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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 991 records · Page 55
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Study on Water-ingress Accident during the Initial Core and the Running-in Phase of HTR-PM - 18429

Water ingress accident is one of the typical accidents with potential hazard to the high temperature gas-cooled reactor (HTGR), which will cause reactivity introduction, as well as the chemical corrosion of the graphite fuel elements and reflector structure material. Besides, increase of the primary pressure may result in the opening of the safety valves, consequently leading the release of radioactive isotopes. Therefore high attention has been paid to water ingress phenomenon during the design of the HTGR to verify its inherent safety characteristics. The online continuous discharge and multi-pass of the fuel elements is another important characteristics for the pebble-bed HTGR which can effectively increase the reactor operation availability. From the initial core, a long period of time, namely the running-in phase, is needed to reach the final long-term stable equilibrium core. During this period of time, the core states are totally different, including the number of the fuel elements, the fuel enrichment, the accumulation of the fission product, the temperature feedback coefficient of the reactivity, and so on. In this paper, based on the design of 250 MW Pebble-bed Module High Temperature gas-cooled Reactor (HTR-PM) and a candidate proposal for its initial core and running-in phase, considering three typical states during the running-in phase, the design basis accident of a double-ended guillotine break of one heating tube has been analyzed with the TINTE code, and the results are compared with that of the equilibrium core. Some important parameters during the transient process, including the power, the temperature of the fuel element, the pressure of the primary coolant, as well as the graphite corrosion rate, are studied. To a certain extent, the results can verity the design of the initial core and the running-in phase, and also further prove the safety feature of the HTR. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

A Methodology for Loading the Advanced Test Reactor Driver Core for Experiment Analysis

The Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) is one of the premier irradiation facilities in the world. The ATR hosts experiments from a variety of laboratories, companies, and disciplines. Because of its flexibility in providing customized irradiation conditions to experiments, such as coolant chemistry, temperature, and reactor power, analysis for experiments, both from a programmatic and safety perspective, can be a challenging endeavor. Because of these experiment variables, there is always an effort to improve upon the existing analysis methodology in order to provide the sponsor of any given experiment with the most accurate predictions possible. The paper discusses the effort to improve upon the methodology for loading the ATR driver core for experiment analysis. Experiment analysis at INL is generally broken up into three separate disciplines: neutronics, thermal-hydraulics, and structural. The focus of this paper is neutronics analysis. After the experiment has a preliminary design, the neutronics analyst will create a model in the Monte Carlo N-Particle (MCNP) transport code to calculate fuel and material heating rates from MCNP tallies and provide those to the thermal-hydraulics analyst, and calculate reactivity worths, perturbation on the axial fission profile of the ATR driver core by the experiment, void worths, and experiment backup worths. Also, there are usually several programmatic parameters that the experiment sponsor requests that the neutronics analyst calculates. These can include heat generation rates, burnup and displacements per atom (DPA). In order to accurately analyze these parameters, an important input into the physics analyst's MCNP model is the loading of the ATR driver core. The ATR driver core consists of 40, 19-plate highly-enriched uranium (HEU) fuel elements, arranged in a serpentine shape, as pictured in Figure 1. The ATR driver core is responsible for the vast majority of the thermal power of the reactor and the nuclear interactions with the experiments being irradiated in ATR. A unique feature of ATR is the 'power tilt' across the reactor. A power tilt means that the Northwest, Northeast, Southwest, and Southeast Lobes of ATR frequently operate at different powers, providing different irradiation conditions for the experiments in and around the lobes of the reactor. In order to achieve lobe power splits, different fuel elements for each of the 40 positions are chosen by ATR Reactor Engineering so that lobe power splits can be maintained during an operating cycle. A typical fuel loading for an ATR cycle consists of 14 to 18 new fuel elements and 2 to 26 depleted fuel elements. These loadings are typically chosen from three different types of fuel elements. All three ATR fuel element types share a common design, spelled out in Chapter 4 of the ATR Safety Analysis Report (SAR). The fuel element meat is HEU UAlx dispersion fuel sandwiched between aluminum cladding with a length of 121.92 cm (4 ft), with 19 plates per fuel element. The three variations of this common design are designated by the first two letters in their serial numbers: XA, YA, and NB. XA elements are most commonly used. These elements have nominally 1075 g of uranium-235 and 0.66 g of boron-10. The NB elements have the same uranium-235 loading as the XA elements, but no boron in any plates. These elements are used when excess reactivity is needed. The YA elements have the same loading as the XA elements, except that the outermost plate (Plate 19) has no boron or uranium. These elements are used to mitigate aged reflector damage. In order to accurately model the ATR driver core, all three types of fuel elements must be accounted for in the MCNP model. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

