Fusion Power Reactor Research Articles

DONES Systems identification and requirements allocation

• Systems identification of IFMIF-DONES for work distribution and responsbility allocation. • System-of-Systems to Group-of-Systems type mapping. • Alignment of Systems identification and requirement allocation. The mission of the DONES program is to develop a database of fusion-like neutron irradiation effects in the materials required for the construction of the Demonstration Fusion Power Reactor (DEMO) for benchmarking of radiation response of materials. In order to do this, the DONES facility will be built to provide a neutron source producing high-energy neutrons at sufficient intensity and irradiation volume. DONES will generate a neutron flux with a broad energy distribution covering the typical neutron spectrum of a (d-t) fusion reactor. This is achieved by utilizing Li(d,xn) nuclear reactions taking place in a liquid Li target when bombarded by a deuteron beam. The energy of the deuterons (40 MeV) and the current of the accelerator (125 mA) have been tuned to maximize the neutron flux to get irradiation conditions comparable to those in the first wall of a fusion power reactor. For the success of this mission, the needs of all the stakeholders shall be very clearly stated, considering the whole lifecycle of the facility. Once the needs are clear, they shall be translated into specific requirements, following a top-down approach, to be allocated to the different systems or groups-of-systems (GoS). To do this, it is essential to properly identify which parts shall be treated as a System or GoS, as they shall be designed accordingly, and the requirements allocated in consequence.

Concept selection for the preliminary DTT remote maintenance strategy

Divertor Tokamak Test (DTT) is the next Italian facility for nuclear fusion research aiming at bringing alternative divertor solutions to a sufficient readiness level to be adopted by the European DEMOnstrating fusion power reactor (EU-DEMO). Since a non-negligible activation is expected on plasma-facing components after DTT shutdown, remote maintenance is mandatory. This work deals with the concept selection for the DTT remote maintenance strategy, in the 2019 reference configuration. First, we present the criteria that we have derived for evaluation of design alternatives. Then, we briefly present the design alternatives developed so far for divertor and first wall remote maintenance. With this regards, three alternatives have been developed for divertor remote maintenance and two alternatives have been developed for first wall remote maintenance. The concept selection process is based on the use of ELIGERE, a decision support tool for concept selection based on the Fuzzy Analytical Hierarchy Process. More than 20 experts from several European institutions have been involved in the concept selection process. The work concludes by presenting the results of the concept selection process, in terms of optimal strategies for divertor and first wall remote maintenance.

Design of the CIEMAT Corrosion Loop for Liquid Metal Experiments

The main components of a liquid breeder blanket in a fusion power reactor are lead lithium alloy (PbLi) and the steel structure in which the liquid is enclosed (EUROFER). Several compatibility tests have shown that structural materials always suffer from corrosion attacks. The governing mechanism can be attributed to the dissolution of the steel by the liquid breeder and is strongly related to the PbLi chemistry, velocity profile, and temperature. A new facility, CiCLo-C (CIEMAT Corrosion Loop, Internally Coated), is dedicated to the study of corrosion in materials under the severe breeding blanket condition. An effort was made to design an experimental facility with a specific test section able to work at quite ambitious operation parameters: up to 550 °C and a 1 m/s flow of PbLi. Furthermore, an innovative tantalum coating was introduced in the whole loop to avoid impurities coming from the pipeline, which can disturb the measurements, and to better preserve the installation.

Activities in Divertor Reflector and Linear Plates Using WCLL and HCPB Breeding Blanket Concepts

