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In the course of a severe accident in light water reactors with core degradation, so-called debris beds can be formed inside the reactor pressure vessel or in the reactor cavity. The strategy to analyse the coolability of such debris beds with both experiments and numerical simulations is discussed. The numerical simulations are carried out with MEWA (MElt WAter) code, being developed at the institute for the prediction of the thermal-hydraulic conditions inside a debris bed, including the prediction of dryout heat flux. The simulations show good agreement with experimental data of the DEBRIS experiments. [ABSTRACT FROM AUTHOR]

Abstract: The High Performance Light Water Reactor (HPLWR), how the European Supercritical Water Cooled Reactor is called, is a pressure vessel type reactor operated with supercritical water at 25MPa feedwater pressure and 500°C average core outlet temperature. It is designed and analyzed by a European consortium of 10 partners and 3 active supporters from 8 Euratom member states in the second phase of the HPLWR project. Most emphasis has been laid on a core with a thermal neutron spectrum, consisting of small fuel assemblies in boxes with 40 fuel pins each and a central water box to improve the neutron moderation despite the low coolant density. Peak cladding temperatures of the fuel rods have been minimized by heating up the coolant in three steps with intermediate coolant mixing. The containment design with its safety and residual heat removal systems is based on the latest boiling water reactor concept, but with different passive high pressure coolant injection systems to cause a forced convection through the core. The design concept of the steam cycle is indicating the envisaged efficiency increase to around 44%. Moreover, it provides the constraints to design the components of the balance of the plant. The project is accompanied by numerical studies of heat transfer of supercritical water in fuel assemblies and by material tests of candidate cladding alloys, performed by the consortium and supported by additional tests of the Joint Research Centre of the European Commission. Besides the scientific and technical progress, the HPLWR project turned out to be most successful in training the young generation of nuclear engineers in the technologies of light water reactors. More than 20 bachelor or master theses and more than 10 doctoral theses on HPLWR technologies have been submitted at partner organizations of this consortium since the start of this project. [Copyright &y& Elsevier]

Abstract: The High Performance Light Water Reactor (HPLWR) is a thermal spectrum nuclear reactor cooled and moderated with light water operated at supercritical pressure. It is an innovative reactor concept, which requires developing and applying advanced analysis tools as described in the paper. The relevant water density reduction associated with the heat-up, together with the multi-pass core design, results in a pronounced coupling between neutronic and thermal-hydraulic analyses, which takes into account the strong natural influence of the in-core distribution of power generation and water properties. The neutron flux gradients within the multi-pass core, together with the pronounced dependence of water properties on the temperature, require to consider a fine spatial resolution in which the individual fuel pins are resolved to provide precise evaluation of the clad temperature, currently considered as one of the crucial design criteria. These goals have been achieved considering an advanced analysis method based on the usage of existing codes which have been coupled with developed interfaces. Initially neutronic and thermal-hydraulic full core calculations have been iterated until a consistent solution is found to determine the steady state full power condition of the HPLWR core. Results of few group neutronic analyses might be less reliable in case of HPLWR 3-pass core than for conventional LWRs because of considerable changes of the neutron spectrum within the core, hence 40 groups transport theory has been preferred to the usual 2 groups diffusion theory. Successively, with the usage of a developed pin-power reconstruction technique capable to account for the innovative fuel assembly design, sub-channel investigations of the individual fuel assemblies have been performed evaluating pin-wise clad temperatures. Obtained results will be discussed giving a detailed insight of the revolutionary HPLWR 3 pass core concept and understanding the physical reasons, which influence the local clad temperatures. The obtained results represent a new quality in core analyses, which takes into full consideration the coupling between neutronics and thermal-hydraulics as well as the spatial effects of the fuel assembly heterogeneity in determining the local pin-power and the associated maximum clad temperature. [Copyright &y& Elsevier]

Abstract: The High Performance Light Water Reactor (HPLWR), a supercritical water cooled reactor concept with multiple heat-up steps, requires efficient mixing of the coolant between these steps to minimize hot spots in the core. Analyzing and improving the mixing in the mixing chamber above the core, situated between evaporator and superheater assemblies, and below the core, between the first and second superheater, is one of the challenges in the design process of the HPLWR. Different measures to enhance mixing have been studied with CFD analyses, in which a new design approach has been applied to the upper mixing chamber. It simplifies the complex structures and takes the effects of the disregarded structures into account by introducing source terms into the momentum equations. [ABSTRACT FROM AUTHOR]

