Tristructural Isotropic Research Articles

Detection and analysis of particles with failed SiC in AGR-1 fuel compacts

As the primary barrier to release of radioactive isotopes emitted from the fuel kernel, retention performance of the SiC layer in tristructural isotropic (TRISO) coated particles is critical to the overall safety of reactors that utilize this fuel design. Most isotopes are well-retained by intact SiC coatings, so pathways through this layer due to cracking, structural defects, or chemical attack can significantly contribute to radioisotope release. In the US TRISO fuel development effort, release of 134Cs and 137Cs are used to detect SiC failure during fuel compact irradiation and safety testing because the amount of cesium released by a compact containing one particle with failed SiC is typically ten or more times higher than that released by compacts without failed SiC. Compacts with particles that released cesium during irradiation testing or post-irradiation safety testing at 1600–1800°C were identified, and individual particles with abnormally low cesium retention were sorted out with the Oak Ridge National Laboratory (ORNL) Irradiated Microsphere Gamma Analyzer (IMGA). X-ray tomography was used for three-dimensional imaging of the internal coating structure to locate low-density pathways through the SiC layer and guide subsequent materialography by optical and scanning electron microscopy. All three cesium-releasing particles recovered from as-irradiated compacts showed a region where the inner pyrocarbon (IPyC) had cracked due to radiation-induced dimensional changes in the shrinking buffer and the exposed SiC had experienced concentrated attack by palladium; SiC failures observed in particles subjected to safety testing were related to either fabrication defects or showed extensive Pd corrosion through the SiC where it had been exposed by similar IPyC cracking.

Assembly Design of Pressurized Water Reactors with Fully Ceramic Microencapsulated Fuel

Fully ceramic microencapsulated (FCM) fuel consists of TRISO (tristructural-isotropic) fuel particles embedded in a ceramic matrix (SiC) to form fuel pellets and rods and offers improved fission product retention and lower operating temperature with expected superior performance in normal and off-normal conditions compared to conventional fuel. When coupled with SiC cladding, FCM fuel eliminates zirconium altogether and is expected to drastically reduce hydrogen generation during a beyond-design-basis accident. In order to be deployed in current or future pressurized water reactors (PWRs), FCM fuel must meet or exceed the neutronic performance of conventional fuel. Limited by low heavy metal loading, an FCM fuel assembly requires increased enrichment and large fuel rods to match the cycle length of a conventional fuel assembly.This study investigated the core design, neutronics, and thermal hydraulics of a PWR loaded with FCM fuel and sought to optimize the assembly design to minimize the enrichment required to reach fuel performance similar to that of conventional fuel. It was found that the implementation of FCM fuel in a 17 × 17 assembly requires close to 20% enrichment and large fuel rods. Such design performs comparably to conventional fuel (4.5% enrichment) in terms of cycle length, reactivity coefficients, intra-assembly power peaking factor, burnable poison penalty, and control rod worth but requires an increase of pumping power. A parametric analysis spanned a large design space varying fuel outer diameter and pitch-to-diameter ratio (P/D) and downselected two alternate assembly designs: 11 × 11 (1.65-cm outer diameter and 1.18 P/D) and 9 × 9 (2.12-cm outer diameter and 1.12 P/D). These designs meet the cycle length requirement with 18.6% and 16.2% enrichments, respectively, but feature a smaller minimum departure from nucleate boiling ratio (MDNBR) compared to a reference assembly. It was estimated that a slight increase in rod outer diameter increases MDNBR to the desired level and implies a pressure drop increase of 10%.

