Graphite Blocks Research Articles

Reference thermal neutron field at KRISS for calibration of neutron detectors

A reference thermal neutron field has been established at the Korea Research Institute of Standards and Science (KRISS) by using a 241Am-Be neutron source and a high-purity graphite pile constructed by stacking graphite blocks. The properties of the graphite blocks such as impurities, density, and dimensions were studied thoroughly to understand the characteristics of the generated field. The energy spectrum and thermal neutron fractions were simulated with the Monte Carlo N-Particle eXtended code using measured physical parameters. The neutron effective temperature was 308 K, and the fraction of thermal neutrons was approximately 95% at the reference position of the thermal neutron field. The thermal neutron fluence rate was determined by adopting the Westcott convention method based on neutron activation analysis using a gold foil. The Westcott fluence rate for thermal neutron at the reference position was (2326.7 ± 8.4) cm−2s−1. The true thermal neutron fluence rate at the reference position in the KRISS thermal neutron field was (2700 ± 29) cm−2s−1 (at the reference date of June 30, 2014). The response of a spherical proportional counter with He-3 (SP9 neutron detector) was evaluated in the newly established field as (3.083 ± 0.045) cm2 for the reference calibration condition (a parallel neutron beam with a Maxwellian energy distribution having a most probable energy of 0.025 eV).

Investigation on local cooling in reducing freckles for directionally solidified superalloy specimens with abruptly varying cross-sections

In order to suppress the freckle formation in specimens with abruptly varying cross-sections, superalloy CMSX-4 was directionally solidified, and the local cooling method was investigated, in which a graphite block was inserted into the bottom of platform of specimen. The experimental and simulation results were coupled to evaluate this method in terms of freckling number, primary dendrite arm spacing (PDAS), dendrite morphology, and solidification interface shape. The results show that, with the application of local cooling method, freckles appearing in the specimens were absolutely removed. Dendrite microstructure was refined with reduced PDAS, and a dendritic to cellular transition was promoted. In addition, the concave solidification interface was flattened. Therefore, the local cooling method with the application of graphite block can be employed to reduce the freckling tendency in directionally solidified CMSX-4 superalloy specimens with abruptly varying cross-sections.

DEVELOPMENT OF ISOTOPICALLY ENRICHED BORON-DOPED ALUMINA DOSIMETER FOR THERMAL NEUTRONS.

A novel optically stimulated luminescence (OSL) detector containing isotopically enriched boron was developed for thermal neutron dosimetry. Alumina containing isotopically enriched boron (Al2O3:B) was synthesised by the sol-gel method. The Al2O3:B was annealed up to ~1800 K. For X-ray diffractometer (XRD) analysis, the diffraction pattern of the Al2O3:B had reflex peaks corresponding to α-Al2O3. The sensitivity of Al2O3:B to photons was slightly 2% of that of a commercial Al2O3:C. The Al2O3:B detector had satisfactory linearity in X-ray dose measurement. A thermal neutron field was constructed using a 241Am-Be neutron source and graphite blocks. A pair of Al2O3:10B and Al2O3:11B detectors were set in the thermal neutron field. The response of Al2O3:10B was larger than that of Al2O3:11B owing to the 10B(n,α)7Li reactions. The sensitivity of Al2O3:10B to thermal neutrons was estimated to be two orders less than the photon sensitivity. Therefore, the pair of Al2O3:10B and Al2O3:11B detectors were useful for thermal neutron dosimetry.

INFLUENCE OF CARBON BASED ELECTRODES ON THE PERFORMANCE OF THE MICROBIAL FUEL CELL

In this study electricity generation was evaluated in a two chambered microbial fuel cell. Performance of microbial fuel cells using two bacteria, Klebsiella pneumoniae and Bacillus coagulans and using three different electrodes namely graphite blocks, carbon cloth and graphite sheet was studied. The device was operated under anaerobic condition in the anode chamber and parameters were recorded for a period of 48 hours. The performance of MFC was analyzed by the measurement of open circuit voltage, polarization curves, impedance curves and cyclic voltammetry. Among different combinations of electrode tested, carbon cloth electrode produced high power density (80 mW/m2). Graphite block gave much high power compared to sheet. Finally, performance was compared with Shewanellaputrefaciens. The current study explores the applicability of carbon electrode for MFC applications.

Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Core design and safety analysis

Engineering demonstration reactors are nuclear reactors built to establish proof of concept for technology options that have never been built. Examples of engineering demonstration reactors include Peach Bottom 1 for high temperature gas-cooled reactors (HTGRs) and Experimental Breeder Reactor II (EBR-II) for sodium-cooled fast reactors. Engineering demonstrations have historically played a vital role in advancing the technology readiness level of reactor technologies. This paper details a preconceptual design for a fluoride salt-cooled engineering demonstration reactor. The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would use tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 7LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable.Core design characteristics, fuel cycle performance, and safety analysis of the FHR DR preconcept have been evaluated. The FHR DR core design features a negative or negligible void coefficient throughout a reactor operating cycle. Both single-batch (cartridge) and multiple-batch fuel cycles can be demonstrated in the FHR DR. The single-batch cycle length of the FHR DR core is estimated at between 12 and 18months, assuming the successful qualification of composite carbon (C/C) or silicon carbide (SiC/SiC) structural fuel block tie rod material. Fuel cycle performance of the FHR DR is similar to a high temperature gas-cooled reactor. Preliminary safety analysis of the FHR DR indicates that the reactor could be used to demonstrate the inherent safety characteristics of FHR designs.

О выборе критерия прочности при математическом моделировании поведения реакторного графита

During the operation of Type RBMK-1000 nuclear power-generating reactors (high-power pressure-tube reactors), it was decided to extend the lifetime of this type of reactors beyond the initially assigned one. Such an approach is more economically profitable than the commissioning of new power units. To ensure the optimal operating conditions and safe operation of the reactor, it is necessary, apart from carrying out its technical modernization, to substantiate the acceptable strength and shape distortion of graphite blocks after their having been exposed to long-term neutron irradiation by means of computer modeling. The graphite blocks used in high-power pressure-tube (RBMK-type) reactors are elements that in many respects determine the reactor’s lifetime as a whole due to their limited amenability to repair. In view of this circumstance, a task was set forth to develop a computer model of thermo-irradiation behavior of graphite blocks that would take into account the maximum possible number of parameters affecting the thermomechanical characteristics of the graphite block. The anisotropy of graphite was taken into account in the stressed-and-strain state equations. Part of the study is devoted to a detailed analysis of creep effect on the graphite block behavior. The strength calculation was carried out taking into consideration the crack incipience and growth phenomena. A few fracture criteria that are used for the conditions of combined stressed-and-strain state are applied in the article for carrying out a strength calculation of the graphite block as an element of the graphite brickwork, which serves as a moderator and deflector in the RBMK-1000 nuclear reactor. The aim of the investigation is to estimate the influence of different fracture criteria on the results obtained from modeling the thermomechanical behavior of a graphite block subjected to long-term neutron irradiation carried out using the finite element method. A 20-node isoparametric element was used as a single finite element.

Controlled Preparation and Thermal Conductivity of Highly Oriented Graphite Blocks

Controlled Preparation and Thermal Conductivity of Highly Oriented Graphite Blocks

Preconceptual design of a fluoride high temperature salt-cooled engineering demonstration reactor: Motivation and overview

Engineering demonstration reactors are nuclear reactors built to establish proof of concept for technology options that have never been built. Examples of engineering demonstration reactors include Peach Bottom 1 for high temperature gas-cooled reactors and the Experimental Breeder Reactor-II for sodium-cooled fast reactors. Engineering demonstrations have historically played a vital role in advancing the technology readiness level of reactor concepts. This paper details a preconceptual design for a fluoride salt-cooled engineering demonstration reactor. The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output. It would use tristructural-isotropic (TRISO) particle fuel in compacts within prismatic graphite blocks. FLiBe (2 7LiF-BeF2) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is an intermediate step to enable near-term commercial FHRs. The design philosophy of the FHR DR was focused on safety, near-term deployment, and flexibility. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated as an engineering demonstration with minimal risk and cost. These technologies include TRISO particle fuel, replaceable core structures, and consistent structural material selection for core structures and the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers.Important capabilities to be demonstrated by building and operating the FHR DR include:•core design methodologies,•heat exchanger performance (including passive decay heat removal),•pump performance,•reactivity control,•salt chemistry control to maximize plant life,•salt procurement, handling, maintenance and ultimate disposal, and•tritium management.Non-nuclear separate and integral test efforts (e.g., heated salt loops or loops using simulant fluids) are necessary to develop the technologies that will be demonstrated in the FHR DR.

