Tristructural Isotropic Particles Research Articles

Evaluation of the Mechanical Properties of TRISO Particles Using Nanoindentation and Ring Compression Testing

Tristructural-isotropic (TRISO) fuel particles are being investigated as a potential fuel for both current and advanced reactors. Due to the size of the particles (<1 mm in diameter with a silicon carbide layer thickness of 30–50 μm), traditional mechanical testing is a challenge, and investigations are limited to micro-scale mechanical testing techniques. In the work presented here, ring crush tests of TRISO particles were performed, and the results are compared with nanoindentation measurements of particles from the same manufactured batch. The elastic modulus results from the nanoindentation and the ring crush tests are found to be in good agreement. In addition, an innovative technique for studying helium implantation effects on the material was investigated utilizing an ORION NanoFab Helium Ion Microscope.

3D Nondestructive Visualization and Evaluation of TRISO Particles Distribution in HTGR Fuel Pebbles Using Cone-Beam Computed Tomography

A nonuniform distribution of tristructural isotropic (TRISO) particles within a high-temperature gas-cooled reactor (HTGR) pebble may lead to excessive thermal gradients and nonuniform thermal expansion during operation. If the particles are closely clustered, local hotspots may form, leading to excessive stresses on particle layers and an increased probability of particle failure. Although X-ray digital radiography (DR) is currently used to evaluate the TRISO distributions in pebbles, X-ray DR projection images are two-dimensional in nature, which would potentially miss some details for 3D evaluation. This paper proposes a method of 3D visualization and evaluation of the TRISO distribution in HTGR pebbles using cone-beam computed tomography (CBCT): first, a pebble is scanned on our high-resolution CBCT, and 2D cross-sectional images are reconstructed; secondly, all cross-sectional images are restructured to form the 3D model of the pebble; then, volume rendering is applied to segment and display the TRISO particles in 3D for visualization and distribution evaluation. For method validation, several pebbles were scanned and the 3D distributions of the TRISO particles within the pebbles were produced. Experiment results show that the proposed method provides more 3D than DR, which will facilitate pebble fabrication research and production quality control.

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.

Comparison study of silicon carbide coatings produced at different deposition conditions with use of high temperature nanoindentation

The elastic modulus and hardness of different silicon carbide (SiC) coatings in tristructural-isotropic (TRISO) fuel particles were measured by in situ high temperature nanoindentation up to 500 °C. Three samples fabricated by different research institutions were compared. Due to varied fabrication parameters the samples exhibited different grain sizes and one contained some visible porosity. However, irrespective of the microstructural features in each case the hardness was found to be very similar in the three coatings around 35 GPa at room temperature. Compared with the significantly coarser grained bulk CVD SiC, the drop in hardness with temperature was less pronounced for TRISO particles, suggesting that the presence of grain boundaries impeded plastic deformation. The elastic modulus differed for the three TRISO coatings with room temperature values ranging from 340 to 400 GPa. With increasing measurement temperature the elastic modulus showed a continuous decrease.

Preparation of Silicon Carbide Coating Layer by Fluidized Bed Chemical Vapor Deposition Using a Halogen-Free Precursor

Tristructural-isotropic (TRISO) particle, with spherical ceramic fuel particle kernels followed by three layers of pyrolytic carbon and one layer of silicon carbide (SiC), has been successful now in high temperature gas cooled reactor (HTGR). The silicon carbide (SiC) layer used in TRISO coated fuel particles is normally produced at high temperatures (~1600°C) via fluidized bed chemical vapor deposition from methyltrichlorosilane (MTS) in a hydrogen environment. The precursor is strong corrosive and the process is not environmentally friendly. In this work, hexamethyldisilane (HMDS) was used instead of MTS and the deposition behavior was investigated via fluidized bed chemical vapor deposition method. Different experimental parameters were tested, such as deposition temperature (800~1450°C) and gas flow ratio of Ar: H2. The deposition rates were obtained and compared. It was found that the optimization parameters of highest deposition rate is 1000°C with the ratio of Ar: H2 of 1:1. The microstructures of the products were further investigated by SEM, XRD and Raman scattering. From the X-ray diffraction pattern it could be inferred that the β-SiC phase was obtained, and free carbon was also found in deposition products. Different types of SiC layer, including dense and porous layer can be prepared. The experimental results validated that HMDS was an alternative precursor for preparing the SiC layer in producing the TRISO particle and other SiC-coated materials in lower temperatures

