In water-cooled D-T fusion tokamaks such as the Joint European Torus (JET), neutron activation of flowing water produces short-lived radionuclides that generate a distributed high-energy gamma and neutron radiation source in the water cooling circuit. The energy and intensity of the source are crucial factors in the design of cooling and shielding systems in future water-cooled deuterium-tritium fusion tokamaks such as ITER. While several past experiments provide fusion-relevant experimental data on water activation, the JET water activation experiment carried out during the 2023 DD and DT campaigns is the first such experiment in a fusion tokamak environment. In the water activation experiment at JET, two types of scintillation detectors were used to measure gamma rays from activated water in the basement below JET Octant 4. The system successfully recorded and processed data from over 1500 JET pulses, including both DD and DT operations, providing a robust data set for analyzing N-16 activity in cooling water. This manuscript describes the preparation of the JET water activation experiment, the calibration of the detectors and the analysis of the experimental data.

The development of analysis tools to calculate the activation of flowing water under irradiation is essential for fusion technology and for the ITER project. The Radio-Species Transport Model (RSTM) method is a simulation methodology developed by Fusion for Energy based on the Ansys Fluent (R) user-defined scalar (UDS) approach. It predicts the activation of a flowing fluid in domains where neutron fields and flow regimes require the coupling of activation and fluid dynamic effects. RSTM was successfully applied to ITER First Wall (FW) studies and benchmarked against experiments at the Frascati Neutron Generator (FNG). This research focuses on the application of the RSTM to the KATANA closed water activation loop at the JSI TRIGA Mark II fission reactor, as part of the EUROfusion Preparation of ITER Operation (PrIO) programme. Future work will compare these results with experimental data and other predictive tools including ActiFlow, GammaFlow (by UKAEA) and FLUNED (by UNED). In the next phase of operation, the KATANA facility aims to tackle experiments more relevant to ITER conditions. For this reason, Fusion for Energy is applying RSTM to design an alternative irradiation head that resembles the cooling circuit of an ITER First Wall panel, this being one of the main components where the water undergoes activation. This study shows the results of the RSTM tool for the current configuration of the JSI KATANA water activation loop. Conclusions are drawn on the relevance of the alternative ITER-relevant irradiation head for the next phase of the JSI water activation loop.

This paper provides an overview on the engineering analyses workflow typically executed in support of the assembly and operation of superconducting magnet systems, in particular for ITER. Accurate field computations are required to evaluate (1) Lorentz forces as primary loads on magnets (2) cable temperature margins (3) critical error field harmonics and (4) field lines tracing. Although the required accuracy increases with each of these magnetic computations, their common aim, beyond use during machine design, is to provide estimates on the final alignment of magnetic fields and, possibly, to guide and adjust magnet installation. Two global mechanical models (18 TF and 2 TF coils) featuring their interfaces to the Pre-Compression Rings, CS and PF coils are used to assess the impact of tolerances and misalignments on the final coils' positions in operation so to guide, with the aid of metrology data, the assembly process. With these, we also evaluate the mechanical stresses during current tests at 4 K if performed before final installation. A global and several local models of the Pre-Compression Rings that must be mounted in the tokamak were developed and used to design the tightening sequence during PCRs assembly and the required assembly tooling. Magnet systems also require sophisticated models to predict conductor temperature margins and to assist during thermal transients (magnets cooldown). Thermal loads due to Joule losses in cold structures are evaluated with 3D eddy-current codes and nuclear heating with detailed Monte-Carlo models suited to simulate streaming and deep penetration 3D phenomena leading to nuclear power deposition in the cables.

The ITER blanket system safeguards the Tokamak's Vacuum Vessel and ex-vessel components from thermal and nuclear loads induced by the plasma. It comprises two parts: the First Wall panel facing the plasma and the Shield Block that provides the bulk of the shielding. The First Wall panels consist of Beryllium tiles on a CuCrZr layer, supported by stainless-steel structures which are cooled using a pressurized water circuit at 40 bar with an inlet temperature of 70 °C.Fusion for Energy has updated the thermal and mechanical analyses of ITER First Wall Panel 11 in preparation for the serial manufacturing. The panel will face heat flux up to 2 MW/m2 from plasma radiation, in addition to the volumetric nuclear heating, and the electromagnetic loads. A model with an enhanced detailed 3D geometry and heterogeneous materials indicates a 25% reduction of the deposited heating compared to the previous 2D model, from 625.4 kW (2D) to 455.4 kW (3D).The study embraces numerous thermal and mechanical analyses using Ansys software v19.2, incorporating 2D and 3D nuclear heating data. Responses are then compared and thus the possible impact of the two distributions on the thermal and mechanical behavior of the Fist Wall Panel 11 is assessed.