Application of 3D coupled code ATHLET-QUABOX/CUBBOX for RBMK-1000 transients after graphite block modernization

Abstract This work describes the application and the results of transient calculations for the RBMK-1000 with the coupled code system ATHLET 2.2A-QUABOX/CUBBOX which was developed in GRS. Within these studies the planned modernization of the graphite blocks of the RBMK-1000 reactor is taken into account. During the long-term operation of the uranium-graphite reactors RBMK-1000, a change of physical and mechanical properties of the reactor graphite blocks is observed due to the impact of radiation and temperature effects. These have led to a deformation of the reactor graphite columns and, as a result, a deformation of the control and protection system (CPS) and of fuel channels. Potentially, this deformation can lead to problems affecting the smooth movement of the control rods in the CPS channels and problems during the loading and unloading of fuel assemblies. The present paper analyzes two reactivity insertion transients, each taking into account three graphite removal scenarios. The presented work is directly connected with the modernization program of the RBMK-1000 reactors and has an important contribution to the assessment of the safety-relevant parameters after the modification of the core graphite blocks.

Nuclear Heat Supply Fluctuation Tests by Non-Nuclear Heating With HTTR

The nuclear heat utilization systems connected to high-temperature gas-cooled reactors (HTGRs) will be designed on the basis of non-nuclear-grade standards in terms of easier entry for the chemical plant companies and the construction economics of the systems. Therefore, it is necessary that the reactor operations can be continued even if abnormal events occur in the systems. The Japan Atomic Energy Agency has developed a calculation code to evaluate the absorption of thermal-load fluctuations by the reactors when the reactor operations are continued after such events, and has improved the code based on the high-temperature engineering test reactor (HTTR) operating data. However, there were insufficient data on the transient temperature behavior of the metallic components and the graphite core support structures corresponding to the fluctuation of the reactor inlet coolant temperature for further improvement of the code. Thus, nuclear heat supply fluctuation tests with the HTTR were carried out in non-nuclear heating operation to focus on the thermal effect. In the tests, the coolant helium gas temperature was heated to 120°C by the compression heat of the gas circulators in the HTTR, and a sufficiently large fluctuation of 17°C for the reactor inlet coolant was achieved by devising a new test procedure under the ideal condition without the effect of the nuclear power. Then, the temperature responses of the metallic components and the graphite core support structures were investigated. The test results adequately showed as predicted that the temperature responses of the metallic components are faster than those of the graphite blocks, and the mechanism of the thermal-load fluctuation absorption by the metallic components was clarified.

Thermal-Hydraulic Analyses of Transportable Fluoride Salt–Cooled High-Temperature Reactor with CFD Modeling

The technology for the 20-MW(thermal) Transportable Fluoride Salt–Cooled High-Temperature Reactor (TFHR) is proposed by Massachusetts Institute of Technology for off-grid applications such as Antarctic bases and remote mining sites. The preliminary thermal-hydraulic analyses and improvements based on a 1/12th full-core model were performed using three-dimensional computational fluid dynamics (CFD). A benchmark study was conducted by comparing the CFD results against empirical correlations and experimental data obtained by Cooke, Silverman, and Grele. In the 1/12th full-core analysis, three practical considerations that may challenge the TFHR temperature limits are evaluated as bounding analysis. These include (1) helium gap between fuel compact and graphite block, (2) thermal conductivity degradations of graphite matrix due to neutron irradiation, and (3) full-core scale power distribution obtained from neutronic calculations. These design considerations lead to insufficient margin between the normal operating condition and the predefined thermal limits. In this context, additional design features are implemented to improve the thermal-hydraulic safety of the TFHR. First, bypass flow in the interstitial gaps between the active core and the reflector is found capable of reducing the temperature peaks at the core periphery. Second, improvements of the flow distribution from the central downcomer to individual coolant channels enable a higher mass flow rate to the regions with compromised cooling access. Overall, thermal-hydraulic performance was significantly improved with a fuel temperature margin from 10 to 150 K and a coolant temperature margin from 16 to 160 K, as well as the more uniform temperature distribution across the reactor core. Furthermore, thermal-hydraulic safety can be maintained at a 20% overpower operating condition [i.e., 24 MW(thermal)]. Overall, this study provides an engineering basis for the TFHR thermal-hydraulic design to improve its safety margin.

