Smesh system pays special attention to achieving highly accurate and reliable results for neutron flux intensities and spectral shape in the initial MCNP5 neutron transport step. To this end, non-uniform fine and coarse meshes are adopted. Also several enhancements to the performance of activation calculations by FISPACT2007 and to sampling from repeated structures geometry have been implemented. The R2Smesh system has been verified by application to an ITER (International Thermonuclear Experimental Reactor) computational benchmark. An extensive validation exercise has been conducted on the FNG (Frascati Neutron Generator) shutdown dose rate benchmark. In this case several variants of the fine/coarse mesh spacing have been tested showing evidence for convergence toward R2S values obtained with the original cell-based approach. It thus has been shown that R2Smesh is validated for ITER applications and offers additional capabilities for high resolution shutdown dose rate calculations.

This paper addresses neutronics aspects of the design development of the Diagnostic Generic Equatorial Port Plug (EPP) in ITER. To secure the personnel access at the EPP back-end interspace, parametric neutronics analyses of the EPP radiation environment have been performed and practical shielding solutions have been found. Radiation transport was performed with the Monte Carlo MCNP5 code. Activation calculations were conducted with the FISPACT-2007 inventory code. The R2Smesh approach was applied to couple transport and activation calculations. Newly created EPP local MCNP5 model was devised by extracting the EPP and adjacent blanket modules from the ITER Alite-4.1 model with proper modification of the EPP geometry in accordance with recent 3D CAD CATIA model. The EPP local model reproduces the EPP neutronically important features and allows investigation of the EPP neutronics effects in isolation from all other ITER components. Thorough EPP parametric analyses revealed dominant effect of gaps around EPP and several EPP design improvements were implemented as the outcomes of the analyses. Gap labyrinths and streaming stoppers inserted into the gaps were shown are capable to reduce the shutdown dose rate which is below the 100 μSv/h limit of personnel access and by 2 orders of magnitude less than the value in the model with straight gaps.

A neutronics analysis has been performed to provide the input required for the design strategy for the In-Vessel Viewing System (IVVS) and the Glow Discharge Cleaning (GDC) plug units in the ITER tokamak. The focus of the analysis has been on operational loads to the GDC electrode head in the shielding position and on the activation and the decay photon radiation absorbed in the structural components of the entire system. To estimate the conditions for maintenance scenarios, the occupational dose rate around the isolated IVVS/GDC head has been calculated assuming the ITER SA2 irradiation scenario. The Rigorous 2 Step (R2S) method, developed previously at KIT, has been employed for the calculation of the shutdown dose rates. The GDC head, which is subjected to the highest neutron loads, gets heavily activated and dominates the decay gamma activity of the entire plug. Accordingly, the shutdown dose rate around the IVVS/GDC plug is dominated by the GDC electrode head. It is therefore recommended to separate the GDC head from the system prior to further operations inside the Hot Cell. All components, except the Be protective layer of the GDC probe, were shown to be classifiable as low level radwaste according to the French regulations.

Within the EU Fusion Technology Programme dedicated theoretical and experimental efforts are conducted to provide reliable nuclear data and computational tools for design analyses for fusion devices like ITER including qualified uncertainty estimates. In this respect, the present paper reports on sensitivity and uncertainty analyses for the EU HCLL Test Blanket Module (TBM) of ITER. Neutron flux spectra and tritium production rates have been calculated using MCNP with a modified version of the ITER Alite torus sector model with integrated TBMs. Sensitivities of such parameters to nuclear cross sections of isotopes contained in the TBM as well as in the ITER device have been calculated using the Monte Carlo code MCSEN. Uncertainties could be obtained by using existing covariance data of the important nuclear cross section files, mainly from ENDF/B-VI, SCALE6.0, but also from recent JEFF/EFF evaluations. Like in the HCLL mock-up experiment two positions at front and back of the TBM have been selected. In both cases the calculated uncertainties of the responses (tritium production rate, neutron flux) are in the range of 2–4%.