Photoneutrons may be generated in beryllium by energetic gamma rays via the reaction 9Be(gamma,n)8Be. In ITER, the beryllium layer of the first wall may be the source of such photoneutrons. During plasma operation, these are of insignificant intensity compared with D-T neutrons from the plasma, but after shutdown, photoneutrons produced by decay gammas from neutron-activated material may be significant enough to impact sensitive electronic components in diagnostic or remote handling equipment that would not otherwise be exposed to neutrons.Studies have been performed to characterize the expected photoneutron source and to evaluate the fluxes arising in detailed three-dimensional models of the ITER tokamak. The results show photoneutron fluxes approaching 105 n/cm2 center dot s within the vessel and up to 103 n/cm2 center dot s elsewhere within the bioshield 14 days after shutdown. When first-wall panels are being transported to the Hot Cell Facility after irradiation, a photoneutron flux exceeding 104 n/cm2 center dot s within the transfer cask is predicted 21 days after shutdown. The peak values in the surrounding building are between 102 and 103 n/cm2 center dot s at the same time.

ITER is of key importance in the European fusion roadmap as it aims to prove the scientific and technological feasibility of fusion as a future energy source. The EUROfusion consortium of labs within Europe is contributing to the preparation of ITER scientific exploitation and operation and aspires to exploit ITER outcomes in view of DEMO. The paper provides an overview of the major progress obtained recently, carried out in the frame of the new (initiated in 2021) EUROfusion work-package called 'Preparation of ITER Operation' (PrIO). The overview paper is directly supported by the eleven EUROfusion PrIO contributions given at the 29th Fusion Energy Conference (16-21 October 2023) London, UK [www.iaea.org/events/fec2023]. The paper covers the following topics: (i) development and validation of tools in support to ITER operation (plasma breakdown/burn-through with evolving plasma volume, new infra-red synthetic diagnostic for off-line analysis and wall monitoring using Artificial Intelligence techniques, synthetic diagnostics development, development and exploitation of multi-machine databases); (ii) R&D for the radio-frequency ITER neutral beam sources leading to long duration of negative deuterium/hydrogen ions current extraction at ELISE and participation in the neutral beam test facility with progress on the ITER source SPIDER, and, the commissioning of the 1 MV high voltage accelerator (MITICA) with lessons learned for ITER; (iii) validation of neutronic tools for ITER nuclear operation following the second JET deuterium-tritium experimental campaigns carried out in 2021 and in 2023 (neutron streaming and shutdown dose rate calculation, water activation and activated corrosion products with advanced fluid dynamic simulation; irradiation of several materials under 14.1 MeV neutron flux etc).

The development of nuclear fusion as a safe and virtually limitless power source is receiving growing attention in the context of looming energy crisis and climate change. ITER project stands as the flagship international initiative and is advancing steadily. The construction of the Tokamak Complex is nearly finished, and the assembly of core components has begun on site. Simultaneously, the design is being finalized, and the safety case is becoming more concrete. Current approaches to radiation safety demonstration using 3D nuclear analysis with the Monte Carlo code MCNP require sophisticated artifacts to sew together simulations in separate models for the Tokamak and the rest of the facility. This results in cumbersome studies and, consequently, challengeable conclusions. To address this issue, we have built the an integral MCNP model of the ITER facility: the ITER full model. Along with improvements to the D1SUNED code, we illustrate its computational practicality and pertinence in two meaningful simulations for ITER safety case. This work represents the culmination of a two-decade-long effort of ITER modelling aiming to demonstrate adequate radiation safety. Beyond supporting the remaining design tasks, this model simplifies the corresponding 3D nuclear analysis and improves the robustness of the ITER safety case. In this work, authors present an integral MCNP model of the ITER facility for radiation transport, addressing limitations of separate models for the Tokamak and Tokamak Complex. This model streamlines safety simulations, enhancing the robustness and simplicity of radiation safety analysis.