Preparation of Hydrophilic Carbon Nano-Particles and Its Application to the Counter Electrodes of DSSCs

Hydrophilic carbon nano-particles have been prepared from a synthesized graphite block using electro-oxidation method in pure water. They are sub-micro meter in diameter depending on the electrolysis condition. They are used as the activator of lead-acid batteries and solid lubricants in non-oily lubricants as already reported at PRiME2012. In this presentation, we will introduce the application to the counter electrodes (CEs) of dye-sensitized solar cells (DSSCs) instead of Pt. We will report the preparation method, the results in characterization, and applications of hydrophilic carbon nano-particles especially on CEs in DSSCs. The weight % of solid carbonous material is 0.4 by the conventional electrolysis method around 10 V DC for 45 days. Increasing in the ratio of the graphite to the cell volume, the electrolysis time decreased and the weight % becomes 0.6 within 3 weeks. By XRD analysis, the dried powder of the produced material consists of mellitic acid (benzenehexacarboxylic acid) and graphite. CE was prepared by doctor blading a mixture of hydrophilic carbon (HC) nano-particles with TiO2 colloids on a FTO glass substrate. Thereafter, HC-CE was characterized using X-ray diffraction, profilometry, four-point probe testing, and CV. A 2 \U0001d707m thick HC-CE revealed a comparable catalytic activity to that of the Pt electrode under the same experimental conditions. A DSSC based on HC-CE was analyzed and showed \U0001d43dsc of 6.87 mA/cm2 close to that of DSSC with Pt-CE (7.0 mA/cm2). More importantly, DSSC based on HC-CE yielded a power-conversion efficiency (\U0001d702) of 2.93% under AM 1.5 irradiation (100 mW/cm2), which was comparable to that of DSSC based on conventional Pt-CE.Hydrophilic carbon nano-particles have been prepared from a synthesized graphite block using electro-oxidation method in pure water. They are sub-micro meter in diameter depending on the electrolysis condition. They are used as the activator of lead-acid batteries and solid lubricants in non-oily lubricants as already reported at PRiME2012. In this presentation, we will introduce the application to the counter electrodes (CEs) of dye-sensitized solar cells (DSSCs) instead of Pt. We will report the preparation method, the results in characterization, and applications of hydrophilic carbon nano-particles especially on CEs in DSSCs. The weight % of solid carbonous material is 0.4 by the conventional electrolysis method around 10 V DC for 45 days. Increasing in the ratio of the graphite to the cell volume, the electrolysis time decreased and the weight % becomes 0.6 within 3 weeks. By XRD analysis, the dried powder of the produced material consists of mellitic acid (benzenehexacarboxylic acid) and graphite. CE was prepared by doctor blading a hydrophilic carbon (HC) nano-particle on a FTO glass substrate. Thereafter, HC-CE was characterized using X-ray diffraction, profilometry, four-point probe testing, and CV. A 2 \U0001d707m thick HC-CE revealed a comparable catalytic activity to that of the Pt electrode under the same experimental conditions. A DSSC based on HC-CE was analyzed and showed \U0001d43dsc of 6.87 mA/cm2 close to that of DSSC with Pt-CE (7.0 mA/cm2). More importantly, DSSC based on HC-CE yielded a power-conversion efficiency (\U0001d702) of 2.93% under AM 1.5 irradiation (100 mW/cm2), which was comparable to that of DSSC based on conventional Pt-CE.

Polynanocrystalline Graphite: A New Carbon Anode with Superior Cycling Performance for K-Ion Batteries.

We synthesized a new type of carbon-polynanocrystalline graphite-by chemical vapor deposition on a nanoporous graphenic carbon as an epitaxial template. This carbon is composed of nanodomains being highly graphitic along c-axis and very graphenic along ab plane directions, where the nanodomains are randomly packed to form micron-sized particles, thus forming a polynanocrystalline structure. The polynanocrystalline graphite is very unique, structurally different from low-dimensional nanocrystalline carbon materials, e.g., fullerenes, carbon nanotubes, and graphene, nanoporous carbon, amorphous carbon and graphite, where it has a relatively low specific surface area of 91 m2/g as well as a low Archimedes density of 0.92 g/cm3. The structure is essentially hollow to a certain extent with randomly arranged nanosized graphite building blocks. This novel structure with disorder at nanometric scales but strict order at atomic scales enables substantially superior long-term cycling life for K-ion storage as an anode, where it exhibits 50% capacity retention over 240 cycles, whereas for graphite, it is only 6% retention over 140 cycles.