In fusion devices, such as European Demonstration Fusion Power Reactor (EU DEMO), primary neutrons can cause material activation due to the interaction between the source particles and the targeting material. Subsequently, the reactor’s inner components become activated. For safety and safe performance purposes, it is necessary to evaluate neutron-induced activities. Activities results from divertor reflector and liner plates are presented in this work. The purpose of liner shielding plates is to protect the vacuum vessel and magnet coils from neutrons. As for reflector plates, the function is to shield the cooling components under plasma-facing components from alpha particles, thermal effects, and impurities. Plates are made of Eurofer with a 3 mm layer of tungsten, while the water is used for cooling purposes. The calculations were performed using two EU DEMO MCNP (Monte Carlo N-Particles) models with different breeding blanket configurations: helium-cooled pebble bed (HCPB) and water-cooled lithium lead (WCLL). The TENDL–2017 nuclear data library has been used for activation reactions cross-sections and nuclear reactions. Activation calculations were performed using the FISPACT-II code at the end of irradiation for cooling times of 0 s–1000 years. Radionuclide analysis of divertor liner and reflector plates is also presented in this paper. The main radionuclides, with at least 1% contribution to the total value of activation characteristics, were identified for the previously mentioned cooling times.

Activation Analysis and Evaluation of Radionuclide Inventory Decay Heat for EU DEMO Vacuum Vessel Components

Activation inventory data were assessed for different European DEMOnstration Fusion Power Reactor (EU DEMO) vacuum vessel (VV) segments in order to evaluate the activity, dose rates, and decay heat of the structural materials and the impact of heterogeneity as a part of the Safety and Environment project of the EUROfusion/Power Plant Physics Technology program. Activity is a good indicator of the condition of irradiated material while decay heat and dose rate estimates are crucial for power plant operation and public safety. In addition, the VV structures were analyzed in terms of different breeder blanket configurations that were implemented in the model. Helium-cooled pebble bed and water-cooled lithium-lead breeder blanket modules were considered. These modules directly shield the VV from neutron irradiation; therefore, their design has a great influence on subsequent structure activation.

Verification and Validation of COMSOL Magnetohydrodynamic Models for Liquid Metal Breeding Blankets Technologies

Liquid metal breeding blankets are extensively studied in nuclear fusion. In the main proposed systems, the Water Cooled Lithium Lead (WCLL) and the Dual Coolant Lithium Lead (DCLL), the liquid metal flows under an intense transverse magnetic field, for which a magnetohydrodynamic (MHD) effect is produced. The result is the alteration of all the flow features and the increase in the pressure drops. Although the latter issue can be evaluated with system models, 3D MHD codes are of extreme importance both in the design phase and for safety analyses. To test the reliability of COMSOL Multiphysics for the development of MHD models, a method for verification and validation of magnetohydrodynamic codes is followed. The benchmark problems solved regard steady state, fully developed flows in rectangular ducts, non-isothermal flows, flow in a spatially varying transverse magnetic field and two different unsteady turbulent problems, quasi-two-dimensional MHD turbulent flow and 3D turbulent MHD flow entering a magnetic obstacle. The computed results show good agreement with the reference solutions for all the addressed problems, suggesting that COMSOL can be used as software to study liquid metal MHD problems under the flow regimes typical of fusion power reactors.

Monte Carlo simulations of pure tungsten and advanced “smart” tungsten alloys under helium bombardment for future fusion power plants

Self‐passivating tungsten‐based alloys may provide a major safety advantage in comparison with pure tungsten, which is presently the main candidate material for the plasma‐facing protection of future fusion power reactors. Advanced “smart” tungsten‐based alloys adjust their properties to the environment: during the plasma operation, preferential sputtering will form almost pure tungsten surface facing the plasma. In case of an accident, the remaining alloying elements form a protective layer, preventing tungsten mobilization. The major challenge in such a scenario is to suppress the sublimation which is responsible for a release of radioactivity. Subsequent oxidation confirmed superior oxidation resistance that is to say resistance to sputtering of advanced “smart” tungsten‐based alloys compared with pure tungsten (W). In this work, we study an alloy containing tungsten (W), chromium (Cr), silicon (Si), titanium (Ti), and yttrium (Y). We have applied the Monte Carlo SRIM‐2013 simulation program to calculate the sputtering yields, number of vacancies, and backscattering yields of the first wall of pure tungsten (W) and advances “smart” tungsten‐based alloys (W‐Cr, W‐Si, W‐Cr‐Si, W‐Cr‐Ti, and W‐Cr‐Y) by helium ion bombardment. It is found that the sputtering yields of advanced “smart” tungsten alloys are very low than that of pure (W) at each incident energy and incidence angle. On the other hand, the sputtering yield depends on the properties of both the incident particle and the target.