There is increasing demand for fast and accurate neutronics analysis to be conducted on specific ITER components. The ATLAS (accelerated torus-like angular source) algorithm has been developed to accelerate such neutronics calculations by approximating the 3-D plasma volume source as a surface source at the first-wall region of interest. The ATLAS routine rapidly calculates the angular distribution of the unscattered neutron flux at the chosen first wall location, on a discrete angular grid, and writes a surface source for both MCNP and Attila. This discretised source can then be used to accurately reproduce the full 3-D toroidal volume source in angle and energy, and to quickly simulate the irradiation of individual components or regions in the ITER model. While designed for ITER, the method is general in implementation and can be applied to any tokamak.

Within the European Fusion Technology Programme dedicated computational methods, tools and data have been developed and validated for sensitivity and uncertainty analyses of fusion neutronics experiments. The present paper is devoted to this kind of analyses on the recent neutronics experiment on a mock-up of the Helium-Cooled Lithium Lead Test Blanket Module for ITER at the Frascati neutron generator. They comprise both probabilistic and deterministic methodologies for the assessment of uncertainties of nuclear responses due to nuclear data uncertainties and their sensitivities to the involved reaction cross-section data. We have used MCNP and MCSEN codes in the Monte Carlo approach and DORT and SUSD3D in the deterministic approach for transport and sensitivity calculations, respectively. In both cases JEFF-3.1 and FENDL-2.1 libraries for the transport data and mainly ENDF/B-VI.8 and SCALE6.0 libraries for the relevant covariance data have been used. With a few exceptions, the two different methodological approaches were shown to provide consistent results. A total nuclear data related uncertainty in the range of 1–2% (1σ confidence level) was assessed for the tritium production in the HCLL mock-up experiment.

In support of the breeder blanket development program, the EU is conducting a dedicated neutronics R&D effort to provide the basis for the design of nuclear tests to be performed in ITER on the Test Blanket Modules (TBMs). It includes the development of computational tools comprising both Monte-Carlo and deterministic transport, sensitivity and uncertainty codes, the generation of high quality neutron cross-section and covariance data libraries. These are validated experimentally in view of their application in the ITER TBM and the DEMO design. To this purpose, two neutronics experiments have been carried out on mock-ups of both the Helium Cooled Pebble Bed (HCPB) and the Helium Cooled Lithium Lead (HCLL) variants of ITER TBMs, at 14-MeV neutron sources. Redundant experimental techniques have been used to measure the resulting tritium production rate and the neutron and gamma ray spectra which are needed to predict the blanket shielding performance, nuclear power production and all nuclear loads. The comparison of experiment and corresponding calculation is obtained with the associated uncertainty margin based on experimental as well as calculational uncertainties. At the same time, suitable nuclear measuring techniques for TBMs in ITER, in particular for the tritium production, are being developed, optimised, and tested in the mock-up experiments. In particular, the present paper summarizes the preliminary results of the latest of such experiments, i.e. the one conducted on a mock-up of the HCLL blanket. It describes also the dedicated neutronics R&D activities, including the efforts to specify the neutronics tests and objectives in TBMs in ITER, considering the various design concepts.

This paper details progress in experimental characterisation work at JET for the long-term irradiation station, conducted as part of a project to perform activation experiments using ITER materials. The aim is to take advantage of the significant 14 MeV neutron yield expected during JET operations to irradiate samples of materials that will be used in the manufacturing of ITER tokamak components, such as Nb3Sn, SS316L steels from a range of manufacturers, SS304B, Alloy 660, W, CuCrZr, OF-Cu, XM-19, Al bronze, NbTi and EUROFER. This paper presents an assessment of the nuclear environment at the relevant irradiation locations at JET, measured using a range of high purity dosimetry foils: Ti, Ni, Y, Fe, Co, Sc, and Ta, irradiated with fusion neutrons at JET over a period of 15 months. Experimental results are presented and compared to simulation predictions using a JET MCNP model coupled with the FISPACT-II inventory code. Comparisons are made for a total of 11 nuclear reactions using a range of nuclear data libraries in calculations. © EURATOM 2018.