Neutronic Analysis of LBE-Uranium Spallation Target Accelerator Driven System Loaded with Uranium Dioxide in TRISO Particles

This study presents the neutronic performances of fissile breeding and energy production of a gas cooled accelerator-driven system with LBE-uranium dioxide (UO2) spallation target. The accelerator-driven system is designed and optimized by considering various target materials, in terms of neutronic. Two different materials, LBE + natural UO2 and LBE + 15% enrichment UO2 are selected as target materials. The target zone is divided into two parts, one within the other; the outer part is pure LBE target part, and the inner part is UO2 target part cooled with the helium gas. Tristructural-isotropic (TRISO)-coated fuel particles, containing UO2 fuel, are embedded in a carbon matrix pebble with the packing fraction of a 29%, and the pebbles are placed in the UO2 target part and in the fuel core with the packing fraction of a 60%. The fuel core is cooled with helium that is a high-temperature coolant. The target is bombarded with the continuous beams of a 1 GeV protons to produce high-flux neutrons that enter the fuel core. The fuel core is surrounded with a graphite reflector zone serving as both effective moderation and reflection of these neutrons. Furthermore, the whole system is enclosed by boron carbide, B4C (shielding zone), to prevent the neutrons leakage out of the accelerator-driven system. The high-energy Monte Carlo code MCNPX along with the LA150 library is used for neutronic calculations. The numerical results bring out that the investigated accelerator-driven system has a high neutronic performance, from the energy production and fissile breeding points of view. Namely, it can be obtained over the thermal power of a 350 MW and produced over the fissile breeding of a 300 g/day.

Evaluation of the mechanical performance of silicon carbide in TRISO fuel at high temperatures

The HTR design envisions fuel operating temperatures of up to 1000°C and in case of an accident even 1600°C are conceivable. To ensure safety in all conditions a thorough understanding of the impact of an extreme temperature environment is necessary. This work assesses the high temperature mechanical performance of the silicon carbide (SiC) layer within the tristructural-isotropic (TRISO) fuel particle as it poses the main barrier against fission product release into the primary circuit. Therefore, simulated fuel was fabricated by fluidized bed chemical vapour deposition; varying the deposition conditions resulted in strongly differing SiC microstructures for the various samples. Subsequently the TRISO particles were annealed in inert atmosphere at temperatures ranging from 1600°C up to 2200°C. Scanning electron microscopy and Raman spectroscopy showed that strong disintegration of the SiC layer occurred from 2100°C onwards, but initial signs of porosity formation were visible already at 1800°C. Still, the elastic modulus and hardness as measured by nanoindentation were hardly impaired. After annealing stoichiometric SiC coatings showed a reduction in fracture strength as determined by a modified crush test, however the actual annealing temperature from 1600°C to 2000°C had no measureable effect.Furthermore, a technique was developed to measure the elastic modulus and hardness in situ up to 500°C using a high temperature nanoindentation facility. This approach allows conducting tests while the specimen and indenter tip are heated to a specific measurement temperature, thus obtaining reliable values for the temperature dependent mechanical properties of the material. For the SiC layer in TRISO particles it was found that the elastic modulus decreased slightly from room temperature up to 500°C, whereas the hardness was reduced more severely to approximately half of its ambient temperature value.

