TRISO coated fuel particles with enhanced SiC properties

The silicon carbide (SiC) layer used for the formation of TRISO coated fuel particles is normally produced at 1500-1650 °C via fluidized bed chemical vapor deposition from methyltrichlorosilane in a hydrogen environment. In this work, we show the deposition of SiC coatings with uniform grain size throughout the coating thickness, as opposed to standard coatings which have larger grain sizes in the outer sections of the coating. Furthermore, the use of argon as the fluidizing gas and propylene as a carbon precursor, in addition to hydrogen and methyltrichlorosilane, allowed the deposition of stoichiometric SiC coatings with refined microstructure at 1400 and 1300 °C. The deposition of SiC at lower deposition temperatures was also advantageous since the reduced heat treatment was not detrimental to the properties of the inner pyrolytic carbon which generally occurs when SiC is deposited at 1500 °C. The use of a chemical vapor deposition coater with four spouts allowed the deposition of uniform and spherical coatings.     

THETRIS: A micro-scale temperature and gas release model for TRISO fuel

The dominating mechanism in the passive safety of gas-cooled, graphite-moderated, high-temperature reactors (HTRs) is the Doppler feedback effect. These reactor designs are fueled with submillimeter-sized kernels formed into tristructural-isotropic (TRISO) particles that are imbedded in a graphite matrix. The best spatial and temporal representation of the feedback effect is obtained from an accurate approximation of the fuel temperature. Micro-scale models of TRISO particles are necessary in order to obtain accurate predictions during fast transients or when parameters internal to the TRISO are needed. Most accident scenarios in HTRs are characterized by large time constants and slow changes in the fuel and moderator temperature fields. In these situations, a meso-scale, or pebble- and compact-scale, solution provides a good approximation of the fuel temperature as the fission thermal energy transports out of the kernel and into the surrounding matrix with a much shorter time constant. Therefore, in most cases, the matrix can be assumed to be in quasi-static equilibrium with the kernels. These models, however, fail to provide accurate information on the state of the various components of the TRISO during the early stages of transients. Since the coated particles constitute one of the fundamental design barriers for the release of fission products, it becomes important to understand the transient behavior inside this containment system. An explicit TRISO fuel temperature model named THETRIS has been developed and incorporated into the CYNOD–THERMIX-KONVEK suite of coupled codes. The code includes gas-release models that provide a simple predictive capability of the internal pressure during transients. The new model yields similar results to those obtained with other micro-scale fuel models of TRISO particles, but with the added capability to analyze gas release, internal pressure buildup, and effects of a gap in the TRISO. Analysis of bounding benchmark transients yield good agreement with other codes in which the TRISO particles are modeled explicitly. In addition, a sensitivity study of the potential effects on the transient behavior of high-temperature reactors due to the presence of an inter-layer gap is included. Although the formation of a gap occurs under special conditions, its consequences on the dynamic behavior of the reactor can yield responses during fast transients that depart significantly from those in which no gap is present in the model. The new model was applied to an extreme (beyond design basis) scenario in order to observe the behavior of the fuel during a large prompt critical reactivity insertion. Although a large amount of fission energy was deposited rapidly into the fuel, the kernel temperature is shown to stay well below the melting point and the silicon carbide layer remained well below the temperature above which failure is expected to occur. The explicit treatment of the TRISO particle geometry leads to much lower estimations of power peaking during the transient and a greater degree of negative Doppler feedback.     

TRISO fuel volume fraction and homogeneity:a nondestructive characterization

A new nondestructive method to estimate the volume fraction and homogeneity of tristructural isotropic(TRISO)-coated fuel particles in fuel compacts designed for high-temperature reactors has been developed using image analysis of conventional X-radiographs. The method is demonstrated on surrogate fuel compacts containing TRISO-coated particles with kernels made of zirconium dioxide. The methodology incorporates a correction for superimposed images of TRISO particles such that a single X-ray image obtained in any one random orientation is sufficient to characterize the fuel compact in terms of volume fraction and homogeneity. The method is based on the virtual segregation of images of each particle inside the compact with the aid of a calibration standard.     