MOCUM Verification with a Heterogeneous MOX Whole Core C5G7 Benchmark Problem

Recent advancements in computer technology has allowed for the possibility to solve complex reactor problems without homogenization, but the present benchmark problems are all limited to small scale with few fuel assemblies and homogeneity at the core level. Therefore, to verify the capabilities of the advanced neutronics codes, heterogeneous full core model shall be developed. Recently several whole-core benchmark problems such as BEAVRS and VERA are developed with measured data to which computational codes can be compared. With a particular research interest on the MOX fuel, this study develops a two-dimensional (2-D) whole core MOX neutronics benchmark based on the C5G7 benchmark concept. The developed MOX core benchmark will then be used to verify the accuracy of the MOCUM code, which is developed using the method of characteristic (MOC) as the flux solver with an advanced geometry processor designed for arbitrary core geometries. MOCUM results will be compared against MCNP5 results as the reference solutions. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

PARCS Analysis of the APR1400 Initial Core

The Design Control Document (DCD) for the APR1400 describes the core loading for the initial core and presents the results of various nuclear design calculations in Chapter 4. In the current work, the initial core cycle depletion was simulated using the Nuclear Regulatory Commission (NRC) suite of nuclear analysis tools and the results of these calculations were compared to reference results provided in the APR1400 DCD. The regulatory purpose of this work was to confirm the results reported in the DCD through independent analysis. The NRC code suite consists of SCALE/POLARIS, which is used for lattice physics analysis, GENPMAXS, which is a tool that converts the lattice physics output to a library file used by the core simulator, and PARCS/PATHS, which is a nodal core simulator. Calculations were performed that simulate the depletion of the initial core through one cycle. The authors compared the calculation results in terms of the power distributions at various points in the cycle to reference results described in the DCD. In addition to these comparisons, calculations were performed to determine point kinetics parameters that are used in downstream transient calculations using TRACE.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Improved Core Design of the Supercritical Water-cooled Reactor CSR1000

The SCWR (Supercritical water-cooled reactor) is one of the Generation IV reactor concepts. The economic performance of the SCWR has a core focus, and in this study forms one of the key design goals. Based on the technology of ABWR and PWR, a conceptual design of CSR1000 (China Supercritical water-cooled Reactor with the rated electric power of 1000 MWe) core was proposed with features such as single water rod, combined square fuel assembly (FA),two flow-pass of coolant in core, high-leakage reload core with 157 FAs, burn-up of 350 EFPD. The refueling cycle of the CSR1000 design was about 12-months, which is much shorter than the generation 3 PWR of 18-months. Thus, it is necessary to enhance the economic performance of CSR1000 by prolonging the refueling cycle. Based on conceptual design of the CSR1000, the FA and the core design are improved for prolonging the refueling cycle of CSR1000. This paper represents an important reference and technical reserve for further SCWR core design study.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Watts Bar I ex-core analyses using VERA - 25430

The integration and direct coupling of the Oak Ridge National Laboratory (ORNL) Shift Monte Carlo code within the Consortium for Advanced Simulation of Light Water Reactors (CASL) Virtual Environment for Reactor Applications (VERA) offers unique capabilities that combine high-fidelity in-core and ex-core radiation transport. In this recent development activity, the deterministic neutronics code MPACT performs the in-core radiation transport with temperature feedback using COBRA-TF (CTF), and provides the fission source to Shift for a follow-on fixed source radiation transport calculation to determine ex-core quantities of interest. The coupling of MPACT to Shift allows the neutron source for the ex-core transport to be calculated from the time-dependent, fully coupled MPACT solution. This paper presents the first use of VERA to perform coupon irradiation for code-to-code comparison of its ex-core capabilities. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Geographic variations in lowermost mantle structure from the ray parameters and decay constants of core-diffracted waves: CMBR Structure From P diff and S diff