ITER is the flagship fusion project, conceived as an experiment to select and develop the technologies for the first demonstration reactor, DEMO. Nuclear analysis is a core discipline in support of the design, commissioning and operation of the machine. To date, it has been conducted with increasingly detailed partial models, which represented toroidal segments of the tokamak. However, the limitations of this methodology became evident as estimates of quantities relevant to design, safety and operation showed unquantifiable uncertainties, which is a risk. Here, we present a detailed and realistic 360° MCNP model of the ITER tokamak called E-lite. We demonstrate the model’s usability and practicality. Two examples are used to illustrate qualitatively and quantitatively how it solves previously intractable problems with marked benefits for the future nuclear analysis of ITER, with applications to DEMO and future reactors. E-lite constitutes a milestone in the field of nuclear analysis in terms of realism in the evaluation of key quantities. The development of nuclear fusion reactors requires detailed nuclear analyses to ensure they run safely and effectively. These are currently typically done using partial models. Juarez et al. now present a complete 360° model of the ITER tokamak, which allows for more realistic and complete modelling of the system.

The planned in situ maintenance tasks in the ITER port interspace are fundamental to ensure the operation of equipment to control, evaluate and optimize the plasma performance during the entire facility lifetime. They are subject to a limit of shutdown dose rates (SDDR) of 100 µSv h−1 after 106 s of cooling time, which is nowadays a design driver for the port plugs as well as the application of ALARA. Three conceptual shielding proposals outside the ITER ports are studied in this work to support the achievement of this objective. Considered one by one, they offer reductions ranging from 25% to 50%, which are rather significant. This paper shows that, by combining these shields, the SDDR as low as 57Δ µSv h−1 can be achieved with a local approach considering only radiation from one port (no cross-talk form neighboring ports). The locally evaluated SDDR are well below the limit which is an essential pre-requisite for achieving 100µSv h−1 in a global analysis including all contributions. Further studies will have to deal with a realistic port plug design and the cross-talks from neighbour ports.

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. (C) 2016 Elsevier B.V. All rights reserved.

This paper assesses the quality of the EAF-2007 and 2010 activation cross sections for relevant reactions in the determination of the Shutdown Dose Rate (SDDR) in the Port Cell (PC) and Port Interspace (PI) areas of ITER.For each of relevant ITER materials, dominant radionuclides responsible of SDDR and their production pathways are listed. This information comes from a review of the recent reports/papers about SDDR in ITER and own calculations. A total of 26 relevant pathways are found. The quality of these cross sections pathways is assessed following EAF validation procedure, and for those found as not validated last TENDL library versions have been investigated in order to check possible improvements when compared to EAF.The use of EAF libraries is found as trustworthy and it is recommended for the prediction of SDDR in the ITER PC and PI. However, 3 cross section reactions are considered for further improvement: Co59(n,2n)Co58, Cu63(n,g)Cu64 and Cr50(n,g)Cr51. (C) 2016 Elsevier B.V. All rights reserved.

Dust and tritium inventories in the vacuum vessel have upper limits in ITER that are set by nuclear safety requirements. Erosion, migration and re-deposition of wall material together with fuel co-deposition will be largely responsible for these inventories. The diagnostic suite required to monitor these processes, along with the set of the corresponding measurement requirements is currently under review given the recent decision by the ITER Organization to eliminate the first carbon/tungsten (C/W) divertor and begin operations with a full-W variant Pitts et al. [1]. This paper presents the result of this review as well as the status of the chosen diagnostics.

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. (C) 2014 Karlsruhe Institute of Technology. Published by Elsevier B.V. All rights reserved.

Nuclear analyses provide essential input to the conceptual design, optimisation, engineering and safety case of fusion technology in current experiments, ITER, next-step devices and power plant studies. Calculations are intricate and computer-intensive, typically requiring detailed geometry models, sophisticated acceleration algorithms, high-performance parallel computations, and coupling of large and complex transport and activation codes and databases. This paper reports progress on some key areas in the development of tools and methods to meet the specific needs of fusion nuclear analyses. In particular, advances in the production and modernisation of reference models, in the preparation and quality assurance of acceleration algorithms and coupling schemes, and in the evaluation and adaptation of alternative transport codes are presented. Emphasis is given to ITER-relevant activities, which are the main driver of advances in the field. Discussion is made of the importance of efforts in these and other areas, considering some of the more pressing needs and requirements. In some cases, they call for a more efficient and coordinated use of the scarce resources available. (C) 2013 Elsevier B.V. All rights reserved.