Neutronic Design Features of a Transportable Fluoride-Salt-Cooled High-Temperature Reactor

The fluoride-salt-cooled high-temperature reactor (FHR) is a new reactor concept, which combines low-pressure liquid salt coolant and high-temperature tristructural isotropic (TRISO) particle fuel. The refractory TRISO particle coating system and the dispersion in graphite matrix enhance safeguards (nuclear proliferation resistance) and security. Compared to the conventional high-temperature reactor (HTR) cooled by helium gas, the liquid salt system features significantly lower pressure, larger volumetric heat capacity, and higher thermal conductivity. The salt coolant enables coupling to a nuclear air-Brayton combined cycle (NACC) that provides base-load and peak-power capabilities. Added peak power is produced using jet fuel or locally produced hydrogen. The FHR is, therefore, considered as an ideal candidate for the transportable reactor concept to provide power to remote sites. In this context, a 20-MW (thermal power) compact core aiming at an 18-month once-through fuel cycle is currently under design at Massachusetts Institute of Technology (MIT). One of the key challenges of the core design is to minimize the reactivity swing induced by fuel depletion, since excessive reactivity will increase the complexity in control rod design and also result in criticality risk during the transportation process. In this study, burnable poison particles (BPPs) made of B4C with natural boron (i.e., 20% B10 content) are adopted as the key measure for fuel cycle optimization. It was found that the overall inventory and the individual size of BPPs are the two most important parameters that determine the evolution path of the multiplication factor over time. The packing fraction (PF) in the fuel compact and the height of active zone are of secondary importance. The neutronic effect of Li6 depletion was also quantified. The 18-month once-through fuel cycle is optimized, and the depletion reactivity swing is reduced to 1 beta. The reactivity control system, which consists of six control rods and 12 safety rods, has been implemented in the proposed FHR core configuration. It fully satisfies the design goal of limiting the maximum reactivity worth for single control rod ejection within 0.8 beta and ensuring shutdown margin with the most valuable safety rod fully withdrawn. The core power distribution including the control rod’s effect is also demonstrated in this paper.

Precession electron diffraction for SiC grain boundary characterization in unirradiated TRISO fuel

Precession electron diffraction (PED), a transmission electron microscopy-based technique, has been evaluated for the suitability for evaluating grain boundary character in the SiC layer of tristructural isotropic (TRISO) fuel. This work reports the effect of transmission electron microscope (TEM) lamella thickness on the quality of data and establishes a baseline comparison to SiC grain boundary characteristics, in an unirradiated TRISO particle, determined previously using a conventional electron backscatter diffraction (EBSD) scanning electron microscope (SEM)-based technique. In general, it was determined that the lamella thickness produced using the standard focused ion beam (FIB) fabrication process (∼80nm), is sufficient to provide reliable PED measurements, although thicker lamellae (∼120nm) were found to produce higher quality orientation data. Also, analysis of SiC grain boundary character from the TEM-based PED data showed a much lower fraction of low-angle grain boundaries compared to SEM-based EBSD data from the SiC layer of a TRISO-coated particle made using the same fabrication parameters and a SiC layer deposited at a slightly lower temperature from a surrogate TRISO particle. However, the fractions of high-angle and coincident site lattice (CSL)-related grain boundaries determined by PED are similar to those found using SEM-based EBSD. Since the grain size of the SiC layer of TRSIO fuel can be as small as 250nm (Kirchhofer et al., 2013), depending on the fabrication parameters, and since grain boundary fission product precipitates in irradiated TRISO fuel can be nano-sized, the TEM-based PED orientation data collection method is preferred to determine an accurate representation of the relative fractions of low-angle, high-angle, and CSL-related grain boundaries. It was concluded that although the resolution of the PED data is better by more than an order of magnitude, data acquisition times may be significantly longer or the number of areas analyzed needs to be significantly greater than the SEM-based method to obtain a statistically relevant distribution. Also, grain size could be accurately determined but significantly larger analysis areas would be required than those used in this study.

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.