We propose an array-based approach for constraining seismic velocity structure in the lowermost mantle by measuring the frequency dependence of the ray parameter p and decay constant γ for core-diffracted waves (P diff and SH diff ). The approach uses an iterative multichannel cross-correlation algorithm that solves for relative arrival times and amplitudes of core-diffracted waveforms from multiple peaks in normalized correlograms from pairs of coazimuthal stations. The approach is applied to 60 m b ≥ 5.8 earthquakes for P diff and 36 for SH diff during 2001–2002, with nearly 50,000 unique profile station pairs with epicentral-distance differences of Δd ≥ 10° and azimuth differences of Δφ ≤ 10°, sampling a significant portion of the lowermost mantle. Cap-averaging the resulting ray parameter estimates produces geographic variations that are largely consistent with the distribution of lowermost mantle large low-velocity provinces and seismically fast slab-accumulation regions seen in seismic tomography models, whose locations also greatly influence the geographic distribution of heat flow out of the core. Geographic variations in station-pair diffracted-wave decay constants differ from those of the ray parameters, suggesting that variations in the decay of core-diffracted waves are more linked to lowermost mantle seismic velocity gradients than absolute values of seismic velocity. The ray parameters and decay constants of core-diffracted waves are strongly frequency-dependent, and these frequency variations also vary significantly with geographic location. Combining lateral and vertical seismic-velocity variations with mineral-physics data on elasticity and conductivity of lowermost mantle species can provide constraints on D" composition and CMB heat flux.

58 GEOSCIENCES↗
DOE OSTI2017

Fabrication of magnetic core PEI-silica shell particles

Highlights: • PEI-silica shell fabrication was developed for iron-carbonyl core material. • Magnetic separation of core-free by-product of the fabrication proved successful. • Varying fabrication method/concentration successfully tuned shell characteristics. • Sonication methods produced smooth coatings with a trend towards multicore particles. • Raspberry-like coatings were made by the slow addition, one- and two-pot methods. - Abstract: Magnetic core-silica shell particles combine the physico-chemical properties of the silica shell with the magnetic properties of the core. Seven possible particle coating methods utilizing the biomimetic polyamine polyethyleneimine (PEI) as the silication catalyst, have been identified and the objective of this study was to link the properties of the final PEI-silica coating with the methodology. The PEI-silica coated carbonyl iron particles were characterized via scanning electron microscopy, energy dispersive X-ray spectroscopy and Fourier transfer infrared spectroscopy. Varying the fabrication method and silane concentration successfully tuned the shell characteristics. Methods using sonication produced smooth more evenly distributed coatings with a tendency towards multicore particles. In contrast, raspberry-like coatings were produced via the slow addition, one- and two-pot methods. The magnetic separation process after synthesis achieved a degree of purification that makes these core–shell particles useful for potential applications that require specific surface interactions and the ease of magnetic separation.

36 MATERIALS SCIENCE↗
DOE OSTI2019

EGS Collab Experiment 2: Core Logs

Core logs and photos from the EGS Collab project Experiment 2 for the Top Vertical well (TV4100) and the Top Horizontal well (TV 4100) on the 4100 Level of SURF (the Sanford Underground Research Facility). The core logs are stored in a single PDF file with 5-ft run intervals. In the monitoring well IDs, "O" indicates that the well is orthogonal to the anticipated fracture plane, "T" refers to top, and "H" refers to horizontal. A core log CT scan for TV4100 and a layout image of the 4100 wells are included as well. Logs include: experiment number; borehole ID; depth interval; run number; final packed core box number; scribe line (yes/no; red-on-right convention); logging dates; logger initials; as well as sketches of core foliation, folding, and fracturing with additional details and notes on other features of interest. Shift reports include: date, location, personnel, summary of site activity, and field notes.