As part of CCFE's nuclear data and technology programmes a series of material irradiations have been performed at the ASP accelerator to determine integral reaction cross-sections for fusion relevant materials. The integral reaction cross section can be used as part of the validation of the cross section data. The ASP machine accelerates deuterons onto a tritiated target to produce approximately 14MeV neutrons via the D–T fusion reaction. These neutrons interact with the material creating radioactive isotopes. The gamma emissions from the activated products are then measured using a high resolution gamma spectroscopy system. An important part of evaluating the results of these and future experiments lies in an accurate determination of the neutron energy spectrum. Initially a neutron spectrum determined by MCNP modelling was used based on a source term calculated using relativistic kinematics. The work reported here improves the understanding of the neutron spectrum using a combination of enhanced modelling and experimental data as input information to be used in spectrum unfolding. Recent advances in simulation techniques allow us to use deuteron cross sections for low energy deuterons and hence model the production of neutrons by the deuterons explicitly. This means that it is possible to model the effect of changing various deuteron beam parameters such as radius, energy and position on the target to understand what influence these have on the spectrum seen at the material of interest. The spectrum unfolding, based on threshold reactions, has not previously been performed for the ASP accelerator and provides an experimental method to improve the understanding of the neutron spectrum in the irradiation position. The combination of both modelling and experimental work to improve the understanding of the neutron spectrum has led to better understanding of the facility and the influences on the neutron spectrum. The combination of modelling and experimental work performed in this work to characterise the ASP neutron spectrum may also be applicable to other current and future neutron irradiation facilities such as IFMIF where an accurate knowledge of the neutron spectrum in each irradiation location will be required.

This work presents new integral data measured at the ASP 14 MeV neutron irradiation facility at Aldermaston in the UK, which has recently become available for fusion-related work through the CCFE materials programme. Measurements of reaction products from activation experiments using elemental foils were carried out using gamma spectrometry in a high efficiency, high-purity germanium (HPGe) detector and associated digital signal processing hardware. Following irradiation and rapid extraction to the measurement cell, gamma emissions were acquired with both energy and time bins. Integral cross section and half-life data have been derived from these measurements. Selected integral cross section values are presented from the measurement campaigns.

A study is reported which computes the radiation transport and activation response throughout the ITER machine and updates the ITER radioactive waste assessment using modern 3D models and up-to-date methods. The latest information on component design, maintenance, replacement schedules and materials is adopted. The radwaste classification is revised for all the major components of ITER, as well as several representative port plugs. Results include categorisation snapshots at different decay times, time histories of radiological quantities throughout the machine, and guidelines on interim decay times for components. All plasma-facing materials except tungsten are found to classify as type B due to the transmutation of their main constituents. Major contributors to the IRAS index of all materials are reported. Elemental concentration limits for type A classification of first wall and divertor materials are obtained; for the steels, only a reduction in service lifetime can reduce the waste class. Comparison of total waste amounts with earlier assessments is limited by the fact that analyses of some components are still preliminary; the trend, however, indicates a potential reduction in the total amount of waste if component segregation is demonstrated. Crown Copyright (C) 2012 Published by Elsevier B.V. All rights reserved.

In support of the technological requirements for fusion, the UK fusion technology programme is conducting several experiments using the ASP DT neutron irradiation facility at Aldermaston. The present experimental programme covers two key areas of technology development: improving the quality of nuclear cross-section data required for fusion-related materials; and benchmarking fusion nuclear analysis tools such as the shutdown activation dose system MC-R2S. The first of these areas is the focus of this paper.The work presented here gives a technical overview of the ASP facility, including the accelerator, the irradiation cell, the recently re-commissioned fast sample extraction system and ancillary radiation metrology equipment. Results from preliminary activation experiments using high purity elemental- and fusion-relevant material foils, are presented. These include measurements of short-lived activation products conducted via gamma spectrometry methods. (C) 2012 Elsevier B.V. All rights reserved.