Experiments have been performed to qualify the radiation shielding performance of concrete for the test cell of the IFMIF-DONES (International Fusion Materials Irradiation Facility-Demo Oriented NEutron Source) facility. A mock-up of ordinary concrete (OC) with local lime-dolomite aggregate was prepared, representing the structural concrete. Also, a mock-up of heavy concrete (HC) with magnetite aggregate was prepared, representing the radiation shielding concrete, particularly for removable biological shielding blocks inside the test cell. These mock-ups were irradiated by neutrons with a continuous energy spectrum up to 33 MeV using a fast neutron generator at the U-120M cyclotron facility at the Nuclear Physics Institute of the Czech Academy of Sciences, & Rcaron;e & zcaron;. Activation foils were used for neutron diagnostics and were positioned throughout the mock-ups. There were five foil materials used: Au, In, Fe, Ti and Al. Gamma spectrometry of the foils was performed at the experimental site to measure the activities of the post-irradiation radionuclides. The experimental results are compared to calculated results obtained via radiation transport code modelling of the experiments and subsequent inventory code simulations of transmutation and activation. The comparison of calculated/experimental (C/E) results determines the confidence in the calculation methods to accurately represent the shielding performance of the concrete materials. Within the uncertainties, 100% of the OC results and 73% of the HC showed C/E agreement. Future calculations using these concretes should include safety factors of 2 and 3 for OC and HC respectively. The discrepancies for HC in particular should be considered in future calculations.

The Fusion Evaluated Nuclear Data Library (FENDL) is a comprehensive and validated collection of nuclear cross section data coordinated by the International Atomic Energy Agency (IAEA) Nuclear Data Section (NDS). FENDL assembles the best nuclear data for fusion applications selected from available nuclear data libraries and has been under development for decades. FENDL contains sub-libraries for incident neutron, proton, and deuteron cross sections including general purpose and activation files used for particle transport and nuclide inventory calculations. We describe the history, selection of evaluations for the various sub-libraries (neutron, proton, deuteron) with the focus on transport and reactor dosimetry applications, the processing of the nuclear data for application codes, and the development of the TENDL-2017 library which is the currently recommended activation library for FENDL. We briefly describe the IAEA IRDFF library as the recommended library for dosimetry fusion applications. We also present work on validation of the neutron sub-library using a variety of fusion relevant computational and experimental benchmarks. A variety of cross section libraries are used for the validation work including FENDL-2.1, FENDL-3.1d, FENDL-3.2, ENDF/B-VIII.0, and JEFF-3.2 with the emphasis on the FENDL libraries. The results of the experimental validation showed that the performance of FENDL-3.2b is at least as good and in most cases better than FENDL-2.1. Future work will consider improved evaluations developed by the International Nuclear Data Evaluation Network (INDEN). Additional work will be needed to investigate differences in gas production in structural materials. Covariance matrices need to be updated to support the development of fusion technology. Additional validation work for high-energy neutrons, protons and deuterons, and the activation library will be needed.

In 2021 JET exploited its unique capabilities to operate with T and D-T fuel with an ITER-like Be/W wall (JET-ILW). This second major JET D-T campaign (DTE2), after DTE1 in 1997, represented the culmination of a series of JET enhancements-new fusion diagnostics, new T injection capabilities, refurbishment of the T plant, increased auxiliary heating, in-vessel calibration of 14 MeV neutron yield monitors-as well as significant advances in plasma theory and modelling in the fusion community. DTE2 was complemented by a sequence of isotope physics campaigns encompassing operation in pure tritium at high T-NBI power. Carefully conducted for safe operation with tritium, the new T and D-T experiments used 1 kg of T (vs 100 g in DTE1), yielding the most fusion reactor relevant D-T plasmas to date and expanding our understanding of isotopes and D-T mixture physics. Furthermore, since the JET T and DTE2 campaigns occurred almost 25 years after the last major D-T tokamak experiment, it was also a strategic goal of the European fusion programme to refresh operational experience of a nuclear tokamak to prepare staff for ITER operation. The key physics results of the JET T and DTE2 experiments, carried out within the EUROfusion JET1 work package, are reported in this paper. Progress in the technological exploitation of JET D-T operations, development and validation of nuclear codes, neutronic tools and techniques for ITER operations carried out by EUROfusion (started within the Horizon 2020 Framework Programme and continuing under the Horizon Europe FP) are reported in (Litaudon et al Nucl. Fusion accepted), while JET experience on T and D-T operations is presented in (King et al Nucl. Fusion submitted).