15 GEOTHERMAL ENERGY↗
DOE OSTI2016

Initial Comparison of Direct and Legacy Modeling Approaches for Radial Core Expansion Analysis

Radial core expansion in sodium-cooled fast reactors provides an important reactivity feedback effect. As the reactor power increases due to normal start up conditions or accident scenarios, the core and surrounding materials heat up, causing both grid plate expansion and bowing of the assembly ducts. When the core restraint system is designed correctly, the resulting structural deformations introduce negative reactivity which decreases the reactor power. Historically, an indirect procedure has been used to estimate the reactivity feedback due to structural deformation which relies upon perturbation theory and coupling legacy physics codes with limited geometry capabilities. With advancements in modeling and simulation, radial core expansion phenomena can now be modeled directly, providing an assessment of the accuracy of the reactivity feedback coefficients generated by indirect legacy methods. Recently a new capability was added to the PROTEUS-SN unstructured geometry neutron transport solver to analyze deformed meshes quickly and directly. By supplying the deformed mesh in addition to the base configuration input files, PROTEUS-SN automatically processes material adjustments including calculation of region densities to conserve mass, calculation of isotopic densities according to material models (for example, sodium density as a function of temperature), and subsequent re-homogenization of materials. To verify the new capability of directly simulating deformed meshes, PROTEUS-SN was used to compute reactivity feedback for a series of contrived yet representative deformed configurations for the Advanced Burner Test Reactor design. The indirect legacy procedure was also performed to generate reactivity feedback coefficients for the same deformed configurations. Interestingly, the legacy procedure consistently overestimated reactivity feedbacks by 35% compared to direct simulations by PROTEUS-SN. This overestimation indicates that the legacy procedures are in fact not conservative and could be overestimating reactivity feedback effects that are closely tied to reactor safety. We conclude that there is indeed value in performing direct simulation of deformed meshes despite the increased computational expense. PROTEUS-SN is already part of the SHARP multi-physics toolkit where both thermal hydraulics and structural mechanical feedback modeling can be applied but this is the first comparison of direct simulation to legacy techniques for radial core expansion.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Development of integrated core disruptive accident analysis code for FBR - ASTERIA-FBR

The evaluation of consequence at the severe accident is the most important as a safety licensing issue for the reactor core of liquid metal cooled fast breeder reactor (LMFBR), since the LMFBR core is not in an optimum condition from the viewpoint of reactivity. This characteristics might induce a super-prompt criticality due to the core geometry change during the core disruptive accident (CDA). The previous CDA analysis codes have been modeled in plural phases dependent on the mechanism driving a super-prompt criticality. Then, the following event is calculated by connecting different codes. This scheme, however, should introduce uncertainty and/or arbitrary to calculation results. To resolve the issues and obtain the consistent calculation results without arbitrary, JNES is developing the ASTERIA-FBR code for the purpose of providing the cross-check analysis code, which is another required scheme to confirm the validity of the evaluation results prepared by applicants, in the safety licensing procedure of the planned high performance core of Monju. ASTERIA-FBR consists of the three major calculation modules, CONCORD, dynamic-GMVP, and FEMAXI-FBR. CONCORD is a three-dimensional thermal-hydraulics calculation module with multi-phase, multi-component, and multi-velocity field model. Dynamic-GMVP is a space-time neutronics calculation module. FEMAXI-FBR calculates the fuel pellet deformation behavior and fuel pin failure behavior. This paper describes the needs of ASTERIA-FBR development, major module outlines, and the model validation status. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Utah FORGE: Powder X-ray Diffraction Data from Well 16A(78)-32 Core

This dataset from Lawrence Livermore National Laboratory (LLNL) consists of four raw X-ray diffraction (XRD) scans and preliminary results of quantitative XRD analysis. The scanned samples were prepared from four subcores, which came from various depths of the FORGE well 16A(78)-32 core. Desired core lengths were selected from available core photos (on GDR), provided by FORGE personnel, and subcored at LLNL. The XRD scans were collected in May 2023 at LLNL as pre-experimental characterization data for these subcores, which will be used in core-flooding experiments at LLNL and in triaxial direct shear experiments at Los Alamos National Laboratory as part of DOE Project 5-2428. XRD scans are in RAW file format (e.g., FORGE-5477-full.raw) and are suitable for viewing and analysis using open-source quantitative XRD software (e.g., Profex; www.profex-xrd.org ) and/or other proprietary instrument software.