Utilization of TRISO fuel with reactor grade plutonium in CANDU reactors

Large quantities of plutonium have been accumulated in the nuclear waste of civilian LWRs and CANDU reactors. Reactor grade plutonium and heavy water moderator can give a good combination with respect to neutron economy. On the other hand, TRISO type fuel can withstand very high fuel burn-up levels. The paper investigates the prospects of utilization of TRISO fuel made of reactor grade plutonium in CANDU reactors. TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 68%. The fuel compacts conform to the dimensions of CANDU fuel compacts are inserted in rods with zircolay cladding.In the first phase of investigations, five new mixed fuel have been selected for CANDU reactors composed of 4% RG-PuO₂ + 96% ThO₂; 6% RG-PuO₂ + 94% ThO₂; 10% RG-PuO₂ + 90% ThO₂; 20% RG-PuO₂ + 80% ThO₂; 30% RG-PuO₂ + 70% ThO₂. Initial reactor criticality (k_∞,0 values) for the modes , , , and are calculated as 1.4294, 1.5035, 1.5678, 1.6249, and 1.6535, respectively. Corresponding operation lifetimes are ∼0.65, 1.1, 1.9, 3.5, and 4.8 years and with burn ups of ∼30 000, 60 000, 100 000, 200 000 and 290 000 MW d/tonne, respectively. The higher initial plutonium charge is the higher burn ups can be achieved.In the second phase, a graphical-numerical power flattening procedure has been applied with radially variable mixed fuel composition in the fuel bundle. Mixed fuel fractions leading to quasi-constant power production are found in the 1st, 2nd, 3rd and 4th row to be as 100% PuO₂, 80/20% PuO₂/ThO₂, 60/40% PuO₂/ThO₂, and 40/60% PuO₂/ThO₂, respectively. Higher plutonium amount in the flattened case increases reactor operation lifetime to >8 years and the burn up to 580 000 MW d/tonne.Power flattening in the bundle leads to higher power plant factor and quasi-uniform fuel utilization, reduces thermal and material stresses, and avoids local thermal peaks. Extended burn-up grade implies drastic reduction of the nuclear waste material per unit energy output for final waste disposal.     

Selection and properties of alternative forming fluids for TRISO fuel kernel production

Current Very High Temperature Reactor (VHTR) designs incorporate TRi-structural ISOtropic (TRISO) fuel, which consists of a spherical fissile fuel kernel surrounded by layers of pyrolytic carbon and silicon carbide. An internal sol–gel process forms the fuel kernel using wet chemistry to produce uranium oxyhydroxide gel spheres by dropping a cold precursor solution into a hot column of trichloroethylene (TCE). Over time, gelation byproducts inhibit complete gelation, and the TCE must be purified or discarded. The resulting TCE waste stream contains both radioactive and hazardous materials and is thus considered a mixed hazardous waste. Changing the forming fluid to a non-hazardous alternative could greatly improve the economics of TRISO fuel kernel production. Selection criteria for a replacement forming fluid narrowed a list of ～10,800 chemicals to yield ten potential replacement forming fluids: 1-bromododecane, 1-bromotetradecane, 1-bromoundecane, 1-chlorooctadecane, 1-chlorotetradecane, 1-iododecane, 1-iodododecane, 1-iodohexadecane, 1-iodooctadecane, and squalane. The density, viscosity, and surface tension for each potential replacement forming fluid were measured as a function of temperature between 25 °C and 80 °C. Calculated settling velocities and heat transfer rates give an overall column height approximation. 1-bromotetradecane, 1-chlorooctadecane, and 1-iodododecane show the greatest promise as replacements, and future tests will verify their ability to form satisfactory fuel kernels.     

Optical anisotropy measurements of TRISO nuclear fuel particle cross-sections: The method

The analysis of two-modulator generalized ellipsometry microscope (2-MGEM) data to extract information on the optical anisotropy of coated particle fuel layers is discussed. Using a high resolution modification to the 2-MGEM, it is possible to obtain generalized ellipsometry images of coating layer cross-sections with a pixel size of 2.5 μm and an optical resolution of ∼4 μm. The most important parameter that can be extracted from these ellipsometry images is the diattenuation, which can be directly related to the optical anisotropy factor (OAF or OPTAF) used in previous characterization studies of tristructural isotropic (TRISO) coated particles. Because high resolution images can be obtained, the data for each coating layer contains >6000 points, allowing considerable statistical analysis. This analysis has revealed that the diattenuation of the inner pyrocarbon (IPyC) and outer pyrocarbon (OPyC) coatings varies significantly throughout the layer. The 2-MGEM data can also be used to determine the principal axis angle of the pyrocarbon layers, which is nearly perpendicular to the TRISO radius (i.e., growth direction) and corresponds to the average orientation of the graphene planes.     