Recent advancements in the neutronics activities of the IFMIF-DONES project, developed within the EUROfusion framework, are presented. These include updates to radiation dose maps during commissioning and normal operation of the accelerator systems; material irradiation analyses and shielding optimization of the test systems; activation inventories of 7Be, 3H, and activated corrosion products in the Li systems; shutdown dose analyses of transportation and storage of radioactive waste, cooling water and atmosphere gas activations, skyshine to the public, etc. The development of simulation tools, nuclear data evaluation, and nuclear experiments for the specific needs of DONES neutronics are highlighted, as well as the nuclear analysis handbook and database. Several challenges for future development are also discussed to ensure the provision of high-quality nuclear analyses.

As the operational period of the Joint European Torus (JET) nears its scheduled conclusion in 2023, accurate predictions of the nuclear waste inventory and end-of-life shutdown dose rates (SDRs) based on actual operations are required for the transition to the repurposing and decommissioning of the reactor. The novel nuclear waste analysis method based on the Rigorous Two-Step (R2S), Cell-under-Voxel (CuV) approach used in the newly updated Mesh Coupled R2S (MCR2S) code has been employed to give precise predictions for the waste inventory and SDRs in JET. From this analysis, SDRs from the reactor and individual in-vessel components after extraction were obtained to guide waste management considerations. Also, crucially, the CuV method increased the precision of waste inventory predictions, allowing for a more rigorous decommissioning of the machine.

Future fusion reactors using deuterium-tritium fuel will exhibit high fluences of high-energy neutrons inside and around the reactor vacuum vessel (VV). As well as causing material damage, fusion neutrons will activate materials, the decay of which leads to radiation fields in and around the reactor after shutdown. Gamma -ray emission from activated materials is a particular radiological hazard during periods of reactor shutdown. This must be accounted for in the design of the reactor shielding to ensure that risks are reduced as low as reasonably achievable. Recent neutronics work has evaluated the shutdown dose rates (SDDRs) in the EU DEMOnstration power plant (DEMO) around the ports and throughout the cryostat, incorporating prospective shielding improvements to the VV and ports. Prior to the proposed shielding design improvements, calculations for the model including the helium-cooled pebble bed (HCPB) blanket showed that radiation leakage through the blanket and VV leads to biological-equivalent SDDRs (following 12 days' decay) above 103 & mu;Sv/h throughout the cryostat, ignoring additional contribution from radiation streaming through the port openings. Inclusion of the proposed VV changes reduces this dose rate to below 100 & mu;Sv/h. The work finds an approximate order-of-magnitude reduction in SDDR throughout the cryostat when all proposed shielding improvements are applied, leading to dose rates in the cryostat in the range of hundreds to thousands of & mu;Sv/h for the full model. The work shows that to further reduce dose rates inside the cryostat, improving the shielding performance of the ports is required, with particular emphasis on the lower port and the equatorial electron-cyclotron launcher which currently dominate the dose rates.

Global research programs seeking to achieve a commercially viable model of a fusion power plant are being accelerated at an unprecedented rate. One critical element to the design and licensing is an accurate understanding of the radiation environment throughout the plant lifetime and subsequent decommissioning phase. The radiation field, which results from the nuclear fusion reaction, gives rise to highly complex phenomena such as flux leakage, materials activation, and decay gamma fields. Demonstration of compliance with limits, the integrity of components, and the permissibility of operations are all fundamental to regulatory approval and the overall safety of a nuclear device. As such, neutronics, which is used in the general sense to refer to the mapping of radiation fields in nuclear devices, is a critical design driver. The Applied Radiation Technology group at the United Kingdom Atomic Energy Authority is a world leader in this field, developing new methods and deploying state-of-the-art codes to conduct nuclear analysis. As well as applied neutronics in areas spanning fusion reactors, medical applications, spallation neutron sources, and nuclear fission, there is an extensive parallel experimental program undertaking critical radiation field characterization and conducting measurements using an array of bespoke particle detection systems. This paper highlights recent technical developments made by this group in the context of outstanding challenges in this field, as well as providing an overview of current methods and capabilities for the broader interest of the community.