Steady- and Transient-State Analyses of Fully Ceramic Microencapsulated Fuel with Randomly Dispersed Tristructural Isotropic Particles via Two-Temperature Homogenized Model—II: Applications by Coupling with COREDAX

In Part I of this paper, the two-temperature homogenized model for the fully ceramic microencapsulated fuel, in which tristructural isotropic particles are randomly dispersed in a fine lattice stochastic structure, was discussed. In this model, the fuel-kernel and silicon carbide matrix temperatures are distinguished. Moreover, the obtained temperature profiles are more realistic than those obtained using other models. Using the temperature-dependent thermal conductivities of uranium nitride and the silicon carbide matrix, temperature-dependent homogenized parameters were obtained. In Part II of the paper, coupled with the COREDAX code, a reactor core loaded by fully ceramic microencapsulated fuel in which tristructural isotropic particles are randomly dispersed in the fine lattice stochastic structure is analyzed via a two-temperature homogenized model at steady and transient states. The results are compared with those from harmonic- and volumetric-average thermal conductivity models; i.e., we compare keff eigenvalues, power distributions, and temperature profiles in the hottest single channel at a steady state. At transient states, we compare total power, average energy deposition, and maximum temperatures in the hottest single channel obtained by the different thermal analysis models. The different thermal analysis models and the availability of fuel-kernel temperatures in the two-temperature homogenized model for Doppler temperature feedback lead to significant differences.

Steady- and Transient-State Analyses of Fully Ceramic Microencapsulated Fuel with Randomly Dispersed Tristructural Isotropic Particles via Two-Temperature Homogenized Model—I: Theory and Method

As a type of accident-tolerant fuel, fully ceramic microencapsulated (FCM) fuel was proposed after the Fukushima accident in Japan. The FCM fuel consists of tristructural isotropic particles randomly dispersed in a silicon carbide (SiC) matrix. For a fuel element with such high heterogeneity, we have proposed a two-temperature homogenized model using the particle transport Monte Carlo method for the heat conduction problem. This model distinguishes between fuel-kernel and SiC matrix temperatures. Moreover, the obtained temperature profiles are more realistic than those of other models. In Part I of the paper, homogenized parameters for theFCM fuel in which tristructural isotropic particles are randomly dispersed in the fine lattice stochastic structure are obtained by (1) matching steady-state analytic solutions of the model with the results of particle transport Monte Carlo method for heat conduction problems, and (2) preserving total enthalpies in fuel kernels and SiC matrix. The homogenized parameters have two desirable properties: (1) they are insensitive to boundary conditions such as coolant bulk temperatures and thickness of cladding, and (2) they are independent of operating power density. By performing the Monte Carlo calculations with the temperature-dependent thermal properties of the constituent materials of the FCM fuel, temperature-dependent homogenized parameters are obtained.

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%.

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.

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.

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.

Neutronics Studies of Uranium-Bearing Fully Ceramic Microencapsulated Fuel for Pressurized Water Reactors

This study evaluated the neutronics and some of the fuel cycle characteristics of using uranium-based fully ceramic microencapsulated (FCM) fuel in a pressurized water reactor (PWR). Specific PWR lattice designs with FCM fuel have been developed that are expected to achieve higher specific burnup levels in the fuel while also increasing the tolerance to reactor accidents. The SCALE software system was the primary analysis tool used to model the lattice designs. A parametric study was performed by varying tristructural isotropic particle design features (e.g., kernel diameter, coating layer thicknesses, and packing fraction) to understand the impact on reactivity and resulting operating cycle length. To match the lifetime of an 18-month PWR cycle, the FCM particle fuel design required roughly 10% additional fissile material at beginning of life compared with that of a standard uranium dioxide (UO2) rod. Uranium mononitride proved to be a favorable fuel for the fuel kernel due to its higher heavy metal loading density compared with UO2. The FCM fuel designs evaluated maintain acceptable neutronics design features for fuel lifetime, lattice peaking factors, and nonproliferation figure of merit.