Abstract: The paper describes a novel core concept for a nuclear reactor cooled with supercritical water, in which the coolant is heated up from 280°C at the reactor inlet to 500°C at the outlet in four steps: a first heat-up step is provided by heat transfer from fuel assemblies to the moderator water in gaps and moderator boxes, a second step is foreseen in a central “evaporator” and two further steps in a first and a second superheater surrounding it. The coolant flow scheme includes upward and downward flow through the core with intermediate mixing in chambers above and below the core to eliminate hot streaks. A preliminary single channel analysis, concentrating on an average flow channel and on the hottest one only, indicates that such core design can match the limits of cladding materials available today. Even though the resultant pressure drop of the coolant will be higher than usual, it is expected that the assembly boxes can be designed with acceptable deformations. [Copyright &y& Elsevier]

Objective: To investigate the effects of tibolone on carbohydrate metabolism, and to compare these effects with those of a sequential regimen of conjugated equine estrogens and medrogestone. Methods: This was an open-label, multicentre, comparative study. Seventy-two postmenopausal women were randomized to receive either tibolone 2.5 mg/day or conjugated equine estrogens 0.6 mg plus sequential medrogestone 5 mg (CEE/M) for six 28-day cycles. Carbohydrate metabolism was evaluated at baseline and after three and six cycles of treatment by an oral glucose tolerance test (OGTT). A blood sample was taken at 30, 60, 90 and 120 mm after glucose 75 mg dosing for determination of plasma glucose, insulin and connecting peptide (C-peptide) levels. Results: The changes from baseline of glucose, insulin and C-peptide area-under-the-curve (AUC) values were not statistically significant after 3 and 6 months of tibolone or CEE/M treatment. There was a small transitory decrease in HbA1C after three cycles of treatment with tibolone. Conclusion: The effects of tibolone and CEE/M on carbohydrate metabolism were considered to have no clinical significance. [Copyright &y& Elsevier]

• New experimental sCO 2 dataset with 196 experiments and 8950 data points. • Thermal inflow lengths were quantified with up to 480 d i for fully developed flow. • Experiments categorised in normal, enhanced and deteriorated heat transfer regimes. • Heat transfer deterioration in the new dataset is caused by buoyancy effects. • Two heat transfer correlations predict the experimental data with 30 % accuracy. Supercritical CO 2 shows great potential as a working fluid in power plant cycles due to its moderate critical pressure of 7.38 MPa and critical temperature of 30.98 °C, closely matching typical heat sink temperatures. Understanding the heat transfer characteristics of sCO 2 is crucial for designing cycle components. This publication presents a systematic analysis of sCO 2 heat transfer in two heated vertical pipes with 4 and 8 mm inner diameters, with both upward and downward flow at pressures of approximately 7.75, 8.00, and 9.50 MPa, and flow inlet temperatures between 5 and 40 °C, based on 196 experiments. The study investigates a range of mass fluxes from 400 to 2000 kg/m2s and heat fluxes from 10 to 195 kW/m2, resulting in a heat to mass flux ratio of 6 to 275 J /kg. The findings reveal enhanced, normal, and deteriorated heat transfer within the experimental dataset. Buoyancy effects are identified as the main cause of deteriorated heat transfer in the investigated parameter range, with flow acceleration showing no significant influence on heat transfer. A new proposed dimensional criterion categorises 36 out of 119 experiments with upward flow in the deteriorated heat transfer regime, accompanied by temperature peaks rising to 47 K. Thermal inflow lengths range from 0 to 480 inner pipe diameters, with some experiments not achieving a thermally fully developed flow over the entire pipe length. Based on a comparison with approximately 8950 experimental data points, two Nusselt correlations are found, capable of reproducing the experimental results with a mean absolute deviation of around 30 %. This publication provides valuable data for validating numerical models and developing correlations to predict the heat transfer of supercritical fluids. [ABSTRACT FROM AUTHOR]