STEM-EDS analysis of fission products in neutron-irradiated TRISO fuel particles from AGR-1 experiment

Historic and recent post-irradiation-examination from the German AVR and Advanced Gas Reactor Fuel Development and Qualification Project have shown that 110 m Ag is released from intact tristructural isotropic (TRISO) fuel. Although TRISO fuel particle research has been performed over the last few decades, little is known about how metallic fission products are transported through the SiC layer, and it was not until March 2013 that Ag was first identified in the SiC layer of a neutron-irradiated TRISO fuel particle. The existence of Pd- and Ag-rich grain boundary precipitates, triple junction precipitates, and Pd nano-sized intragranular precipitates in neutron-irradiated TRISO particle coatings was investigated using Scanning Transmission Electron Microscopy and Energy Dispersive Spectroscopy analysis to obtain more information on the chemical composition of the fission product precipitates. A U-rich fission product honeycomb shape precipitate network was found near a micron-sized precipitate in a SiC grain about ∼5 μm from the SiC-inner pyrolytic carbon interlayer, indicating a possible intragranular transport path for uranium. A single Ag-Pd nano-sized precipitate was found inside a SiC grain, and this is the first research showing such finding in irradiated SiC. This finding may possibly suggest a possible Pd-assisted intragranular transport mechanism for Ag and may be related to void or dislocation networks inside SiC grains. Preliminary semi-quantitative analysis indicated the micron-sized precipitates to be Pd2Si2U with carbon existing inside these precipitates. However, the results of such analysis for nano-sized precipitates may be influenced by the SiC matrix. The results reported in this paper confirm the co-existence of Cd with Ag in triple points reported previously.

Reduction of Environmental Impact of Fixed Bed Nuclear Reactor (FBNR) Waste

The Fix Bed Nuclear Reactor (FBNR) is a pressurized water reactor but its fuel elements are made of Tristructural-Isotropic (TRISO) type particles. Its spent fuel elements may be used as a source of radiation for irradiation purposes in medicine, industry and agriculture. Thereafter, the waste treatment problem is the same as for the fourth generation high-temperature nuclear reactors using TRISO particles. It is found that using the proposed simplified TRISO particles increases the reactivity of the reactor, resulting in higher fuel burnup; while in recycling of its spent fuel, the amount of radioactive carbon is reduced by 57%.

Time-Dependent Neutronic Analysis of a Power-Flattened Gas Cooled Accelerator Driven System Fuelled with Thorium, Uranium, Plutonium, and Curium Dioxides TRISO Particles

This study presents the power flattening and time-dependent neutronic analysis of a conceptual helium gas cooled Accelerator Driven System (ADS) loaded with TRISO (tristructural-isotropic) fuel particles. Target material is lead-bismuth eutectic (LBE). ThO2, UO2, PuO2, and CmO2TRISO particles are used as fuel. PuO2and CmO2fuels are extracted from PWR-MOX spent fuel. Subcritical core is radially divided into 10 equidistant subzones in order to flatten the power produced in the core. Tens of thousands of these TRISO fuel particles are embedded in the carbon matrix fuel pebbles as five different cases. The high-energy Monte Carlo code MCNPX 2.7 with the LA150 library is used for the neutronic calculations. Time-dependent burnup calculations are carried out for thermal fission power (Pth) of 1000 MW using the BURN card. The energy gain of the ADS is in the range of 99.98–148.64 at the beginning of a cycle. Furthermore, the peak-to-average fission power density ratio is obtained between 1.021 and 1.029 at the beginning of the cycle. These ratios show a good quasi-uniform power density for each case. Furthermore, up to 155.1 g233U and 103.6 g239Pu per day can be produced. The considered system has a high neutronic capability in terms of energy multiplication, fissile breeding, and spent fuel transmutation with thorium utilization.

Sustainability of thorium-uranium in pebble-bed fluoride salt-cooled high temperature reactor