Vessel eddy current characteristics in SST-1 tokamak

Eddy current distribution in the vacuum vessel of the Steady state superconducting (SST-1) tokamak has been determined from the experimental data obtained using an array of internal voltage loops (flux loop) installed inside the vacuum vessel. A simple circuit model has been employed. The model takes into account the geometric and constructional features of SST-1 vacuum vessel. SST-1 vacuum vessel is a modified ‘D’ shaped vessel having major axis of 1.285m and minor axis of 0.81m and has been manufactured from non-magnetic stainless steel. The Plasma facing components installed inside the vacuum vessel are graphite blocks mounted on Copper Chromium Zirconium (CuCrZr) heat sink plates on inconel supports. During discharge of the central solenoid, eddy currents get generated in the vacuum vessel and passive supports on it. These eddy currents influence the early magnetic NULL dynamics and plasma break-down and start-up characteristics. The computed results obtained from the model have been benchmarked against experimental data obtained in large number of SST-1 plasma shots. The results are in good agreement. Once bench marked, the calculated eddy current based on flux loop signal and circuit equation model has been extended to the reconstruction of the overall B- field contours of SST-1 tokamak in the vessel region. A comparison of the field lines with and without the plasma column in identical conditions of the central solenoid and equilibrium field profiles has also been done with an aim to quantify the diagnostics responses in vacuum shots.

Characterization of TiC and α-Ti coating on graphite prepared by Powder Immersion Reaction Assisted Coating (PIRAC) Oxidized at 1000°C in Air

Study of the formation of TiC and α-Ti coating on graphite has been successfully carried out by means PIRAC method. Graphite blocks were immersed into a mixed of titanium powder containing 4 wt% iodine and then proceed at 1000°C for 10 hours in argon atmosphere. Two different treatment of graphite blocks were carried out; with and without dispersion ZrO2 in the graphite surface before heat treatment. Characterizations of coated surface were include X-ray diffraction for phase and crystalline analysis, SEM and EDX analysis to determine the topography and distribution of elements in coated film. For sample without ZrO2, as a result of interaction between titanium and graphite, carbon diffusion occurred and TiC coated film were formed. Whereas sample with the dispersant of ZrO2 on the graphite surface caused diffusion of carbon into the titanium did not occur, as a result α-Ti were formed. Oxidation test at 1000°C in air showed that TiC and α-Ti coated film underwent oxidation. In both cases TiO2 was observed as a results of oxidation.

New thermal neutron calibration channel at LNMRI/IRD

A new standard thermal neutron flux unit was designed in the National Ionizing Radiation Metrology Laboratory (LNMRI) for calibration of neutron detectors. Fluence is achieved by moderation of four 241Am-Be sources with 0.6 TBq each, in a facility built with graphite and paraffin blocks. The study was divided into two stages. First, simulations were performed using MCNPX code in different geometric arrangements, seeking the best performance in terms of fluence and their uncertainties. Last, the system was assembled based on the results obtained on the simulations. The simulation results indicate quasi-homogeneous fluence in the central chamber and H*(10) at 50 cm from the front face with the polyethylene filter.

Decolorization and degradation of monoazo and diazo dyes in Pseudomonas catalyzed microbial fuel cell

A laboratory‐scale H‐shaped Pseudomonas catalyzed microbial fuel cell (MFC) was investigated for its performance in decolourizing synthetic wastewater containing azo dyes. The azo‐dyes investigated in this study were methyl orange (MO), congo red, reactive blue 172 (RB), reactive yellow 145, and reactive red 2. Among the azo dyes in anode chamber MO resulted in the highest power density (4100 μW m−2) with graphite electrodes and a decolourization efficiency of 94%. The azo bonds were cleaved in all the dyes tested, and their metabolites in the anolytes were characterized by UV‐visible spectral and HPLC analyses. To reduce the internal resistance of MFC and maximize the power density, different metal salt doped graphite epoxy composites) were tested. In the case of azo dyes MO and RB, the power output increased substantially (almost 1.2 fold) when using Mn2+—GECE compared to graphite block. © 2016 American Institute of Chemical Engineers Environ Prog, 35: 1623–1628, 2016