The application of supercritical fluids as an alternative heat transfer medium in thermal processes is becoming increasingly important, whereby the understanding of their pressure drop characteristics is essential for the process and component design. With a total of 96 experiments, this publication shows a systematic analysis of the pressure drop of CO 2 flow at supercritical pressures in a heated smooth pipe with an inner diameter of 4 mm, at a pressure of 7.75 MPa, mass fluxes up to 2000 kg/m2s and heat fluxes up to 235 kW/m2. The hydrostatic pressure drop accounts for between 4 % and 24 % of the total pressure drop and the pressure drop due to flow acceleration for between 12 % and 30 %, with the frictional pressure drop accounting for the largest percentage. It was found that the Filonenko correlation can predict the pipe friction pressure drop in the investigated parameter range with a mean absolute deviation of 7 %. [Display omitted] • Pressure drop database with 34 isothermal and 62 heated scCO 2 pipe flows. • Experiments were categorised in the normal or enhanced heat transfer regime. • Between 4 % and 24 % for hydrostatic and 12 and 30 % for acceleration pressure drop. • Pipe friction pressure drop has the largest share of total pressure drop. • Filonenko correlation predicts friction pressure drop with a 7 % mean absolute deviation. [ABSTRACT FROM AUTHOR]

[Display omitted] • Key elements of the Europe-wide strategic agenda for nuclear education, training, and knowledge management proposed. • Urgent and sustainable strategic support, coordination, and partnership between all nuclear stakeholders, is needed. • Challenges in attraction, development and retaining of nuclear talents are discussed. The key elements of the EU-wide strategic agenda for nuclear education, training, and knowledge management are proposed in this paper. They were developed partly within the ENEN+ and ENEN# projects, to support and consolidate the efforts of the nuclear stakeholders to attract, develop, and retain new talents. The paper considers the projections of needs developed by the European Human Resources Observatory-Nuclear (EHRO-N). It builds on the existing national nuclear education strategies in the EU and beyond and more than two decades of experience with the European Nuclear Education Network (ENEN). The paper outlines and substantiates the most important actions and risks related to human resources, which need to be managed successfully for the EU to contribute to the tripling of the nuclear power generation capacity on the planet by 2050 (as announced in COP28) and to retain its leading role in the nuclear power and non-power applications. The sheer complexity of this challenge calls for high-level strategic support, coordination, and partnership between all nuclear stakeholders, especially those involved in the decision-making. The present situation calls for urgent strategic actions with long-term sustainability. [ABSTRACT FROM AUTHOR]

• Homogenization is non-conservative and only applicable for very small particles sizes. • The regular lattice model is suitable if an adequate equivalent diameter d eq is chosen. • A unique d eq value may not represent adequately the debris in the whole porosity range. • Taking very high d eq do not always guarantee conservative results. • The BCC lattice is preferred due to its simplicity and the covered porosity range. After the Fukushima accident, the interest of the scientific community in severe accident research has been renewed. One of the severe accident research issues that needs to be further investigated is the recriticality potential of the debris bed that is formed after the core meltdown. The uncertainty regarding fuel debris conditions is very high. Consequently, one of the challenges of the criticality evaluation is the neutronic modeling of the debris itself. Conservative assumptions and simplifications have to be performed to overcome the uncertainties and to achieve a computationally feasible debris bed model. This paper presents a suitability analysis of several debris bed models for a criticality evaluation. The objective is to identify the most suitable model, i.e. the model with the best compromise between accuracy and simplicity, from a set of candidates. For that, a detailed near-to-reality model was developed whose higher fidelity results have been used as a reference to evaluate the adequacy of various simplifications. This work focuses on the modeling of the porous internal structure of the debris, concretely on the fuel particles, which are characterized by their shape, size and spatial distribution. The final proposed neutronic model represents the debris bed as a regular 3D arrangement of monosized spherical particles. This is a very suitable model delivering accurate results as long as an appropriate equivalent diameter d eq is chosen. However, it was proved that not always the same d eq value is adequate for representing the debris in the whole porosity range. The Monte Carlo codes MCNP6.1 and Serpent 2.1 were used to construct the debris bed models and to compute the infinite multiplication factor (k ∞). [ABSTRACT FROM AUTHOR]

The latent heat transfer during vapour condensation in the condenser section of passive heat transport devices such as the two-phase closed thermosiphon is limited by film condensation. Dropwise condensation provides an increase of the heat transfer coefficient by up to one order of magnitude and can be achieved with a water-repellant surface. The inner surface of pipes made from stainless steel was functionalized by laser surface texturing with ultrashort laser pulses and subsequent storage in a liquid containing long-chained hydrocarbons. The pipes were separated into half-pipes by wire eroding to enable laser texturing of the inner surface, and were then joined by electron beam welding after laser texturing. As a result, superhydrophobic and water-repellent surfaces with a contact angle of 153° were obtained on the inner surface of the pipes with a length of up to 1 m. The functionalized pipes were used in the condenser section of a two-phase closed thermosiphon to demonstrate a heat transfer rate of 0.92 kW at 45 °C, which is approximately three times the heat transfer rate of 0.31 kW of a smooth reference pipe. [ABSTRACT FROM AUTHOR]