Sustainability of thorium fuel in a Pebble-Bed Fluoride salt-cooled High temperature Reactor (PB-FHR) is investigated to find the feasible region of high discharge burnup and negative Flibe (2LiF-BeF2 ) salt Temperature Reactivity Coefficient (TRC). Dispersion fuel or pellet fuel with SiC cladding and SiC matrix is used to replace the tristructural-isotropic (TRISO) coated particle system for increasing fuel loading and decreasing excessive moderation. To analyze the neutronic characteristics, an equilibrium calculation method of thorium fuel self-sustainability is developed. We have compared two refueling schemes (mixing flow pattern and directional flow pattern) and two kinds of reflector materials (SiC and graphite). This method found that the feasible region of breeding and negative Flibe TRC is between 20 vol% and 62 vol% fuel loading in the fuel. A discharge burnup could be achieved up to about 200 MWd/kgHM. The case with directional flow pattern and SiC reflector showed superior burnup characteristics but the worst radial power peak factor, while the case with mixing flow pattern and SiC reflector, which was the best tradeoff between discharge burnup and radial power peak factor, could provide burnup of 140 MWd/kgHM and about 1.4 radial power peak factor with 50 vol% dispersion fuel. In addition, Flibe salt displays good neutron properties as a coolant of quasi-fast reactors due to the strong 9 Be(n,2n) reaction and low neutron absorption of 6 Li (even at 1000 ppm) in fast spectrum. Preliminary thermal hydraulic calculation shows good safety margin. The greatest challenge of this reactor may be the decades irradiation time of the pebble fuel.

Analysis of fission gas release-to-birth ratio data from the AGR irradiations

A series of advanced gas reactor (AGR) irradiation tests is being conducted in the advanced test reactor (ATR) at Idaho National Laboratory (INL) in support of development and qualification of tristructural isotropic (TRISO) fuel used in the High temperature gas-cooled reactor (HTGR). Each AGR test consists of multiple independent capsules containing fuel compacts placed in a graphite cylinder shrouded by a steel shell. These capsules are instrumented with thermocouples (TC) embedded in the graphite enabling temperature control. For AGR-1, the first US irradiation of modern TRISO fuel completed in 2009, there were no particle failures detected. For AGR-2, a few exposed kernels existed in the fuel compacts based upon quality control data. For the AGR-3/4 experiment, particle failures in all capsules were expected because of the use of designed-to-fail (DTF) fuel particles whose kernels are identical to the driver fuel kernels and whose coatings are designed to fail under irradiation. The release-rate-to-birth-rate ratio (R/B) for each of krypton and xenon isotopes is calculated from release rates measured by the germanium detectors used in the AGR fission product monitoring (FPM) system installed downstream from each irradiated capsule. Birth rates are calculated based on the fission power in the experiment and fission product generation models. Thus, this R/B is a measure of the ability of fuel particle coating layers and compact matrix to retain fission gas atoms preventing their release into the sweep gas flow. The major factors that govern gaseous diffusion and release processes are found to be fuel material diffusion coefficient, temperature, and isotopic decay constant. To compare the release behavior among the AGR capsules and historic experiments, the R/B per failed particle is used. HTGR designers use this parameter in their fission product behavior models. For the U.S. TRISO fuel, a regression analysis is performed to establish functional relationships between R/B per failed particle of selected noble gas isotopes and temperature. The effect of isotopic half-life is also obtained from the data. To reduce measurement uncertainty of the release rate, the krypton and xenon isotopes selected for regression analysis have a short enough half-life to achieve equilibrium in the capsule, but are also long enough to provide a measurable signal in the FPM detector. The impact of uncertainty of the estimated number of failed particles in each capsule is also examined. This study found that R/B values for AGR test fuel are comparable to R/B obtained in historic tests. The correlation with isotopic half-life is very stable with time indicating no strong influence of burnup on release. The correlation can be used by reactor designers to estimate fission gas release from postulated failed fuel in HTGR cores.