Synthesis by Vacuum Infiltration, Microstructure, and Thermo‐Physical Properties of Graphite–Aluminum Composite

Graphite prefabricated blocks are fabricated using natural flake graphite and pore‐forming agent of urea, and then high volume fraction graphite–aluminum (Gr–Al) composites are fabricated applying graphite prefabricated blocks and aluminum alloy (ZL101) by vacuum infiltration process for the electronic packaging materials. The microstructures of graphite prefabricated block and Gr–Al composites are observed in detail, which shows that the graphite maintains a good integrity and arranges uniformly in aluminum alloy. Besides, the result shows the formation of a tightly adhered interface between the side surfaces of the graphite and aluminum alloy matrix, and no detrimental Al4C3 phase is found. Thermal properties of Gr–Al composites are researched perpendicularly to the pressing direction. Thermal conductivity increases from 282 to 398 W mK−1 whereas the volume fraction of flake graphite increases from 40 to 75% in the composites, this reveals that the vacuum infiltration is a very promising method to synthesize Gr–Al composites.

Modelling of the spatial distribution of the induced activities in the RBMK-1500 reactor graphite blocks and rings/sleeves

There is one nuclear power plant (NPP) in Lithuania – the Ignalina NPP – which is under decommissioning now. The Ignalina NPP has two units with RBMK-1500 reactors, which are the most powerful and the most advanced versions of RBMK-type reactor design. Unit 1 of the Ignalina NPP was shut down at the end of 2004 and Unit 2 was shut down at the end of 2009. RBMK is a water-cooled graphite-moderated channel-type power reactor and the decommissioning of these reactors faces specific challenges for proper characterisation and disposal of irradiated reactor graphite.Apart from radiological inventory, the spatial distribution of radionuclides in the reactor graphite is also very important because it could indicate the possibilities for decontamination/treatment of the irradiated graphite. This is important for consideration of the near surface disposal option for irradiated graphite, as without treatment it usually does not meet the waste acceptance criteria.Based on that, the work presented in this paper is focused on the modelling of the induced activity spatial distribution in the Ignalina NPP RBMK-1500 reactor graphite components: blocks and rings/sleeves. The modelling was performed with MCNP and SCALE computer codes and consisted of two mains stages: modelling of the neutron flux in the reactor graphite components, and then modelling of the neutron activation in them using the already modelled neutron flux. In such a way, the spatial induced activity distribution in the analysed reactor components was obtained. Modelling results show that the thermal neutron flux is more intensive in the outer radial regions of the graphite components and this, in general, results in higher induced activities there.

Enhancement of cross-plane thermal conductivity and mechanical strength via vertical aligned carbon nanotube@graphite architecture

A three-dimensional (3D) carbon nanotube/exfoliated graphite block (CNT/EGB) was prepared by growing vertical aligned carbon nanotube (VACNT) at the surface of SiO2-coated exfoliated graphite plate (EGP) through chemical vapor deposition followed by hot-pressing. In such 3D CNT/EGB, EGPs were bridged by the VACNTs in the cross-plane direction, and the interface between EGPs and VACNTs was covalently bonded by SiC which formed by reaction of SiO2 and the adjacent carbon of EGPs and VACNTs. The length and growth density of VACNTs were adjusted by the growth time and concentration of catalysts. Thermal conductivity and mechanical strength of CNT/EGB were controlled by the growth states of VACNTs and hot-pressing. CNT/EGB showed a maximum cross-plane thermal conductivity (k⊥) of 38 W/mK, which is more than twice as much as that of EGB (14 W/mK). A remarkable increase in k⊥ was attributed to the efficient heat flow of VACNTs bridging EGPs in the cross-plane direction and the thermal conductive SiC interface between VACNTs and EGPs. Additionally, the increased bending (76 MPa) and compressive strength (59 MPa) of CNT/EGB was due to the combination of strong pull-out effect of high-density nanotubes and the strong covalent interconnections between VACNTs and EGPs.