The use of heat pipes (HP) for the DEMO in-vessel plasma-facing components (PFCs) has been considered because of their high capacity to transport the heat from a heat source to a heat sink by means of the vaporization and condensation of the working fluid inside and their ability to enlarge the heat transfer area of the cooling circuit substantially. Recent engineering studies conducted in the framework of the EUROfusion work package Divertor (Wen et al, 2021) indicate that it is possible to design a heat pipe with a capillary limit above 6 kW using a composite capillary structure (wherein axial grooves cover the adiabatic zone and the condenser, and sintered porous material covers the evaporator). This power level would correspond to an applied heat flux of 20 MW/m², rendering such a design interesting with respect to a divertor target concept. To validate the results of the initial engineering analysis, several experiments have been conducted to evaluate the actual performance of the proposed heat pipe concept. The present contribution presents the experiment's results regarding the examination of the operating limits of two different designs for an evaporator: one featuring a plain porous structure, and one featuring ribs and channels. [ABSTRACT FROM AUTHOR]

Nuclear research reactors (RR) are essential facilities in countries implementing nuclear power plants and are used for experiments necessary for commercial reactor development, training and education programs, and many other applications not related to nuclear energy production (e.g., isotope production, neutron sources, materials science). Europe has a broad and very diverse landscape of RRs, many of which have been in operation for 30-60 years, are well maintained and regularly modernized. However, financial pressures caused by a combination of declining interest and the lack of a sound financial model have led to the closure of many of them (e.g. OSIRIS in Saclay, JEEP II research reactor at IFE Kjeller and BER2 in Berlin). These negative trends called for coordinated European action to assess the impact of the declining number of RRs. The Towards Optimized Use of Research Reactors (TOURR) project was a response to this challenge. Its main objective was to assess the status of the EU RR fleet and to develop a strategy for the refurbishment and construction of new RR in Europe. The assessment was based on analysed data obtained through extensive questionnaires sent to all operating European RR. The analysis revealed gaps in terms of lack of long-term funding, lack of manpower and lack of communication between RRs and their customers. It also showed threats of further European RR closures. Regarding the long-term EU RR strategy, the main recommendations of the TOURR project are to build (at least) two RRs, a medium-flux multipurpose reactor and a flexible zero-power facility. Both reactor cores could be part of a single facility built at the European level and accessible to all EU Member States. • Current status of the European Research Reactor fleet analysed. • Main gap in EU RR utilization is the lack of manpower and lack of financing. • RR research opportunities in new fields identified (e.g. processing of chemicals) • At least one new multipurpose medium-flux reactor is needed in EU. • In addition, results indicate the need for flexible zero-power reactor. [ABSTRACT FROM AUTHOR]

Abstract: A novel concept of a pressurized water reactor with a primary loop cooled with supercritical water is introduced and analyzed in this work. A steam cycle analysis has been performed to illustrate the advantages of such a nuclear power plant with respect to specific power and thermal efficiency. Moreover, a reactor pressure vessel concept including its internals and a suitable core and fuel assembly design are presented overcoming the problems, which arise due to the high heat up of the coolant and the density change involved with it. The core power and coolant density distributions are predicted with coupled neutronic and thermal-hydraulic analyses. The method features the definition of inlet orifices for coolant mass flow adjustment within the core as well as an additional tool for the interpolation of local pin power data. The latter one has been used for a successive sub-channel analysis of the hottest fuel assembly of the core, which provides a more detailed spatial resolution and thus predicts peak cladding temperatures, the maximum linear pin power of fuel pins, and maximum fuel temperatures. It can be shown that maximum temperatures of claddings and fuel are well below the material limits. Thanks to an average core exit temperature below the pseudo-critical temperature, the core concept leaves enough margin for additional uncertainties and allowances for operation. [ABSTRACT FROM AUTHOR]

Abstract: The European concept of the High Performance Light Water Reactor (HPLWR) differs from current light water reactors in a higher system pressure beyond the critical point of water, as well as a higher heat-up of the coolant within the core and thus higher core outlet temperatures, leading to a significant increase in turbine power and thermal efficiency of the power plant. The motivation to develop a novel fuel assembly for the HPLWR is caused by the high variation of coolant density in the core by more than a factor of seven. A systematic design study shows that a square fuel assembly with two rows of fuel rods and a central moderator box is best to minimize the structural material, to optimize the moderator to fuel ratio and to reduce differences of fuel rod power. Using neutronic and thermal-hydraulic analyses, a detailed mechanical design of a fuel assembly of the HPLWR has been worked out. Moreover, concepts for the head piece, the foot piece, the steam plenum and the lower mixing plenum, including the lower core plate, have been developed to account for the individual flow paths of this reactor. These allow a leak-tight counter current flow of moderator water and coolant as well as uniform mixing of different mass flows. The assembly design concept can be used as a general key component for any advanced core design of this reactor. [Copyright &y& Elsevier]

