Shutdown dose rate (SDDR) assessments have been performed for the DEMO tokamak model, including the latest design and environmental configurations. The main objective of this study was to evaluate the shutdown radiation fields and establish dose rate limits to ensure safe personnel access to the Vacuum Vessel (VV) and nearby components. The simulations were based on the DEMO baseline model, further refined with the minor updates of the lower port, equatorial port limiter, and upper port assemblies. The computational approach employed the Monte Carlo particle transport code MCNP for neutron and photon transport calculations, coupled with the activation and decay code FISPACT-II to determine time-dependent decay gamma source terms. The mesh-coupled Rigorous Two-Step (R2Smesh) methodology developed in KIT was applied to achieve spatially resolved decay of photon source distributions and to compute corresponding SDDR 3D maps within the DEMO reactor configuration. The results provide a detailed characterization of the residual radiation environment inside the VV, offering insight into the accumulated activity, shielding performance of different materials, and potential access scenarios for maintenance operations in next-generation fusion devices.

We present a computational sensitivity and uncertainty analysis of the effect of geometrical uncertainties on the measured reaction rates in the Water Cooled Lithium Lead experimental benchmark, which is a mock-up of a tritium breeding blanket for the future DEMO fusion facility. We used deterministic transport methods to determine the sensitivity of various parts of the geometry and Monte Carlo particle transport methods along with random sampling from geometrical uncertainties, to assess the expected uncertainty and sensitivity profiles, particularly to establish whether tritium self-sufficiency can be achieved within the uncertainty. We have established that most sensitive parts of the geometry are the source position and orientation, as well as the target detector pack position. We have established the uncertainty to be in the range of 8% to 25%, increasing with the distance from the source. We have also established the sensitivity to the displacement of the target detector pack to be roughly 5%mm-1, while the sensitivity to the perturbation of position of other detector packs and source position and orientation is generally below the 1%mm-1.

One of the fundamental challenges in the development of a fusion power plant is to integrate all the systems required for the operation of the power plant into a machine that fulfills all the design requirements. This includes designing sufficient shielding against neutrons and gamma rays to ensure that the machine can operate reliably throughout its planned lifetime. The electron cyclotron (EC) heating is one of the systems critical to heating and controlling the fusion plasma. In the EU-DEMO reactor, the EC system is integrated into an equatorial port. However, the challenge with this system is that the waveguides required for it to function act as neutron and gamma ray streaming paths, which makes controlling the nuclear loads inside and behind the EC port a challenge. This concerns both the components of the EC system itself and components and systems outside the equatorial port where the EC system is located, e.g. the superconducting toroidal field coils. The latest iteration of the EU-DEMO EC heating system from 2023 was analyzed. The main nuclear loads have been calculated and the shielding performance optimized to ensure that the design limits are met. These analyses include calculations of nuclear heating in EC port, determination of peak nuclear heating in the toroidal field coils and peak neutron-induced damage (DPA) in the components exposed to the plasma neutrons.

The EUROfusion Consortium is conducting a preconceptual design and feasibility study of a volumetric neutron source (VNS) facility to address perceived needs in the development of integrated breeding blanket and fuel cycle testing and qualification, with aim to mitigate risks stemming from the current low reliability and technical maturity of present design concepts for a DEMO fusion reactor. The main requirements driving the selection of key physics and technical concepts have been identified to cope with a construction and operation schedule consistent with the DEMO design activities.The VNS needs to provide a steady-state plasma operation based on reliable beam-target fusion plasmas with a fusion power at 30 MW, generating a peak neutron wall load of at least 0.5 MW/m2 up to neutron fluences of 30 to 50 displacements per atom. Neutronics assessments are requested to provide essential nuclear loads and shielding performances of this device. To support the technical feasibility of the VNS tokamak, a series of analyses were conducted in support of the architecture and system design concept. The preconceptual design of the VNS was based on ITER-like shielding structures at the inboard side (ca. 70 cm radial depth) and enhanced outboard side shielding (about a 120-cm radial depth).According to the assumed port configuration and toroidal segmentation of the VNS tokamak, a torus sector model of 60 deg has been generated from available computer-aided design models and converted to MCNP geometry descriptions. The primary objectives of the initial nuclear analyses and scoping assessments were the radial build on the inboard side to protect, specifically, the toroidal field coils and the shielding environment around the neutral beam injector (NBI) port duct, as well the divertor and lower ports, which are critical areas for ex-vessel shielding objectives. It could be demonstrated that the protection from a radial build adopting tungsten-based shields is adequate, whereas additional measures on shielding and integration of NBI ports and lower ports are required.