The measurement of silver diffusivity in zirconium carbide to study the release behavior of 110mAg in the ZrC TRISO-coated nuclear fuel particle

The tri-structural isotropic (TRISO) coated particle fuel has been developed and used for high temperature gas-cooled reactors (HTGRs). It provides a unique robustness of the first barrier for the fission products. The TRISO fuel particle has typically consisted of a UO2 or UCO kernel, surrounded by successive layers of porous carbon, dense inner pyrocarbon, silicon carbide, and dense outer pyrocarbon. During operation, however, the SiC layer has been known to release radioactive silver 110mAg which makes maintenance more difficult and thus costly. Zirconium carbide has been considered as a promising alternative to the SiC fission product barrier. ZrC exhibits high temperature stability and possibly possesses superior Pd resistance, while the retention properties especially for silver have not been adequately studied.To help elucidate the diffusive behavior of silver in the ZrC coating of the TRISO-coated particle, a new diffusion experimental technique, called the encapsulating source method, has been developed by constructing a constant source diffusion couple between ZrC and Ag gas originated from Zr–Ag solid solution. Scanning electron microscopy (SEM), wavelength-dispersive X-ray spectroscopy (WDS), electron backscatter diffraction (EBSD) and optical methods were used to analyze the diffusion couple annealed at 1500 °C. The resultant diffusion coefficient of Ag in single-crystalline ZrC0.84 at 1500 °C was experimentally determined to be about 2.8 (±1.2) × 10−17 m2/s.

An improved design of TRISO particle with porous SiC inner layer by fluidized bed-chemical vapor deposition

Tristructural-isotropic (TRISO) particle has been successful in high temperature gas cooled reactor (HTGR), but an improved design is required for future development. In this paper, the coating layers are reconsidered, and an improved design of TRISO particle with porous SiC inner layer is proposed. Three methods of preparing the porous SiC layer, called high methyltrichlorosilane (MTS) concentration method, high Ar concentration method and hexamethyldisilane (HMDS) method, are experimentally studied. It is indicated that porous SiC layer can be successfully prepared and the density of SiC layer can be adjusted by tuning the preparation parameters. Microstructure and characterization of the improved TRISO coated particle are given based on scanning electron microscope (SEM), X-ray diffraction (XRD), Raman scattering and energy dispersive X-ray (EDX) analysis. It can be found that the improved TRISO coated particle with porous SiC layer can be mass produced successfully. The formation mechanisms of porous SiC layer are also discussed based on the fluidized bed-chemical vapor deposition principle.

Thermal Predictions of the AGR-2 Experiment with Variable Gas Gaps

A new daily as-run thermal analysis was performed at the Idaho National Laboratory for the advanced gas cooled reactor (AGR) test experiment number two (AGR-2) in the Advanced Test Reactor (ATR). This thermal analysis incorporates gas gaps changing with time during the irradiation experiment due to graphite shrinkage resulting from neutron damage. The purpose of this analysis was to calculate the daily average temperatures of each TRISO (tristructural isotropic)–particle fuel compact. A steady-state thermal analysis was performed daily for each capsule with the commercial finite element heat transfer code ABAQUS. These new thermal predictions show the compact fuel temperature dependence on the variable gas gap method. Comparison between measured and calculated temperatures is discussed.

Comparison of diffusion coefficients and activation energies for Ag diffusion in silicon carbide

Abstract The migration of silver (Ag) in silicon carbide (SiC) and 110mAg through SiC of irradiated tri-structural isotropic (TRISO) fuel has been studied for the past three to four decades. However, there is no satisfactory explanation for the transport mechanism of Ag in SiC. In this work, the diffusion coefficients of Ag measured and/or estimated in previous studies were reviewed, and then pre-exponential factors and activation energies from the previous experiments were evaluated using Arrhenius equation. The activation energy is 247.4 kJ·mol−1 from Ag paste experiments between two SiC layers produced using fluidized-bed chemical vapor deposition (FBCVD), 125.3 kJ·mol−1 from integral release experiments (annealing of irradiated TRISO fuel), 121.8 kJ·mol−1 from fractional Ag release during irradiation of TRISO fuel in high flux reactor (HFR), and 274.8 kJ·mol−1 from Ag ion implantation experiments, respectively. The activation energy from ion implantation experiments is greater than that from Ag paste, fractional release and integral release, and the activation energy from Ag paste experiments is approximately two times greater than that from integral release experiments and fractional Ag release during the irradiation of TRISO fuel in HFR. The pre-exponential factors are also very different depending on the experimental methods and estimation. From a comparison of the pre-exponential factors and activation energies, it can be analogized that the diffusion mechanism of Ag using ion implantation experiment is different from other experiments, such as a Ag paste experiment, integral release experiments, and heating experiments after irradiating TRISO fuel in HFR. However, the results of this work do not support the long held assumption that Ag release from FBCVD-SiC, used for the coating layer in TRISO fuel, is dominated by grain boundary diffusion. In order to understand in detail the transport mechanism of Ag through the coating layer, FBCVD-SiC in TRISO fuel, a microstructural change caused by neutron irradiation during operation has to be fully considered.