General Fusion is developing a Magnetized Target Fusion approach that involves compressing an initial magnetically confined plasma inside a cavity formed in liquid metal. General Fusion has launched a Fusion Demonstration Program to design and construct a large scale, integrated prototype, being built at the Culham Centre for Fusion Energy in the United Kingdom. It is proposed to implement a highly segmented, time-of-flight Neutron Emission Spectrometer to measure the energy distribution of neutrons emitted nominally at 2.45 MeV from deuterium-deuterium fusion. This distribution will be used to extract the ion plasma temperature at the late stages of compression. An incident neutron flux of up to one billion neutrons per second is expected at the spectrometer, resulting in very demanding performance criteria. It is vital that the Fusion Demonstration Program’s prototype machine have a diagnostic system capable of measuring plasma ion temperatures of 10 keV (116 million degrees Celsius) to evaluate the success of a compression and the overall performance. Simulations modeling plasma compression, neutron interactions with the prototype’s apparatus and spectrometer, as well as the development of a prototype time-of-flight spectrometer in partnership with TRIUMF and Simon Fraser University, will be discussed.

This work presents neutron flux calculation results and specific decay heat and activity for the European DEMO reactor divertor structure. Two DEMO 2017 models were used in calculations; one with a homogenised WCLL breeder blanket and the other with a homogenised HCPB breeder blanket. The neutron flux calculations were performed using MCNP6 code with JEFF 3.2 nuclear data. Activation and decay heat calculations were performed using the FISPACT-II code with TENDL - 2017 nuclear data library.

A new method for the calculation of Shutdown Dose Rates (SDDR) has been developed, the Novel-1-Step (N1S) method. The new method retains the benefits of only requiring a single radiation transport calculation, as in the use of the direct-1-step (D1S) method, while removing the need for pre-calculations to determine dominant nuclides and time correction factors. The N1S method uses a time dependent source and decay data for all nuclides. When reactions in the transport occur leading to unstable daughter nuclide, the correct contribution of photon radiation from all the decay products of a nuclide are calculated with no need for additional external activation calculations. Weights of these decay photons are calculated for each decay time of interest and are analytically determined based on the solutions to the Bateman equations. The N1S method has been implemented into MCNP and preliminary verification calculations performed. These calculations included the FNG ITER shutdown dose rate benchmark and the ITER SDDR cross comparison. For the FNG ITER SDDR benchmark the N1S method showed good agreement, within experimental error, for the first campaign apart from the first decay time where a C/E value of 1.34 was obtained. This overestimation was shown to be due to the decay of 64Cu inside the copper cup of the neutron generator. For the second campaign the N1S method showed an under prediction of up to 20% at short decay times and an over prediction up to 20% at longer decay times. These times are dominated by 56Mn and 58Co respectively and it is likely the difference is due to under and over predictions in the reaction rates leading to these nuclides. The ITER cross comparison showed good agreement between the N1S method and MCR2S (and by association other D1S and R2S codes). Differences seen in the results were shown to be due to difference in the calculated reaction rates using EAF2010, TENDL2019 and FENDL3.2.

The DEMOnstration fusion power plant (DEMO) is being developed within EUROfusion and one of the challenges is the integration of all the systems into a fusion reactor by designs that meet strict design criteria required for safe and reliable long-term reactor operation. Neutronics analyses are required as an important contribution to this effort. The work described here has analysed the equatorial port plugs of the electron cyclotron (EC) system in terms of neutronic aspects. The EC port plugs are needed for heating the fusion plasma and for plasma control. Necessary openings in the breeding blanket and port plug structures foreseen for the mm-wave heating beams of microwaves coming from launchers of the EC system in equatorial port plugs are a challenge in terms of neutron shielding. Both the neutron streaming through the EC port and nuclear loads in its critical components like mirrors have to be considered in the system design and integration.

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

The accurate modelling of the activation of flowing material in a fusion reactor, such as coolant water or lithiumlead breeder, has important safety and shielding implications. Two codes developed at CCFE which account for neutron flux variation have been investigated for the effects of incorporating computational fluid dynamics (CFD) and benchmarked against experimental data. With the inclusion of CFD, both codes are found to be reasonably accurate and benchmarking discrepancies identified previously for 16N water activation data have been clarified. Precise paths calculated using CFD have been used in flowing lithium-lead activation analysis for the first time, with results suggesting that simplified linear paths may give comparable results to detailed CFD paths, but low-detail CFD paths should be avoided. This work paves the way for an accurate and benchmarked set of fluid activation codes.