The goal of the European plasma Volumetric Neutron Source (VNS), which is a 14 MeV n-source, is to test and validate technological solutions of breeding blankets in an environment representative of a future fusion power plant such as DEMO. One of the assessed magnetic configurations is a tokamak device with a radius of ti 2.5 m that produces a D-T fusion power of ti 30 MW. The architecture of VNS integrates a single null divertor located at the bottom of the vacuum vessel. The ITER-like divertor is designed to be actively water-cooled for stationary operation and is made up with 36 cassette modules. Each module has a cassette body with a set of plasma facing components (PFCs) with tungsten as plasma facing material. The PFC arrangement is a dome positioned in the central part with reflector plates on both sides located between inboard and outboard target striking surfaces. The selected PFC technological solutions are: the plasma facing units (PFUs) of the targets are armored with tungsten monoblocs bonded onto a cooling pipe made of CuCrZr with an inserted twisted tape, the PFUs of the dome are made up with tungsten flat tiles bonded onto a CuCrZr hypervapotron cooling structure. This paper presents the status of the development of the divertor conceptual design. Preliminary analyses confirmed that the design is compatible with plasma scenarios foreseen for VNS operation.

IFMIF-DONES (International Fusion Materials Irradiation Facility-DEMO Oriented Neutron Source) is a key facility for the study and analysis of the materials properties exposed to irradiation conditions characterised by a high neutron flux with high energies up to 55 MeV. These irradiation conditions are expected by the future DEMOnstration fusion power plant (DEMO). Within IFMIF-DONES, the irradiation modules which host the specimens material are placed. Therefore, real-time monitoring of the neutron flux is essential to detect and correct any deviations, ensuring continuous and uniform irradiation. Among the diagnostics taken into account, the Self-Powered Neutron Detector (SPND) appears to offer strong physical characteristics in such an environment. However, they have so far only been used in fission reactors. In this work, a theoretical study has been carried out to analyse the signals that could be measured with these diagnostics.

The qualification of in-vessel components for a fusion power plant requires a test environment with a high flux of 14 MeV neutrons over a sufficiently large surface and volume. Performance testing and qualification of the complex design and technologies of fusion nuclear components is needed, in particular that of the tritium breeding blanket (BB). Testing in relevant conditions over a relevant time will also allow gaining the necessary confidence regarding the build-up and control of tritium inventories inside the BB, which will be an important radioactive source. An option of such a volumetric neutron source (VNS) is a beam-driven tokamak. A feasibility study of the main machine components and associated plant systems is described in this article. The machine has a major radius of 2.53 m, a single-null divertor configuration, and four tangential 120 keV beamlines that generate a fusion power of approximately 30 MW and provide current drive for a steady-state plasma scenario. The plasma is small with a minor radius of a = 0.55 m to maximize the neutron wall load, up to 0.5 MW/m(2), similar to what is targeted in ITER. Approximately 25 m(2) are available for blanket testing including 4 port plugs, which offer flexibility regarding the test module operating conditions and the implementation of instrumentation. Given the small plasma, much of the tokamak's volume is made up by the neutron shielding structures that are similarly sized as in ITER. To reduce the construction risk, ITER-like concepts were adopted for many components. In some cases, however, lessons learned from ITER led to the development of customized or innovative concepts. Due to the modest fusion power the plasma will burn <1 kg of tritium per year, which can be provided from external sources.

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.

Within the scope of the implementation of a nuclear data pipeline aiming at producing the best possible evaluated nuclear data files, a major point is the production of relevant sensitivity coefficients when including integral benchmark information. Thanks to recent code modifications in the Monte Carlo code Serpent2, it is now possible to produce these coefficients in fixed source simulations. The manuscript describes the verification of this implementation against the deterministic transport code susd3d. The study is completed by an analysis of the computational cost (running time and memory allocation) associated with such calculations with Serpent2. The relative difference between the sensitivity coefficients produced by Serpent2 and susd3d is of the order of the percent at most, except for the low energy range where the lack of neutrons prevents from reducing the Monte Carlo uncertainties. The computational cost of such calculation is similar to the one of criticality calculation mode, although the OpenMP scalability should be further improved.

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).

As part of EUROfusion's efforts to develop fusion into a viable energy source, an initial concept for a European 14 MeV volumetric neutron source (VNS), a medium-sized tokamak-based fusion machine, was developed to mainly test tritium breeding blanket technologies under reactor conditions relevant for EU DEMO. One of the challenging aspects is designing and integrating all relevant systems into the lower port area. This includes the vacuum pumping system, the divertor with its associated piping and ensuring there is enough space for divertor removal during maintenance using remote handling tools. At the same time, adequate protection of the superconducting coils, vacuum vessel and other important systems is required with regard to nuclear loads. Neutronics analyses were carried out to investigate the suitability of the initial design choices and to propose alternative solutions where appropriate. The main parameter of interest in the present analysis is the helium production in divertor cooling pipes - a parameter that is crucial for the re-weldability of the pipes after maintenance or replacement of the divertor.