TRISO 입자를 포함하는 SiC 복합소결체의 소결 및 특성 평가

Fully ceramic micro encapsulated (FCM) nuclear fuel is a concept recently proposed for enhancing the stability of nuclear fuel. FCM nuclear fuel consists of tristructural-isotropic (TRISO) fuel particles within a SiC matrix. Each TRISO fuel particle is composed of a UO₂ kernel and a PyC/SiC/PyC tri-layer which protects the kernel. The SiC ceramic matrix is created by sintering. In this FCM fuel concept, fission products are protected twice, by the TRISO coating layer and by the SiC ceramic. The SiC ceramic has proven attractive for fuel applications owing to its low neutron-absorption cross-section, excellent irradiation resistivity, and high thermal conductivity. In this study, a SiC-matrix composite containing TRISO particles was sintered by hot pressing with Al₂O₃-Y₂O₂ additive system. Various sintering conditions were investigated to obtain a relative density greater than 95%. The internal distribution of TRISO particles within the SiC-matrix composite was observed using an x-ray radiograph. The fracture of the TRISO particles was investigated by means of analysis of the cross-section of the SiC-matrix composite.

Selection and evaluation of potential burnable absorbers incorporated into modified TRISO particles

The quadruple isotropic (QUADRISO) particle concept modifies tri-structural-isotropic (TRISO) particles to add a layer of burnable absorber next to the fuel kernel. Previous studies have evaluated QUADRISO fuel particle designs for the transmutation of neptunium and plutonium, as a modification to the plutonium–thorium fuel cycle, and for excess reactivity management. This study screens the available nuclear isotope database to identify potential burnable absorbers for incorporation into modified TRISO fuel particles in a representative Very High-Temperature Reactor (VHTR). Based on this screening, eleven potential burnable absorbers are tested as QUADRISO burnable absorber layers, in either oxide or carbide forms. The addition of a burnable absorber layer to the TRISO fuel resulted in an increase in reactor operational lifetime in all cases considered in the present work when compared to a reactor with the same excess reactivity without burnable absorbers. The B4C, CdO and Gd2O3 cases resulted in the greatest operational lifetime, for the same initial excess reactivity, and the addition of these burnable absorbers to TRISO fuel particles decreased the magnitude of the excess reactivity change over the reactor lifetime when compared to the unmodified TRISO case. Thermal cross-section effects in QUADRISO particles generally result in excess reactivity decreasing with increasing temperature.

A Neutronic Investigation of the Use of Fully Ceramic Microencapsulated Fuel for Pu/Np Burning in PWRs

An investigation of the utilization of TRistructural-ISOtropic (TRISO)-coated fuel particles for the burning of plutonium/neptunium (Pu/Np) isotopes in typical Westinghouse four-loop pressurized water reactors is presented. Though numerous studies have evaluated the burning of transuranic isotopes in light water reactors (LWRs), this work differentiates itself by employing Pu/Np-loaded TRISO particles embedded within a silicon carbide (SiC) matrix and formed into pellets, constituting the fully ceramic microencapsulated (FCM) fuel concept that can be loaded into standard LWR fuel element cladding. This approach provides the capability of Pu/Np burning and, by virtue of the multibarrier TRISO particle design and SiC matrix properties, will allow for greater burnup of Pu/Np material, plus improved fuel reliability and thermal performance. In this study, a variety of heterogeneous assembly layouts, which utilize a mix of FCM rods and typical UO2 rods, and core loading patterns were analyzed to demonstrate the neutronic feasibility of Pu/Np-loaded TRISO fuel. The assembly and core designs herein reported are not fully optimized and require fine-tuning to flatten power peaks; however, the progress achieved thus far strongly supports the conclusion that with further rod/assembly/core loading and placement optimization, Pu/Np-loaded TRISO fuel and core designs that are capable of balancing Pu/Np production and destruction can be designed within the standard constraints for thermal and reactivity performance in pressurized water reactors.