Optimization of S-CO[formula omitted] power conversion layouts with energy storage for the pulsed DEMO reactor

Nowadays, the technology of fusion power reactors faces various engineering obstacles. One of the most discussed topics is the pulse and dwell period of the fusion power reactor, which may be the most challenging for material engineering, electrical power grid, energy storage systems, etc. The main goal of this research is to calculate the optimized parameters for power cycles to gain the highest efficiency. This article also describes the optimization method of a possible supercritical CO2 power conversion cycle in DEMO. For comparison, there were chosen two Brayton power cycle layouts using S-CO2. The calculation focuses on the optimization due to the difference between the gained electrical power for these two periods, which lowers the oscillations in the power output. Besides, the approach of calculation via upgraded software for the power conversion cycles optimization is described. The optimization of this specific DEMO design points to the simple Brayton cycle as a better choice due to simplicity, higher flexibility and slightly higher gross efficiency.

Performance Optimization of Power Generation System for Helium Gas/Lead Lithium Dual-coolant Blanket System

Helium gas (He-gas) / Lithium Lead (LiPb) Dual-cooled LiPb (DLL) blanket is one of the promising candidates for the DEMO fusion reactor for power generation. Brayton cycle based on He-gas has been used as the power generation system for the fusion and fission power reactors. In this work, a modified and efficient design of the power generation system with concept of two stage expansions and one intercooler named as Schematic-I has been proposed for the DLL blanket of Fusion Power Reactor (FDS-II) to increase the thermal performance (e.g. power density and efficiency) of the system and optimized the parameters of the system. The engineering equation solver tool was used to calculate the performance of the system. It is found that the use of more than one expansion stage with proper selection of intercooler is very effective as compared with either splitting the flow equally for the expansion in two turbines or a single stage expansion. The concept of two-stage expansions in the Schematic-I increased the thermal efficiency of the system up to 51%, which was 4% higher than previous design of DLL blanket power generation system. The main parameters of the system such as isentropic efficiency of turbine and compressor, pressure ratio, recuperator performance, the terminal temperature difference of the recuperator, the inlet temperature of DLL blanket and effect of different pressures on the system have all been analyzed and optimized to achieve high performance.

Optimization of the supercritical CO2 power conversion system based on the net efficiency under conditions of the pulse-operated fusion power reactor DEMO

Fusion power plants represent a new energy technology with specific features including multiple heat sources with different outlet temperature and power, fluctuating high self-consumption, and in the case of the demonstration fusion power plant, the pulse operation of the heat sources. The supercritical CO2 (S-CO2) Brayton cycle was applied and optimized under DEMO conditions with the nuclear net efficiency as the main optimization criterion. The optimization was performed for helium-cooled and water-cooled reactor blanket concepts, and for simple and recompression S-CO2 cycle layouts with four heat sources. The optimization process used a brute-force search problem-solving technique applied to multiparametric space. The model data was taken from the European model of the fusion power plant DEMO1 2019. When applying the simple S-CO2 cycle, the net efficiency of the model power plant with the helium-cooled and water-cooled blanket was found to be 19.5% and 12.3%, respectively, whereas when applying the recompression S-CO2 cycle, the net efficiency was found to be 16.4% and 8.1%, respectively. The simple S-CO2 cycle provided higher net efficiency for the model fusion power plants with the given configuration of multiple heat sources than the recompression S-CO2 cycle, and the model power plants using the helium-cooled blanket achieved higher net efficiency compared to the power plants using the water-cooled blanket despite significantly higher self-consumption. The application of the simple S-CO2 cycle allowed the model fusion power plant with the helium-cooled blanket to achieve the net efficiency higher than the reference net efficiency of 17.9% of the current DEMO model using the steam cycle. For the power plants with the water-cooled blankets, S-CO2 cycles were found to be less efficient than the steam cycles.