Irradiation performance of experimental fuel particles coated with silicon-alloyed pyrocarbon: A review

A review has been conducted on the use of silicon-alloyed pyrocarbon (Si-PyC) as an improved coating material for the two types of fuel particles used in the cores of high-temperature gas-cooled reactors. Based on recent data from extensive irradiation testing and postirradiation annealing of such experimental fuel particles, it is concluded that Si-PyC coatings offer considerable promise as replacements for the standard pure pyrocarbon (PyC) coatings used on thorium-based fertile fuels that have BISO coating designs. The primary advantage here is improved retention of fission products from bred U-233, with diffusion coefficients being as much as 100 times smaller for Si-PyC than for PyC. However, there is no significant improvement in mechanical performance of Si-PyC coatings over standard PyC coatings under irradiation. As a result, there is no incentive for using these coatings on TRISO particle designs of the type used on uranium-based fissile fuels, because here a silicon carbide barrier layer provides superior fission-product retention.     

Fission product monitoring of TRISO coated fuel for the advanced gas reactor-1 experiment

The US Department of Energy has embarked on a series of tests of TRISO coated particle reactor fuel intended for use in the Very High Temperature Reactor (VHTR) as part of the Advanced Gas Reactor (AGR) program. The AGR-1 TRISO fuel experiment, currently underway, is the first in a series of eight fuel tests planned for irradiation in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). The AGR-1 experiment reached a peak compact averaged burnup of 9% FIMA with no known TRISO fuel particle failures in March 2008. The burnup goal for the majority of the fuel compacts is to have a compact averaged burnup greater than 18% FIMA and a minimum compact averaged burnup of 14% FIMA. At the INL the TRISO fuel in the AGR-1 experiment is closely monitored while it is being irradiated in the ATR. The effluent monitoring system used for the AGR-1 fuel is the Fission Product Monitoring System (FPMS). The FPMS is a valuable tool that provides near real-time data indicative of the AGR-1 test fuel performance and incorporates both high-purity germanium (HPGe) gamma-ray spectrometers and sodium iodide [NaI(Tl)] scintillation detector-based gross radiation monitors. To quantify the fuel performance, release-to-birth ratios (R/B's) of radioactive fission gases are computed. The gamma-ray spectra acquired by the AGR-1 FPMS are analyzed and used to determine the released activities of specific fission gases, while a dedicated detector provides near-real time count rate information. Isotopic build up and depletion calculations provide the associated isotopic birth rates. This paper highlights the features of the FPMS, encompassing the equipment, methods and measures that enable the calculation of the release-to-birth ratios. Some preliminary results from the AGR-1 experiment are also presented.     

Determining the shear properties of the PyC/SiC interface for a model TRISO fuel

The fracture behavior of TRISO-coated fuel particles is dependent on the shear strength of the interface between the inner pyrolytic carbon (PyC) and silicon carbide coatings. This study evaluates the interfacial shear properties and the crack extension mechanism for TRISO-coated model tubes using a push-out technique. The interfacial debond shear strength was found to increase with increasing sample thickness and finally approached a constant value. The intrinsic interfacial debond shear strength of ∼280 MPa was estimated. After the layer is debonded, the applied load is primarily transferred by interfacial friction. A non-linear shear-lag model predicts that the residual clamping stress at the interface is ∼350 MPa, and the coefficient of friction is ∼0.23, yielding a frictional stress of ∼80 MPa. These relatively high values are attributed to the interfacial roughness. Of importance in these findings is that this unusually high interfacial strength could allow significant loads to be transferred between the inner PyC and SiC in application, potentially leading to failure of the SiC layer.     

Modification in the stress calculation of PyC material properties in TRISO fuel particles under irradiation

Tristructural-isotropic (TRISO) coated fuel particles are a key component in the high-temperature gas-cooled reactor fuel elements. The stresses in a TRISO-coated particle depend upon the coating layer properties which are affected by neutron irradiation. The modifications of pyrolytic carbon (PyC) material properties under irradiation require the use of time-dependent material parameters in the stress calculations, which was performed with an analytic solution in this study. The experimental results indicated that PyC density increased at the early stages of irradiation while PyC anisotropy increased obviously at the later stages of irradiation, both of which caused particle stresses to increase. The PyC creep coefficient was modeled as increasing with neutron fluence. This significantly decreased the silicon carbide and inner pyrolytic carbon tangential and interface radial stresses. Increasing the Young's modulus of the PyC had little effect on the particle stresses. As a result, compared with constant material properties, the dynamic variation of PyC material parameters under irradiation further improved particle stress calculations and made the stress model closer to actual reactor conditions.     