Ag out-surface diffusion in crystalline SiC with an effective SiO2 diffusion barrier

For applications of tristructural isotropic (TRISO) fuel particles in high temperature reactors, release of radioactive Ag isotope (110mAg) through the SiC coating layer is a safety concern. To understand the diffusion mechanism, Ag ion implantations near the surface and in the bulk were performed by utilizing different ion energies and energy-degrader foils. High temperature annealing was carried out on the as-irradiated samples to study the possible out-surface diffusion. Before and after annealing, Rutherford backscattering spectrometry (RBS) and secondary ion mass spectrometry (SIMS) measurements were employed to obtain the elemental profiles of the implanted samples. The results suggest little migration of buried Ag in the bulk, and an out-diffusion of the implanted Ag in the near-surface region of single crystal SiC. It is also found that a SiO2 layer, which was formed during annealing, may serve as an effective barrier to reduce or prevent Ag out diffusion through the SiC coating layer.

Observations of Ag diffusion in ion implanted SiC

The nature and magnitude of Ag diffusion in SiC has been a topic of interest in connection with the performance of tristructural isotropic (TRISO) coated particle fuel for high temperature gas-cooled nuclear reactors. Ion implantation diffusion couples have been revisited to continue developing a more complete understanding of Ag fission product diffusion in SiC. Ion implantation diffusion couples fabricated from single crystal 4H-SiC and polycrystalline 3C-SiC substrates and exposed to 1500–1625°C, were investigated by transmission electron microscopy and secondary ion mass spectrometry (SIMS). The high dynamic range of SIMS allowed for multiple diffusion régimes to be investigated, including enhanced diffusion by implantation-induced defects and grain boundary (GB) diffusion in undamaged SiC. Estimated diffusion coefficients suggest GB diffusion in bulk SiC does not properly describe the release observed from TRISO fuel.

Deposition of diversely textured buffer pyrolytic carbon layer in TRISO coated particle by controlled manipulation of spouted bed hydrodynamics

Predicting texture and physical properties of pyrolytic carbon (PyC) coatings produced by spouted bed coater/reactor has never been easy for specific yet critical hydrodynamic nature of the reactor. As a part of the TRistructural-ISOtropic (TRISO) coating for nuclear fuel applications, we have deposited PyC on surrogate zirconia particles by cracking of acetylene under variable static bed height, gas velocity and precursor concentration and characterized them thoroughly. From the analysis of samples, a correlation has been drawn between different experimental parameters and physical properties and texture of the said coatings. It has been shown that PyC with larger pores, higher degree of isotropic crystals and larger number of carbon atoms in lamellae are formed at higher spouted bed-height, whereas disordered crystals are produced at higher spouting gas velocities. Likewise, dense and hard coatings with smaller crystallite size are obtained with lower precursor concentration. The results have been adequately explained by invoking the droplet model. Our studies clearly shows that spouted bed hydrodynamics may be used to trap time frame of deposition, control surface temperature of spherical substrates and also the precursor gas concentration and thereby tailor the critical properties of the PyC coatings of TRISO coated particles.