Nuclear performances of the water-cooled lithium lead DEMO reactor: Neutronic analysis on a fully heterogeneous model

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

Physical Background, Computations and Practical Issues of the Magnetohydrodynamic Pressure Drop in a Fusion Liquid Metal Blanket

In blankets of a fusion power reactor, liquid metal (LM) breeders, such as pure lithium or lead-lithium alloy, circulate in complex shape blanket conduits for power conversion and tritium breeding in the presence of a strong plasma-confining magnetic field. The interaction of the magnetic field with induced electric currents in the breeder results in various magnetohydrodynamic (MHD) effects on the flow. Of them, high MHD pressure losses in the LM breeder flows is one of the most important feasibility issues. To design new feasible LM breeding blankets or to improve the existing blanket concepts and designs, one needs to identify and characterize sources of high MHD pressure drop, to understand the underlying physics of MHD flows and to eventually define ways of mitigating high MHD pressure drop in the entire blanket and its sub-components. This article is a comprehensive review of earlier and recent studies of MHD pressure drop in LM blankets with a special focus on: (1) physics of LM MHD flows in typical blanket configurations, (2) development and testing of computational tools for LM MHD flows, (3) practical aspects associated with pumping of a conducting liquid breeder through a strong magnetic field, and (4) approaches to mitigation of the MHD pressure drop in a LM blanket.

Microstructural evolution and strain hardening rule for V-5Cr-5Ti alloy

Abstract V-5Cr-5Ti alloy is a promising candidate structural material for fusion-power reactor blankets. However, its plastic deformation behavior, which is important to the safety and reliability of a fusion reactor, is not sufficiently understood. The present paper develops a physically based strain hardening model to characterize the plastic deformation of V-5Cr-5Ti alloy. Using miniaturized specimens of V-5Cr-5Ti alloy, uniaxial tensile tests and microstructural evolution analysis are performed at different strain levels. Microstructural evolution results show that the existence of Ti-enriched second phase and the dislocation interactions are critical to the plastic deformation of V-5Cr-5Ti alloy. On the above physical basis, the Ti-enriched second phase and the evolution equations of the dislocation density are developed, while the flow stress rule is formulated from the summation of athermal stress, thermally activation stress, and dispersed-phase stress. The finite element method based on the user-material subroutine UMAT (User-defined Material Mechanical Behavior) of commercial ABAQUS code (finite element analysis software) and the implicit stress update algorithm are adopted to realize the proposed physically based strain hardening model. Results suggest that this new strain hardening model is superior to conventional models for providing a more accurate and reasonable prediction of the plastic deformation behavior of V-5Cr-5Ti alloy.

Atom probe characterisation of segregation driven Cu and Mn–Ni–Si co-precipitation in neutron irradiated T91 tempered-martensitic steel

The T91 grade and similar 9Cr tempered-martensitic steels (also known as ferritic-martensitic) are leading candidate structural alloys for fast fission nuclear and fusion power reactors. At low temperatures (300–400 °C) neutron irradiation hardens and embrittles these steels, therefore it is important to investigate the origin of this mode of life limiting property degradation. T91 steel specimens were separately neutron irradiated to 2.14 dpa at 327 °C and 8.82 dpa at 377 °C in the Idaho National Laboratory Advanced Test Reactor. Atom probe tomography was used to investigate the segregation driven formation of Mn–Ni–Si-rich (MNSPs) and Cu-rich (CRP) co-precipitates. The precipitates increase in size and, slightly, in volume fraction at the higher irradiation temperature and dose, while their corresponding compositions were very similar, falling near the Si(Mn,Ni) phase field in the Mn–Ni–Si projection of the Fe-based quaternary phase diagram. While the structure of the precipitates has not been characterised, this composition range is distinctly different than that of the typically cited G-phase. The precipitates are composed of CRP with MNSP appendages. Such features are often observed in neutron irradiated reactor pressure vessel (RPV) steels. However, the Si, Ni, Mn, P and Cu solutes concentrations are lower in the T91 than in typical RPV steels. Thus, in T91 precipitation primarily takes place in solute segregated regions of line and loop dislocations. These results are consistent with the model for radiation induced segregation driven precipitation of MNSPs proposed by Ke et al. Cr-rich alpha prime (α’) phase formation was not observed.