15 GEOTHERMAL ENERGY↗
DOE OSTI2018

Multistage Core Formation in Planetesimals Revealed by Numerical Modeling and Hf-W Chronometry of Iron Meteorites

Iron meteorites provide some of the most direct insights into the processes and timescales of core formation in planetesimals. Of these, group IVB irons stand out by having one of the youngest 182 Hf- 182 W model ages for metal segregation (2.9 ± 0.6 Ma after solar system formation), as well as the lowest bulk sulfur content and hence highest liquidus temperature. Here in this paper, using a new model for the internal evolution of the IVB parent body, we show that a single stage of metal-silicate separation cannot account for the complete melting of pure Fe metal at the relatively late time given by the Hf-W model age. Instead, a complex metal-silicate separation scenario is required that includes migration of partial silicate melts, formation of a shallow magma ocean, and core formation in two distinct stages of metal segregation. In the first stage, a protocore formed at ≈1.5 Ma via settling of metal particles in a mantle magma ocean, followed by metal segregation from a shallow magma ocean at ≈5.4 Ma. As these stages of metal segregation occurred at different times, the two metal fractions had different 182 W compositions. Consequently, the final 182W composition of the IVB core does not correspond to a single differentiation event, but represents the average composition of early- and late-segregated core fractions. Our best fit model indicates an ≈100 km radius for the IVB parent body and provides an accretion age of ≈0.1–0.5 Ma after solar system formation. The computed solidification time is, furthermore, consistent with the Re-Os age for crystallization of the IVB core.

58 GEOSCIENCES↗
DOE OSTI2024

Engineering Ultrathin Alloy Shell in Au@AuPd Core-Shell Nanoparticles for Efficient Plasmon-Driven Photocatalysis

Bimetallic core-shell nanoparticles (NPs) possessing a synergetic coupling of plasmonic and catalytic metals have emerged as promising prototypes for efficient plasmon-driven photocatalysis. Here, Au@AuPd core-shell NPs composed of Au core NP and ultrathin AuPd shell are introduced to acquire improved light utilization efficiency in photocatalytic selective oxidation. With systematical control of the composition of the ultrathin alloy shell, it reveals that the Au@AuPd core-shell NPs with 10 at% Pd (in the single-atom alloy regime) is an effective nanostructure capable of maximizing the quantum yield of plasmon-induced light absorption and optimizing the surface electronic structure for catalytic reactions. This controlled system provides new insights into shell engineering for enhancing photocatalytic performance via the regulation of energy funneling processes in core-shell nanocatalysts.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Synthesis of Core@Shell Cu-Ni@Pt-Cu Nano-Octahedra and Their Improved MOR Activity

Fabrication of 3d metal-based core@shell nanocatalysts with engineered Pt-surfaces provides an effective approach for improving the catalytic performance. The challenges in such preparation include shape control of the 3d metallic cores and thickness control of the Pt-based shells. Herein, we report a colloidal seed-mediated method to prepare octahedral CuNi@Pt-Cu core@shell nanocrystals using CuNi octahedral cores as the template. By precisely controlling the synthesis conditions including the deposition rate and diffusion rate of the shell-formation through tuning the capping ligand, reaction temperature, and heating rate, uniform Pt-based shells can be achieved with a thickness of < 1 nm. Furthermore, the resultant carbon-supported CuNi@Pt-Cu core@shell nano-octahedra showed superior activity in electrochemical methanol oxidation reaction (MOR) compared with the commercial Pt/C catalysts and carbon-supported CuNi@Pt-Cu nano-polyhedron counterparts, demonstrating that both the lattice strain and shape effects arising from the as-prepared CuNi@Pt-Cu octahedral nanocatalysts play a crucial role in MOR performance.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Probing Nitrogen-Doping Effects in the Core-Shell Structured Catalysts for Bifunctional Electrocatalysis.