The supercritical carbon dioxide (sCO 2) heat removal system, which is based on multiple closed Brayton cycles with sCO 2 as the working fluid, is an innovative, self-propelling and modular heat removal system for existing and future nuclear power plants. Previous studies analysed its design, layout, control and operation. In addition, this study considers different sudden failures during the accident progress, e.g. failure of single sCO 2 cycles, control systems and valves. These abnormal conditions were investigated with the thermal-hydraulic system code ATHLET for a generic Konvoi pressurized water reactor. In most cases, the failure of a single sCO 2 cycle can be compensated. On the one hand, failure of the fans of the gas cooler leads to a pressure increase which may be mitigated by an inventory control system or cycle shutdown. On the other hand, unintended fan speed-up can cause compressor surge without adequate countermeasures. Furthermore, the system can operate under the cyclic blow-off from the steam generator safety valves when the relief valves are not available. Finally, the unintended closure of the valve which controls the steam flow through the compact heat exchanger triggers a fast cycle shutdown but a subsequent restart might be possible. • Transient simulation of a supercritical carbon dioxide heat removal system. • Successful operation in interaction with a pressurized water reactor. • Analysis of different sudden failures, e.g. of single cycles, control systems and valves. • Some failures are mitigated by the applied robust design of the heat removal system. • For other failures, fast interventions or further design or control improvements are proposed. [ABSTRACT FROM AUTHOR]

• History of ENEN in support to nuclear knowledge communities. • Biggest nuclear Education and Training Organization in Europe. • European Commission funded education. • Networking among Universities and Research Centers. • Novel technologies in Nuclear. The European Nuclear Education Network (ENEN) just celebrated in 2023 the first twenty years of existence. During this period, the ENEN network grew, reaching today more than ninety Members, Partners and Supporters. The mission of ENEN is the preservation and the further development of expertise in the nuclear fields by higher Education and Training. We target to reach this objective through the co-operation between universities, research organisations, regulatory bodies, the industry and any other organisations involved in the application of nuclear science and ionising radiation. Networking is an important method that we use to achieve success in our actions. Either if we speak about networking between people that aim at delivering specific education and training (E&T) actions (attraction to nuclear, higher education and training, workshops, webinars, summer/winter schools, etc.) or networking between persons who need an upgrade in their professional career through these E&T actions, we strongly believe that collaboration is the answer for a safe usage of nuclear energy in Europe and beyond. Today, ENEN is the response to the European need for highly educated people in the nuclear field by meeting these needs through cooperation between industry, research centers, technical support organizations international organizations and higher education institutions. This paper summarizes the recent evolution of ENEN after twenty years of existence and the vision for the coming period, in the service of nuclear education in Europe. [ABSTRACT FROM AUTHOR]

This publication provides a dataset of 54 experiments, carried out with supercritical CO 2 (sCO 2) in two heated horizontal pipes with inner diameters of 4 and 8 mm at a pressure of approximately 7.75 MPa. The mass fluxes were set to 400 and 800 kg/m2s and the heat flux was varied from 30 to 130 kW/m2, resulting in a heat to mass flux ratio of 38–225 J/kg. The influences of the inner pipe diameter, the Reynolds number and the heat to mass flux ratio on the thermal stratification were studied. Temperature differences up to 90 K between the bottom and the top of the pipe have been detected. By varying the bulk fluid inlet temperature, the thermal inflow lengths have been quantified for a fully developed temperature stratification with up to 180 inner pipe diameters. The Petukhov and Richardson buoyancy criteria show that the experimental results are strongly influenced by buoyancy. For the Jackson buoyancy criterion, a new threshold of higher than 160 is proposed. The validation of four empirical Nusselt correlations shows that the Bishop correlation has the lowest total average deviation of 52%, especially for the bottom site of both pipes with an average deviation of 29%. • 54 Experiments with heated horizontal sCO 2 flows in pipes were carried out. • Thermal stratifications were investigated with temperature differences up to 90 K. • Thermal inflow lengths were quantified with up to 180 d i for fully developed flows. • Three different criteria for buoyancy and one for flow acceleration were tested. • Four different heat transfer correlations were validated against experimental data. [ABSTRACT FROM AUTHOR]