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

The International Fusion Materials Irradiation Facility (IFMIF) uses a high speed (10–20 m/s) lithium (Li)-jet flow as a target for two 40 MeV/125 mA deuteron beams. The major function of the Li target is to provide a stable Li jet for the production of an intense neutron flux. Different hydrodynamic instability mechanisms are possible in the Li-jet flow. The main objective of this study is to provide a detailed numerical analysis of instabilities in the Li-jet flow caused by separation of the flow near the nozzle outlet. Experimental observations have shown that the change of the nozzle geometry at the outlet or nozzle edge defects causes the flow separation and occurrence of longitudinal periodic structures on the free surface. Target surface fluctuations of large magnitude can lead to the penetration of the deuteron beam in the target structure and cause the local overheating of the back plate. This work is focused on the validation the CFD code Star-CD and choice of the suitable simulation technique for further calculations of lithium target flow. Two simulation methods, Large Eddy Simulation (LES) and Reynolds Averaged Navier–Stokes method using the Shear Stress Transport (SST) turbulence model have been validated against the experimental data. Two experimental cases with two different kinds of instability sources were used to test the suitability of turbulence models to predict waves generated near the nozzle edge. While in the first case the instability source is a stationary located obstacle at the nozzle edge, in the second one the position of the instability sources is influenced from turbulent fluctuations in the boundary layer near the nozzle edge. LES has a reasonable agreement with experimental data. SST turbulence model is not able to predict the local flow separations at the nozzle edge caused by turbulence fluctuations and to reproduce instabilities of small magnitudes.

The International Fusion Materials Irradiation Facility (IFMIF) is devised to contribute experimental evidence to an irradiated material properties database for candidate materials exposed to irradiation spectra and doses relevant for future fusion power reactors. Due to neutron fluxes generated by high-energy deuterons reacting in a liquid lithium target, damage rates of 20–50 displacements per atom in one full power year can be achieved in steel specimens inside a volume of approximately 0.5 L. The design of the high flux test module developed at the Forschungszentrum Karslruhe (FZK) allows for maximizing the space available in the high flux neutron field for material irradiation, while at the same time allowing precise adherence of the irradiation temperature of the specimen stacks. Since enhancement of the neutron irradiation requires placement of the specimens as close as possible to the neutron source, the design proposes thin container structures (obeying mechanical constraints) and flat coolant channels between the rigs. A helium gas flow is designated to remove the heat from the rigs to keep the required irradiation temperature, which may be chosen between 250 and 650 °C. As a result of the thin container walls and the small channel dimensions, the helium cooling gas flow is characterized by low pressure, transitional Reynolds numbers and intermediate Mach numbers. Dedicated experimental investigations on such minichannel cooling gas flows have been conducted with the ITHEX helium loop facility. Results obtained by laser Doppler anemometry indicate a complex three-dimensional evolution of the transitional laminar-turbulent flow field in the hydraulic entrance section. In the short cooling channels, a relevant portion of the flow alongside the rigs is influenced by this developing region. Detailed knowledge of the flow development and the resulting heat transfer coefficients is necessary to optimize the flow channel inlet design and to avoid inhomogeneities of the temperature field inside the specimen stacks, which otherwise could be caused by varying local heat transfer coefficients and mass flux redistributions (in the axial and the lateral coordinates). Experimental results are presented. Specific features of the minichannel entrance flow are identified, and conclusions are drawn for an optimized design of the entrance geometry, to enhance the utility of the HFTM for fusion materials research.