The design of core-shell structured catalysts has attracted wide interests due to their remarkable catalytic performances in many fields. Although the nitrogen doping is often integrated in such kind of catalysts, the main contribution is exclusively attributed to the metal sites or the core-shell configuration. The role of nitrogen doping is often believed to be less important, which remains largely unexplored. Here, in an effort to probe the catalytic role of heterogeneous doping in such core-shell structure, the effect of nitrogen doping on the activation of cobalt and iron carbide-based materials for the electrocatalytic reduction of oxygen and the hydrogen evolution reaction is originally revealed. The nitrogen doping in the core-shell structured biphasic interfaces is found to be of critical importance for triggering the electrocatalytic activity and selectivity. Therefore, this study shows the neglected but critical role of nitrogen doping and biphasic interactions in a core-shell configuration, paving a new pathway towards the performance enhancement.

biphasic interfaces↗
DOE OSTI2020

Assessing mean climate change signals in the global CORDEX-CORE ensemble

The new Coordinated Output for Regional Evaluations (CORDEX-CORE) ensemble provides high-resolution, consistent regional climate change projections for the major inhabited areas of the world. It serves as a solid scientific basis for further research related to vulnerability, impact, adaptation and climate services in addition to existing CORDEX simulations. The aim of this study is to investigate and document the climate change information provided by the CORDEX-CORE simulation ensemble, as a part of the World Climate Research Programme (WCRP) CORDEX community. An overview of the annual and monthly mean climate change information in selected regions in different CORDEX domains is presented for temperature and precipitation, providing the foundation for detailed follow-up studies and applications. Initially, two regional climate models (RCMs), REMO and RegCM were used to downscale global climate model output. The driving simulations by AR5 global climate models (AR5-GCMs) were selected to cover the spread of high, medium, and low equilibrium climate sensitivity at a global scale. The CORDEX-CORE ensemble has doubled the spatial resolution compared to the previously existing CORDEX simulations in most of the regions (25 $\,\mathrm {km}$ (0.22 $^{\circ }$ /jats:inline-formula>) versus 50 $\,\mathrm {km}$ (0.44 $^{\circ }$ )) leading to a potentially improved representation of, e.g., physical processes in the RCMs. The analysis focuses on changes in the IPCC physical climate reference regions. The results show a general reasonable representation of the spread of the temperature and precipitation climate change signals of the AR5-GCMs by the CORDEX-CORE simulations in the investigated regions in all CORDEX domains by mostly covering the AR5 interquartile range of climate change signals. The simulated CORDEX-CORE monthly climate change signals mostly follow the AR5-GCMs, although for specific regions they show a different change in the course of the year compared to the AR5-GCMs, especially for RCP8.5, which needs to be investigated further in region specific process studies.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2021

Climate hazard indices projections based on CORDEX-CORE, CMIP5 and CMIP6 ensemble

The CORDEX-CORE initiative was developed with the aim of producing homogeneous regional climate model (RCM) projections over domains world wide. In its first phase, two RCMs were run at 0.22° resolution downscaling 3 global climate models (GCMs) from the CMIP5 program for 9 CORDEX domains and two climate scenarios, the RCP2.6 and RCP8.5. The CORDEX-CORE simulations along with the CMIP5 GCM ensemble and the most recently produced CMIP6 GCM ensemble are analyzed, with focus on several temperature, heat, wet and dry hazard indicators for present day and mid-century and far future time slices. The CORDEX-CORE ensemble shows a better performance than the driving GCMs for several hazard indices due to its higher spatial resolution. For the far future time slice the 3 ensembles project an increase in all temperature and heat indices analyzed under the RCP8.5 scenario. The largest increases are always shown by the CMIP6 ensemble, except for Tx > 35 °C, for which the CORDEX-CORE projects higher warming. Extreme wet and flood prone maxima are projected to increase by the RCM ensemble over the la Plata basin in South America, the Congo basin in Africa, east North America, north east Europe, India and Indochina, regions where a better performance is obtained, whereas the GCM ensembles show small or negligible signals. Furthermore, compound hazard hotspots based on heat, drought and wet indicators are detected in each continent worldwide in region like Central America, the Amazon, the Mediterranean, South Africa and Australia, where a linear relation is shown between the heatwave and drought change signal, and region like Arabian peninsula, the central and south east Africa region (SEAF), the north west America (NWN), south east Asia, India, China and central and northern European regions (WCE, NEU) where the same linear relation is found for extreme precipitation and HW increases. Although still limited, the CORDEX-CORE initiative was able to produce high resolution climate projections with almost global coverage and can provide an important resource for impact assessment and climate service activities.

54 ENVIRONMENTAL SCIENCES↗