In this publication, the cooling heat transfer coefficient (h t c) is investigated in a 3 mm diameter tube with vertical flow orientation. Commonly used calculation methods of the heat transfer coefficient are presented. Although developed for heating of sCO 2 , the mixed convection criterion is used to evaluate the heat transfer deterioration. The effects of the CO 2 mass flux of 141 − 354 k g / m ² s and bulk fluid temperatures of 20 − 50 ° C with a constant pressure of 80 b a r on the heat transfer were examined. The upwards flow shows a steady decrease in the h t c with the reduction of the mass flux. However, the downwards flow shows significant effects of buoyancy. At low mass flux the distinct peak in the h t c at the pseudocritical temperature (T p c) disappears. The deteriorated heat transfer in the downwards flow showed significant lower wall temperatures compared to the upwards flow. The separate evaluation of the liquid-like and gas-like region indicates that the deterioration may vary based on the fluid properties. The results were regressed within +-15% in the forced convection region and +-40% in the mixed convection region. The comparison with the literature data of vertical cooling in 6 mm showed different behaviour which indicates an influence of the tube diameter. • Heat transfer of sCO2 depends strongly on the flow direction. • The flow direction influences the heat transfer much stronger at low mass flux. • Heat transfer deterioration leads to a strong decrease in the wall temperature. [ABSTRACT FROM AUTHOR]

The supercritical carbon dioxide (sCO 2 1) heat removal system, which is based on a closed Brayton cycle with sCO 2 as a working fluid, is an innovative heat removal system for existing and future nuclear power plants. This paper provides the design, layout and control of the system based on assumptions developed in the project sCO2-4-NPP. A self-propelling operational readiness state enables a fast start-up and consumes only 12% of the design thermal power input. The system is analysed over a wide range of ambient and steam-side conditions in ATHLET, using performance maps for the turbomachinery, which were designed recently. The performance analysis suggests that it is a good option to operate the system at the design compressor inlet temperature of 55 °C at any boundary condition. With decreasing thermal power input, the rotational speed of the turbomachinery must be decreased to keep the system self-propelling. Moreover, the turbomachinery design with a higher surge margin is preferred. By controlling the compressor inlet temperature via the air mass flow rate and turbine inlet temperature via the turbomachinery speed, the heat removal system is successfully operated in interaction with a pressurized water reactor. • Design, layout and control of a supercritical carbon dioxide heat removal system. • Self-propelling operational readiness enables fast start-up. • Compressor inlet temperature may be kept constant at any boundary condition. • New turbomachinery design with a higher surge margin is preferred. • Successful operation in interaction with a pressurized water reactor. [ABSTRACT FROM AUTHOR]

• Heat transfer of sCO 2 depends strongly on the fluid temperature and pressure. • Cooling water temperature leads to a trade-off in heat transfer performance. • CO 2 mass flux influences the heat transfer much stronger near the pseudocritical temperature. • The wall fluid properties are most appropriate for sCO 2 cooling Nusselt correlation. [Display omitted] The heat transfer of carbon dioxide near the critical point cooled in a horizontal tube with 2 mm inner diameter was investigated experimentally. The local heat transfer coefficients (htc) were determined by temperature measurements inside the tube wall along the 1.2 m cooled length. The effects of the CO 2 mass fluxes in the range of 400–1300 kg/m2s, CO 2 inlet pressures between 77 and 85 bar, CO 2 bulk fluid temperatures between 10 and 85 °C and cooling water temperatures between 10 and 40 °C on the heat transfer were examined. The experimental results are compared with existing correlations for sCO 2 heat transfer in small diameter tubes. A modified Dittus-Boelter equation was developed to predict the heat transfer coefficients with an accuracy of 20%. [ABSTRACT FROM AUTHOR]

Heat pipes (HP) are considered being used in the divertor target because of their high thermal conductivity and their capability to substantially enlarge the heat transfer area to the cooling circuit. Here, a divertor target concept based on a heat pipe design is introduced being able to dissipate heat fluxes of up to 20 MW / m2. It consists of a 230 mm long water-based heat pipe with a capillary structure that combines axial grooves with sintered porous material. The analysis of the integration of the HP in the divertor target indicates that design changes of the current divertor cassette design are required to accommodate such a solution. [ABSTRACT FROM AUTHOR]