Reactivity control technique for a pressurized water reactor with an inventive TRISO fuel particle composition

In this research paper a reactivity control technique has been suggested for the conceptual design of a compact sized pressurized water reactor (PWR) core with an inventive tristructural-isotropic (TRISO) fuel particle composition. This conceptual design is a light water cooled and moderated reactor which utilizes TRISO fuel particles in PWR technology. The use of TRISO fuel in PWR technology improves integrity of the design due to its fission fragments retention ability. The fuel provides first retention barrier within fuel itself against the release of fission fragments that makes this design concept safer and environment friendly. The suggested TRISO fuel particle composition has a small amount of Pu-240 with 2.0 w/o in the place of U-238 which acts as reactivity suppressor. Reactor codes WIMS-D/4 and CITATION have been used for simulation and core design modeling. Results reveals that the amount of excess reactivity can be reduced significantly by using a small amount of Pu-240 in TRISO fuel which in turns reduces the number of gadolinia rods in the core required for excess reactivity control and completely eliminates the requirement of soluble boron system. Therefore the effective and optimal use of reactivity suppressor and burnable poison suppresses and flattens the core excess reactivity throughout the core life and hence number of control rods can be reduced without compromising on the shutdown margin.     

Kinetic Monte Carlo (KMC) simulation of fission product silver transport through TRISO fuel particle

A mesoscale kinetic Monte Carlo (KMC) model developed to investigate the diffusion of silver through the pyrolytic carbon and silicon carbide containment layers of a TRISO fuel particle is described. The release of radioactive silver from TRISO particles has been studied for nearly three decades, yet the mechanisms governing silver transport are not fully understood. This model atomically resolves Ag, but provides a mesoscale medium of carbon and silicon carbide, which can include a variety of defects including grain boundaries, reflective interfaces, cracks, and radiation-induced cavities that can either accelerate silver diffusion or slow diffusion by acting as traps for silver. The key input parameters to the model (diffusion coefficients, trap binding energies, interface characteristics) are determined from available experimental data, or parametrically varied, until more precise values become available from lower length scale modeling or experiment. The predicted results, in terms of the time/temperature dependence of silver release during post-irradiation annealing and the variability of silver release from particle to particle have been compared to available experimental data from the German HTR Fuel Program (Gontard and Nabielek [1]) and Minato and co-workers (Minato et al. [2]).     

Performance modeling of Deep Burn TRISO fuel using ZrC as a load-bearing layer and an oxygen getter

The effects of design choices for the TRISO particle fuel were explored in order to determine their contribution to attaining high-burnup in Deep Burn modular helium reactor fuels containing transuranics from light water reactor spent fuel. The new design features were: (1) ZrC coating substituted for the SiC, allowing the fuel to survive higher accident temperatures; (2) pyrocarbon/SiC “alloy” substituted for the inner pyrocarbon coating to reduce layer failure and (3) pyrocarbon seal coat and thin ZrC oxygen getter coating on the kernel to eliminate CO. Fuel performance was evaluated using General Atomics Company’s PISA code. The only acceptable design has a 200-μm kernel diameter coupled with at least 150-μm thick, 50% porosity buffer, a 15-μm ZrC getter over a 10-μm pyrocarbon seal coat on the kernel, an alloy inner pyrocarbon, and ZrC substituted for SiC. The code predicted that during a 1600 °C postulated accident at 70% FIMA, the ZrC failure probability is <10^?4.     