Interface design of the blanket system for CFETR

The China Fusion Engineering Test Reactor (CFETR) aims to bridge the gaps between fusion experimental reactor ITER and demonstration fusion power reactor. Blanket, one of the key components of the CFETR, is of great importance to achieve the functions of tritium breeding, heat extraction and radiation shielding. As for the next device for fusion energy research in China, two blanket candidate options have been developing in parallel to decrease the risk of both physical and technical issues. Apart from the research and design work on the breeding blanket module itself, the interfaces definition and integration of the blanket system is vital, particularly due to the system interdependencies. The interface design of the blanket system is a challenge due to several factors, such as the thermal expansion, the electromagnetic loads, the feasibility of remote maintenance. In this paper, a preliminary interface definition and integration design of the CFETR blanket system has been presented, based on the design with R = 7.2 m and a = 2.2 m. The purpose is to complete the interface definition and integrated design of the blanket and other components or systems. The vacuum vessel has 16 sectors, and each sector has two inboard and three outboard blanket segments. The blanket segment will be installed and removed by remote maintenance tools through the upper ports. The work addresses the interfaces between the blanket and other system, such as diagnostic system and heating system. Preliminary assessment of the impact on tritium breeding ratio (TBR) for the candidate water cooled ceramic breeder blanket integration scheme has been presented highlighting the influence of the auxiliary interfaces system to TBR.

Hybrid 1D + 2D Modelling for the Assessment of the Heat Transfer in the EU DEMO Water-Cooled Lithium-Lead Manifolds

The European demonstration fusion power reactor (EU DEMO) tokamak will be the first European fusion device to produce electricity and to include a breeding blanket (BB). In the framework of the design of the EU DEMO BB, the analysis of the heat transfer between the inlet and outlet manifold of the coolant is needed, to assess the actual cooling capability of the water entering the cooling channels, as well as the actual coolant outlet temperature from the machine. The complex, fully three-dimensional conjugate heat transfer problem is reduced here with a novel approach to a simpler one, decoupling the longitudinal and transverse scales for the heat transport by developing correlations for a conductive heat-transfer problem. While in the longitudinal direction a standard 1D model for the heat transport by fluid advection is adopted, a set of 2D finite elements analyses are run in the transverse direction, in order to lump the 2D heat conduction effects in suitable correlations. Such correlations are implemented in a 1D finite volume model with the 1D GEneral Tokamak THErmal-hydraulic Model (GETTHEM) code (Politecnico di Torino, Torino, Italy); the proposed approach thus reduces the 3D problem to a 1D one, allowing a parametric evaluation of the heat transfer in the entire blanket with a reduced computational cost. The deviation from nominal inlet and outlet temperature values, for the case of the Water-Cooled Lithium-Lead BB concept, is found to be always below 1.4 K and, in some cases, even to be beneficial. Consequently, the heat transfer among the manifolds at different temperatures can be safely (and conservatively) neglected.