Steady state magnetic sensors based on the Hall effect are one of the candidate diagnostic concepts for measurement of plasma current, plasma position and shape on the future demonstration fusion power reactor (DEMO). Design of any diagnostic component has to consider the foreseen high neutron fluences which will be accumulated over DEMO's operational time. The total DEMO life time neutron fluence is expected to reach up to 1026 m-2 for in-vessel locations. At these levels of neutron load, transmutation of materials becomes an effect which has to be considered. Neutron spectra and flux rates were determined using MCNP6 code for a selected location considered to be instrumented by the Hall sensors on DEMO. Using these neutron spectra, inventory simulations were conducted with FISPACT-II code for a set of materials presently considered for sensitive layers of the DEMO Hall sensors. Resulting information on evolution of material composition of Hall detectors along the DEMO operation is summarized. Comparison of simulations of various irradiation time patterns e.g. continuous versus more realistic DEMO campaign regime is presented.

The way to arrive at a licensing phase for a nuclear fusion installation is not straightforward mainly because of the lack of operating experience and of dedicated nuclear regulations. In fact, only small/medium experimental facilities exist with limited licensing processes and only one large experiment, ITER, has obtained a construction license. Therefore, the safety assessment and the preparation of the preliminary safety report is almost a first of a kind for DEMO. Taking advantage of the fission power plants experience and considering to the maximum extent the ITER safety studies, the preparation of a Generic Site Safety Report (GSSR) has begun. It will require some years to be completed; however currently, at the starting point, the strategy to develop it is clear and well defined. This paper considers all the safety issues that will be included in the GSSR because they have been clarified in the frame of the European Workprogramme for DEMO from 2014 up to 2018, and should not be modified in the future, such as the safety requirements for the plant and the systems, the tools to be used for the safety assessment, the procedures for the selection of the reference accidents, and so on. Together with these topics, considered as goals achieved, there are others for which an additional effort is necessary because they do not cover all the expected requirements of the likely licensing procedures applicable for DEMO. A complete spectrum of Design Basis Accidents and Beyond Design Basis Accidents that can determine the risk of releases from the main systems of the power plant is still incomplete together with the safety classification of most of the Structures, Systems and Components, and the feasibility and analyses of some accident mitigation systems. The outcome of this study is the quantification, when possible, of the gap between the results achieved and the goals established in the power plant guidelines. It will help also to qualify the effort required in terms of studies, experiments and human resources to reach a good stage for successful DEMO licensing.

Shutdown dose rate calculations provide an essential input to the design and research of fusion power plant technology. They allow the estimation of dose to personnel and equipment during planned and unplanned maintenance. The mesh coupled rigorous 2 step (MCR2S) methodology used at Culham Centre for Fusion Energy (CCFE) was originally developed to link the MCNP particle transport code and the FISPACT-II inventory code. As new particle transport codes are developed there is a strong motivation to move towards a code agnostic approach. This paper details the integration of MCR2S with the FISPACT-II API and two other transport codes, Serpent 2 and OpenMC. Two benchmarks, the FNG shutdown dose rate experimental benchmark and the ITER computational benchmark, have been performed and compared to results produced with MCNP. In general, the results show that MCNP, Serpent 2 and OpenMC give shutdown dose rate results similar to the FNG experiment for both experimental campaigns. However, all codes appeared to slightly overestimate the dose rates for Campaign 1 (all results had a C/E between 1 and 1.5) and underestimate the dose rates for Campaign 2 (all results had a C/E between 0.6 and 1). Differences were seen between OpenMC and MCNP for the ITER port plug benchmark, where the lack of variance reduction in OpenMC meant that the neutron flux estimates at the rear of the model were not converged. This led to differences of up to 13% in the shutdown dose rates. It was shown that Serpent 2 and MCNP, where variance reduction was used, gave shutdown dose rates within 3% of each other. Although some areas for development of the Serpent 2 and OpenMC transport codes have been highlighted, overall the comparisons give confidence that the implementation of these two transport codes into the MCR2S work-flow has been carried out successfully.