Spent UO2 TRISO coated particles – instant release fraction and microstructure evolution

Abstract The impact of burn-up on the instant release fraction (IRF) from spent fuel was studied using very high burn-up UO2 fuel (∼ 100 GWd/t) from a prototype high temperature reactor (HTR). TRISO (TRi-structural-ISO-tropic) particles from the spherical fuel elements contain UO2 fuel kernels (500 μm diameter) which are coated by three tight layers ensuring the encapsulation of fission products during reactor operation. After cracking of the tight coatings 85Kr and 14C as 14CO2 were detected in the gas fraction. Xe was not detected in the gas fraction, although ESEM (Environmental Scanning Electron Micoscope) investigations revealed an accumulation in the buffer. UO2 fuel kernels were exposed to synthetic groundwater under oxic and anoxic/reducing conditions. U concentration in the leachate was below the detection limit, indicating an extremely low matrix dissolution. Within the leach period of 276 d 90Sr and 134/137Cs fractions located at grain boundaries were released and contribution to IRF up to max. 0.2% respectively 8%. Depending on the environmental conditions, different release functions were observed. Second relevant release steps occurred in air after ∼ 120 d, indicating the formation of new accessible leaching sites. ESEM investigations were performed to study the impact of leaching on the microstructure. In oxic environment, numerous intragranular open pores acting as new accessible leaching sites were formed and white spherical spots containing Mo and Zr were identified. Under anoxic/reducing conditions numerous metallic precipitates (Mo, Tc and Ru) filling the intragranular pores and white spherical spots containing Mo and Zr, were detected. In conclusion, leaching in different geochemical environments influenced the speciation of radionuclides and in consequence the stability of neoformed phases, which has an impact on IRF.

Understanding the 1600°C Fuel Temperature Limit of TRISO Coated Fuel Particles

ABSTRACTThe TRISO (tristructural isotropic) coated fuel particle is made of a uranium oxide kernel coated with three layers of pyrolytic carbon and one of silicon carbide. This fuel, originally used in High Temperature Reactors, has been proposed as accident tolerant fuel for Light Water Reactors after the accident in Fukushima. Although this fuel is capable of retaining fission products within the particle up to 1600°C, little is known on the origin of this temperature limit. Therefore, in order to increase the safety of this type of fuel, it is necessary to understand the origin of the degradation of the materials that compose this fuel. We have studied the effect of temperature on the microstructure and diffusion of silver in pyrolytic carbon coatings produced by fluidized bed chemical vapor deposition. Samples were heat treated at 1000°C, 1400°C and 1700°C for 200 hrs. under inert atmosphere. The effect of temperature on the microstructure and silver diffusion behavior were analyzed by Raman spectroscopy, X-Ray diffraction, optical microscopy, SEM and TEM. We observed that the microstructure of PyC changed drastically above 1400°C, showing the increase in anisotropy and the re-orientation of the graphene planes. The diffusion of silver appears to be also correlated with this change in microstructure.

Computational Fluid Dynamics Analysis for Asymmetric Power Generation in a Prismatic Fuel Block of Fluoride-Salt-Cooled High-Temperature Test Reactor

Abstract The fluoride-salt-cooled high-temperature reactor (FHR) is an advanced reactor concept that uses tristructural isotropic (TRISO) high-temperature fuel and low-pressure liquid salt coolant. A 20-MWth test reactor, as the key step in demonstrating the technical feasibility, is currently under design at Massachusetts Institute of Technology. This study focuses on the coupled conduction and convection heat transfer adopting a three-dimensional unit-cell model with one coolant channel and six one-third fuel compacts. The laminar, transitional, and turbulent flows are investigated with the use of computational fluid dynamic (CFD) software, CD-adapco STARCCM+. The model is validated against theory for developing laminar flow in the benchmark study with excellent agreement. The model is also benchmarked for transitional and turbulent flows by Hausen, Gnielinski, Dittus-Boelter, and Sieder-Tate correlations. Azimuthal distributions of temperature, heat flux, and heat transfer coefficient along the coolant-graphite interface were obtained for the asymmetric heat source, graphite materials, and two different types of salt coolant. The results show that the asymmetric power generation has little impact on peak fuel temperature, interface temperature, and heat transfer coefficient for a unit-cell module in laminar flow regime due to effective thermal conduction of the graphite matrix. In the turbulent flow regime, the effect on the azimuthal heat flux and heat transfer coefficient is more pronounced.