Modeling spent TRISO fuel for geological disposal: corrosion and failure under oxidizing conditions in the presence of water

High temperature gas reactors (HTGRs) are being considered for near term deployment in the United States under the GNEP program and farther term deployment under the Gen IV reactor design (U.S. DOE Nuclear Energy Research Advisory Committee, 2002). A common factor among current HTGR (prismatic or pebble) designs is the use of TRISO coated particle fuel. TRISO refers to the three types of coating layers (pyrolytic carbon, porous carbon, and silicon carbide) around the fuel kernel, which is both protected and contained by the layers. While there have been a number of reactors operated with coated particle fuel, and extensive amount of research has gone into designing new HTGRs, little work has been done on modeling and analysing the degradation rates of spent TRISO fuel for permanent geological disposal. An integral part of developing a spent fuel degradation modeling was to analyze the waste form without taking any consideration for engineering barriers. A basic model was developed to simulate the time to failure of spent TRISO fuel in a repository environment. Preliminary verification of the model was performed with comparison to output from a proprietary model called GARGOYLE that was also used to model degradation rates of TRISO fuel. A sensitivity study was performed to determine which fuel and repository parameters had the most significant effect on the predicted time to fuel particle failure. Results of the analysis indicate corrosion rates and thicknesses of the outer pyrolytic carbon and silicon carbide layers, along with the time dependent temperature of the spent fuel in the repository environment, have a significant effect on the time to particle failure. The thicknesses of the kernel, buffer, and IPyC layers along with the strength of the SiC layer and the pressure in the TRISO particle did not significantly alter the results from the model. It can be concluded that a better understanding of the corrosion rates of the OPyC and SiC layers, along with increasing the quality control of the OPyC and SiC layer thicknesses, can significantly reduce uncertainty in estimates of the time to failure of spent TRISO fuel in a repository environment.     

A coupling methodology for mesoscale-informed nuclear fuel performance codes

This study proposes an approach for capturing the effect of microstructural evolution on reactor fuel performance by coupling a mesoscale irradiated microstructure model with a finite element fuel performance code. To achieve this, the macroscale system is solved in a parallel, fully coupled, fully-implicit manner using the preconditioned Jacobian-free Newton Krylov (JFNK) method. Within the JFNK solution algorithm, microstructure-influenced material parameters are calculated by the mesoscale model and passed back to the macroscale calculation. Due to the stochastic nature of the mesoscale model, a dynamic fitting technique is implemented to smooth roughness in the calculated material parameters. The proposed methodology is demonstrated on a simple model of a reactor fuel pellet. In the model, INL’s BISON fuel performance code calculates the steady-state temperature profile in a fuel pellet and the microstructure-influenced thermal conductivity is determined with a phase field model of irradiated microstructures. This simple multiscale model demonstrates good nonlinear convergence and near ideal parallel scalability. By capturing the formation of large mesoscale voids in the pellet interior, the multiscale model predicted the irradiation-induced reduction in the thermal conductivity commonly observed in reactors.     

TRISO燃料颗粒等效导热系数理论模型研究

三层各向同性碳包覆(TRISO)燃料颗粒由核芯和4层包覆层组成,具有良好的裂变产物包容能力,其等效导热系数是计算弥散微封装燃料等效导热系数的重要基础。本文首先从球坐标下基本导热方程出发,基于多相固体宏观等效导热理论,建立了TRISO燃料颗粒等效导热系数理论计算模型;然后,结合固-固二元复合材料等效导热系数Chiew-Glandt模型分析了锆基微封装燃料(M3)芯体等效导热系数。结果表明,本文开发的模型可有效模拟TRISO燃料等效导热系数。基于开发的TRISO等效导热系数模型计算获得了全陶瓷微封装燃料(FCM)的等效导热系数。     

UN核芯TRISO燃料颗粒破损概率模型研究

TRISO燃料颗粒由核芯和4层包覆层组成,具有良好的裂变产物包容能力。TRISO燃料颗粒破损概率是表征TRISO燃料事故安全特性的关键参数。本文基于修正的PANAMA破损概率计算方法,在考虑UN核芯裂变气体释放导致的气体内压以及内外致密热解炭层辐照蠕变和收缩作用的基础上,开发了UN核芯TRISO燃料颗粒压力壳式破损概率计算方法,并采用IAEA基准题6和基准题9对模型进行了验证;基于开发的UN核芯TRISO颗粒破损概率计算方法,采用随机抽样统计方法分析了事故工况下UN核芯和包覆层设计参数（包括包覆层尺寸及密度）对UN核芯TRISO燃料颗粒破损概率的影响。研究结果显示,疏松热解炭（Buffer）层设计参数是影响TRISO颗粒破损概率的关键因素,可通过降低Buffer层尺寸及密度分布设计标准偏差的方法降低UN核芯TRISO燃料颗粒的破损概率。