Within the EUROfusion Power Plant Physics and Technology Department the DEMOnstrational fusion power plant (DEMO) is being developed. One of the fundamental challenges is the integration of ports in the vacuum vessel. The lower port of the DEMO machine is particularly challenging due to tight space constraints imposed by the toroidal field (TF) coils and the requirement to provide a large open duct through both the divertor and inside the port to enable for vacuum pumping. In addition, feeding pipes of divertor and tritium breeding blanket need to be integrated and access space must be provided for various remote handling operations. p Several neutronics requirements need to be fulfilled, e.g. the nuclear heating of the superconducting TF coils and the gamma radiation levels inside the cryostat need to be limited to reduce occupational exposure to per-sonnel during maintenance, and the irradiation damage and neutron heating in different components need to be considered in the design and limited. The results of neutronic analyses show that further shielding optimization is needed as maximum TF coil heating is still 5x the design limit and the SDDR values orders of magnitude above the target values inside the lower port duct. With this in mind the direction of future design developments is discussed.

The DEMOnstration fusion power plant (DEMO) is being developed within the EUROfusion Power Plant Physics and Technology Department. Challenging aspects of this work include the integration of all required systems necessary for reactor operation, to give access during maintenance operations, while providing sufficient shielding from neutrons and gammas. The Upper port of DEMO is challenging due to tight space constraints imposed by the toroidal field (TF) coils, a large port opening required for the tritium breeding blanket (BB) replacement sequence, and the need for various systems being integrated or passing through the port. In the present case, which is considered the most challenging, the port size and configuration are dominated by the space requirements to allow for the extraction of the BB segments and to integrate the BB coolant pipes, which are particularly large in case of helium as a coolant. In addition a plasma limiter plug is integrated inside the port. At the same time the TF coils on the exterior of the port need to be effectively shielded from neutrons to prevent loss of superconductivity due to nuclear heating and the shutdown dose rate (SDDR) in the cryostat needs to be within the limits. The evolution of the design and its effect on the neutronic performance are presented and the plans for future work discussed.

The plasma diagnostic and control (D&C) system for a future tokamak demonstration fusion reactor (DEMO) will have to provide reliable operation near technical and physics limits, while its front-end components will be subject to strong adverse effects within the nuclear and high temperature plasma environment. The ongoing developments for the ITER D&C system represent an important starting point for progressing towards DEMO. Requirements for detailed exploration of physics are however pushing the ITER diagnostic design towards using sophisticated methods and aiming for large spatial coverage and high signal intensities, so that many front-end components have to be mounted in forward positions. In many cases this results in a rapid aging of diagnostic components, so that additional measures like protection shutters, plasma based mirror cleaning or modular approaches for frequent maintenance and exchange are being developed. Under the even stronger fluences of plasma particles, neutron/gamma and radiation loads on DEMO, durable and reliable signals for plasma control can only be obtained by selecting diagnostic methods with regard to their robustness, and retracting vulnerable front-end components into protected locations. Based on this approach, an initial DEMO D&C concept is presented, which covers all major control issues by signals to be derived from at least two different diagnostic methods (risk mitigation).

Waste-production predictions for the future demonstration fusion power plant (DEMO) are necessary to produce an accurate picture of the likely environmental and economic costs of radioactive waste disposal at end-of-life. An integrated simulation process combining MonteCarlo neutron transport simulations, inventory calculations, and extensive and reproducible post-processing algorithms has been used for the evolving European DEMO designs to quantify the time-varying mass inventories in different waste classes for individual regions and components of the reactor vessel, as well as for the reactor as a whole. Waste categories based on UK and French regulations reveal that minor impurities contained in structural steels, particularly Eurofer, as well as in functional materials such as tungsten, and beryllium, can have a significant impact on their waste classification prospects. Predictions for current European DEMO concepts suggest that there may be an issue in disposing of fusion structural-steel waste as low-level waste (LLW) in near-surface repositories. Detailed analysis of the subtleties of these predictions, particularly with regard to the production of long-lived radionuclides such as C-14 and Nb-94, reveal that the threshold for acceptance as LLW is only just exceeded in some situations. Several mitigation approaches are discussed in this context. The computational framework developed for these assessments can be rapidly and continuously applied to the maturing DEMO design, helping to guide design choices to mitigate long-lived waste production and ensure that most waste becomes LLW or better within a few decades.