A mock-up of the European Helium-Cooled Pebble Bed TBM was irradiated with DT neutrons in pulsed and continuous mode at the Fusion Neutronics Laboratory of the University of Technology Dresden. The aim was to measure fast neutron and gamma-ray flux spectra as well as time-of-arrival spectra of the slow neutron flux. The results of the experiments were analysed by the Monte Carlo code MCNP and nuclear data from the European Fusion File (EFF-3), and the Fusion Evaluated Nuclear Data Library (FENDL-2.0/2.1). It was found that the calculation of the fast neutron flux above 3 MeV tends to overestimate while the gamma-ray flux and slow neutron flux in two measurement positions in the mock-up was underestimated. The mock-up was also irradiated at FNG/ENEA Frascati to measure tritium breeding rates by means of small Li 2 CO 3 pellet detectors inserted into the breeding layers. The breeding experiment was analysed at FZ Karl-sruhe with emphasis on determining sensitivities of the TPR to relevant cross section uncertainties of all materials in the mock-up. It was found that the TPR calculation shows a tendency to underestimate. From the sensitivity analysis it was found that the total TPR is most sensitive to the elastic scattering in Be and the 7 Li(n,α T) reaction.

An inspection procedure to assess the mechanical integrity of the International Fusion Materials Irradiation Facility (IFMIF) capsules and rigs during the irradiation campaign is necessary. Due to its penetration ability and contrast mechanism, the X-ray microtomography is the only known tool that could meet these requirements. In the high flux test module (HFTM) of IFMIF miniaturized specimens are densely packed in capsules. The capsules, which wear electric heaters and thermocouples, are housed in rigs. To assure a well-defined thermal contact the heater wires have to be attached to the capsules by brazing them into grooves. The examination of the quality of the braze material layer is of crucial interest in order to assure the best heat coupling of the heater wires to the capsule. A high density of the heaters is necessary to maintain the required temperature and, in addition NaK filling of narrow channels is employed for improving the 3D-heat transfer between the irradiation specimens and the capsule wall. Fully 3D tomographic inspections of a prototypical HFTM instrumented capsule, developed and manufactured at FZK, were conducted. In order to identify the optimum irradiation parameters and scanning configuration we carried out a comparative NDT analysis on two microtomography facilities: a compact, high magnification installation at NILPRP and a high-end industrial tomography facility with higher X-ray energy and intensity at HWM. At optimum inspection parameters of a directional microfocus X-ray source ( U = 220 kV and I = 300 μA) the geometry resolution was about 30 microns for characteristic dimension of the sample of 50 mm. Voids of 30 microns diameter and cracks of about 20 microns width can be detected. The absolute error of geometrical measurements is sufficient for the assessment of the structural integrity of the irradiation capsule and for the geometry description within the thermal-hydraulic modeling. The space resolution and the overall reconstruction quality could be further improved by operating the X-ray source at low intensity and, consequently, very small focus size and, in addition, by implementing procedures to compensate geometry and beam hardening artifacts.

The present paper summarizes work devoted to design and design validation of the HFTM and improvements obtained by neutronics, thermohydraulic and mechanical analyses. The HFTM design requirements and its main features will be described. Remarkable progress has been achieved in the development of appropriate computational tools, data and methods for IFMIF neutronics and activation analysis. Similarly, the thermohydraulic simulations have been validated against experimental data obtained in the ITHEX facility to test turbulence models of gas flow in narrow channels. Iterative analyses lead to several design optimizations of container and helium supply. Crucial manufacturing techniques for the parts of the HFTM have been tested extensively to assure a reliable and reproducible manufacturing route for the final procurement of the HFTM. Finally, the integrated design of the Target and Test Cell has been initiated with contributions from several international design teams.

The engineering process of the IFMIF high flux test module (HFTM) demands concurrent optimization with respect to neutronic, mechanical and thermal-hydraulic issues. The cooling technology of the HFTM makes use of helium gas flowing in mini-channels along the specimen capsules. The coolant gas flows in these mini-channels are characterized by transitional Reynolds numbers, high velocities and high heating rates. Further, the flow in the short channels is influenced by entrance effects on a relevant part of the full length. The present study addresses these issues by experimental and numerical simulations. To this end, the experimental facility ITHEX has been constructed, and several mini-channel test sections were investigated using a variety of experimental techniques. The numerical methods are validated against the experimental data, and general conclusions on the HFTM thermal-hydraulic layout are presented.