The Fast Discharge Units (FDUs) of the Superconducting (SC) Toroidal Field (TF) coils in the European demonstration fusion power plant DEMO warrant the machine integrity over its full lifetime against severe failure events, such as SC coil quenches or any other plant events requiring the safe TF magnet system discharge. A low (75 kA) and a high current (105 kA) configuration are under study for the TF coils for DEMO. The FDUs must be extremely reliable for the purpose of commutating in short time (∼1 s) the currents and to discharge the TF magnets safely into resistors outside of the tokamak building. Malfunctions of the FDUs must be avoided. The FDUs are considered as Safety Important Class (SIC) components that need to discharge high amounts of energy of about 161 GJ (@75 kA) resp. 118 GJ (@105 kA) stored in the DEMO TF coils.The TF FDUs Circuit Breakers (CBs) shall be installed in the lower level of the tokamak to minimize the length of the connecting busbars. The FDUs integration is challenging because of the high neutron and gamma radiation and stray magnetic fields of the tokamak.Since in DEMO the neutron fluence over lifetime is much higher than in ITER, the problems of using FDUs with electronic subsystems was expected to be more severe, so that their integration has been considered from the beginning of the DEMO project. Sufficient shielding or possible re-positioning of the whole FDUs or sensitive FDU components compared to ITER are being investigated, to reduce the neutron fluxes and neutron and gamma ray fluences. Alternative concepts, e.g., fully mechanical CBs are studied in the EUROfusion Work Package Plant Electrical System (WPPES) in parallel.This paper presents the CAD integration work on the DEMO TF FDUs supported by neutronics assessments. It is assumed the same FDU technology as in ITER. The magnet feeder´s integration is commenced 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).

A Volumetric Neutron Source (VNS) is proposed as a compact tokamak to serve as a qualification testbed for fusion in-vessel components, breeding blanket in particular. The main goal is to determine the minimum size of the machine while maximizing the Neutron Wall Load (NWL). This paper discusses the design parameters and constraints involved in specifying the magnetic cage for VNS. The focus of the study is on sizing the TF (toroidal field) coils and related structures with the plasma magnetic field, as well as assessing how various design constraints influence the machine performance at higher magnetic field strengths. In particular, the trade-off between TF neutron heat load and dose and shielding efficiency. A very preliminary (and still evolving) design point for VNS is presented and discussed.

Abstract One of the advances in the test blanket module program within the ITER project in the last few years concerned the evolution of the pipe forest (PF) and bioshield plug (BP) designs. In support of the design phase, nuclear analyses to assess several responses in the fusion neutronics environment inside the port interspace (PI) with the existence of the evolved PF and BP are deemed essential. Nuclear analyses were commenced using the new PF and BP with developing the neutronics models and performing preliminary assessment of the radiation fields and shutdown dose rate (SDDR) in the PI. In this paper, the results of a full suite of nuclear analyses are discussed, which covers more configurations and radiation sources, in two plasma operational modes: on and off. For the plasma-on mode, different shielding options were examined. The results show a clear benefit of combining the installation of shielding panels on the PF enclosure with those in the BP “dogleg,” through which the pipes penetrate to the port cell area. For the plasma-off mode, the SDDR was assessed from different sources: activated components and residual LiPb layers in pipes after drainage. As maintenance operations are foreseen during the lifetime of the facility, the SDDR was also assessed for access conditions, open BP doors, and transport conditions, with PF extracted in the gallery.

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.

Accurate prediction of displacement damage is essential for the safety of nuclear materials irradiated by neutrons, gammas and charged particles. Displacement damage cross-sections are used for the calcula-tion of displacement damage rate. With the development of accelerator facilities and spallation neutron sources, displacement damage rate calculations in these facilities require the availability of displacement damage cross-sections for charged particles. In the present work, new arc-dpa model parameters were obtained for aluminum, iron, copper and tungsten. The displacement damage cross-sections for protons, deuterons, and alpha-particles at energies up to 100 GeV are calculated using the standard NRT and the arc-dpa model with improved parameters for these materials. Calculated displacement damage cross -sections for charged particles calculated using the arc-dpa model with new parameters agree well with available experimental data.(c) 2022 Elsevier B.V. All rights reserved.

The estimation of displacement damage rate requires the prior knowledge of displacement damage cross-section and displacement function is one part in the displacement damage cross-section calculation. The NRT model, which is used to calculate the displacement function for many years, always overestimates the number of displaced atoms in material. Although the arc-dpa model proposed recently can give correct results for the materials irradiated by neutron, there is still no suitable displacement function that can be used to accurately obtain both neutron and charged particles displacement cross-section comparing with experimental data. In the present work, a general displacement function is proposed, which is composed by power function model, arc-dpa model and BCA model. The model parameters are obtained for aluminum, iron, copper and tungsten. The displacement damage cross-sections for the materials irradiated by neutron and charged particles calculated using the general displacement function agree well with available experimental data.

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.

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.