Development of a Thermal-Hydraulic Analysis Code and Transient Analysis for a Fluoride-Salt-Cooled High-Temperature Test Reactor

The fluoride-salt-cooled high-temperature reactor (FHR) is an advanced reactor concept that uses high-temperature tristructural isotropic (TRISO) fuel with a low-pressure liquid salt coolant. Design of the fluoride-salt-cooled high-temperature test reactor (FHTR) is a key step in the development of the FHR technology and is currently in progress both in China and the United States. An FHTR based on pebble-bed core design with a coolant temperature of 600–700°C is being planned for construction by the Chinese Academy of Sciences’ (CAS) Thorium Molten Salt Reactor (TMSR) Research Center, Shanghai Institute of Applied Physics (SINAP). This paper provides preliminary thermal-hydraulic transient analyses of an FHTR using SINAP’s pebble-bed core design as a reference case. A point kinetic model is implemented using computer code by coupling with a simplified porous medium heat transfer model in the core. The founded models and developed code are applied to analyze the safety characteristics of the FHTR by simulating several transient conditions including the unprotected loss of flow, unprotected overcooling, and unprotected transient overpower accidents. The results show that SINAP’s pebble-bed core is a very safe reactor design.

Preparation of the Coated Nuclear Fuel Particle Using the Fluidized Bed-chemical Vapor Deposition (FB-CVD) Method

The tristructural-isotropic (TRISO) type coated fuel particle, which has been commonly used in the current high temperature gas-cooled reactor consists of a microspheric UO2 fuel kernel surrounded by four coated layers: a porous buffer pyrolysis carbon layer (buffer PyC), an inner dense pyrolysis carbon layer (IPyC), a silicon carbide layer (SiC) and an outer dense pyrolysis carbon layer (OPyC). The fluidized bed-chemical vapor deposition (FB-CVD) method was selected for preparing the coated nuclear fuel particle. In this paper, the special design for FB-CVD process is given. The gas distributors with multi-ring inclined holes and the coating furnace with vertical inhomogeneous heating design are developed. Based on experimental study, the TRISO coated nuclear fuel particle has been successfully prepared, and the appropriate parameters in the coating process are obtained.

Experimental measurement and numerical modeling of the effective thermal conductivity of TRISO fuel compacts

Accurate modeling capability of thermal conductivity of tristructural-isotropic (TRISO) fuel compacts is important to fuel performance modeling and safety of Generation IV reactors. To date, the effective thermal conductivity (ETC) of TRISO fuel compacts has not been measured directly. The composite fuel is a complicated structure comprised of layered particles in a graphite matrix. In this work, finite element modeling is used to validate an analytic ETC model for application to the composite fuel material for particle-volume fractions up to 40%. The effect of each individual layer of a TRISO particle is analyzed showing that the overall ETC of the compact is most sensitive to the outer layer constituent. In conjunction with the modeling results, the thermal conductivity of matrix–graphite compacts and the ETC of surrogate TRISO fuel compacts have been successfully measured using a previously developed measurement system. The ETC of the surrogate fuel compacts varies between 50 and 30Wm−1K−1 over a temperature range of 50–600°C. As a result of the numerical modeling and experimental measurements of the fuel compacts, a new model and approach for analyzing the effect of compact constituent materials on ETC is proposed that can estimate the fuel compact ETC with approximately 15–20% more accuracy than the old method. Using the ETC model with measured thermal conductivity of the graphite matrix-only material indicate that, in the composite form, the matrix material has a much greater thermal conductivity, which is attributed to the high anisotropy of graphite thermal conductivity. Therefore, simpler measurements of individual TRISO compact constituents combined with an analytic ETC model, will not provide accurate predictions of overall ETC of the compacts emphasizing the need for measurements of composite, surrogate compacts.