Likely U.S. Regulatory Considerations for D-T Fusion Power Reactors

The Electric Power Research Institute outlined three criteria important for a commercially viable fusion power plant: competitive electric power cost, regulatory simplicity, and public acceptance. In this paper we consider likely U.S. regulatory considerations for deuterium-tritium (D-T) fusion power reactors, relying on existing criteria and past actions by the U.S. Nuclear Regulatory Commission, which has asserted regulatory jurisdiction over U.S. commercial fusion reactors. We begin with consideration of a basic D-T fusion reactor, independent of plasma confinement approach. Because tritium and radioactivity are present, likely regulation will require containment structures and various safety measures for each component. Regulators are certain to require that all nuclear components of the system be housed in an overall containment vessel that must be held at less than atmospheric pressure to contain any leakage of tritium, radioactive corrosion products, radioactive coolant, and activated elements in the air. In addition, regulators are sure to require plant structure and operations that minimize the potential for clandestine plutonium breeding. Next, we add superconducting magnets and a plasma dump (divertor) to the basic system and recognize the small but nonzero probability of those magnets explosively quenching, potentially causing reactor damage and dramatically increasing containment vessel pressure. Finally, we consider ITER as prototypical of a D-T–fueled fusion power reactor. Because ITER-like systems are subject to damaging plasma disruptions, regulators are almost certain to require safeguards against such events significantly damaging first walls and subsystems. Finally, we believe that regulators are not likely to back off significantly in requirements related to the deuterium-deuterium and D3He fuel cycles even though the tritium production and the neutron damage in the latter fuel cycle are significantly below those in a D-T system. However, regulations for p11B and 3He3He fuel cycles are certain to be dramatically less demanding because of the lack of tritium and essentially no neutron production.

Electron microscopy and atom probe tomography of nanoindentation deformation in oxide dispersion strengthened steels

Oxide Dispersion Strengthened (ODS) steels are candidates for fuel cladding materials in sodium-cooled fast reactors and for structural materials in nuclear fusion power reactors. The effect yttrium-titanium-oxygen (Y-Ti-O) nano-oxide precipitates within ODS steels have on the micromechanical deformation mechanisms has been investigated. The aim is to assess the extent of any direct link between the Y-Ti-O dispersion and nanoindentation hardness, using electron backscatter diffraction, transmission electron microscopy (TEM), and atom probe tomography (APT) studies of a Fe-14Cr-3W-0.2Ti-0.25Y2O3 (wt%) ODS steel at room temperature. Y-Ti-O nanoclusters had a non-uniform distribution and average number density that ranged from 5.8 ± 0.1 × 1023 to 1.3 ± 0.1 × 1024 m−3 and Guinier radii ranging from 1.8 ± 0.2 nm to 2.1 ± 0.2 nm. Surprisingly, the local Y-Ti-O distribution did not correlate strongly with the nanohardness, indicating that the dominant hardening mechanisms was at best only weakly related to the Y-Ti-O distribution. Instead, scanning TEM and APT confirmed that the dominant hardening mechanism was due to the grain boundary refinement, and was further enhanced by tungsten enrichment to ~3.5 at.% at grain boundaries.

Nuclear analysis of the Water cooled lithium lead DEMO reactor

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

Shut-Down Dose Rate Distribution Calculation in the Access Cell With Activated HFTM in the IFMIF-DONES Facility

International Fusion Materials Irradiation Facility–DEMOnstration Power Station (DEMO)-Oriented Neutron Source (IFMIF-DONES) is an accelerator-based d-Li neutron source that is aimed at the qualification of materials at the irradiation conditions of a DEMO fusion power reactor. It is being developed in the frame of EUROfusion’s Power Plant Physics and Technology (PPPT) work program. The intense neutron radiation produced in the liquid lithium target results in a strong activation of the test cell (TC) and its high flux test module (HFTM). This article presents the analyses performed within the Early Neutron Source (ENS) project of EUROfusion/PPPT directed to provide up-to-date estimates of the activity inventories and the decay heat production in the DONES TC facility. To this end, a series of coupled Monte Carlo N-Particle code (MCNP)+ McDelicious transport and FISPACT inventory calculations were performed using the 2017 DONES TC model and nuclear cross section data from the Joint Evaluated Fission and Fusion (JEFF)–3.1.2 data library. The article presents the shut-down dose rate maps generated by HFTM that was irradiated at the TC and later relocated to the access cell. Performed calculations used to addresses the issue of the radiation dose loads that are to be expected due to the activated components/systems including the HFTM. In addition to that, activation calculations of HFTM materials are also provided.