Tristructural Isotropic Research Articles

Internal Radiation Dose Evaluation for an Unruptured Post Release Tristructural Istropic Fuel Particle for Advanced and Micro-reactor Applications.

There are unique benefits from advanced/micro-reactor designs and fuel types that offer safety features in the case of an accident that may reduce environmental consequences compared to conventional reactors and fuels. Tristructural isotropic (TRISO) fuel particles are a robust advanced nuclear fuel type that leads to the unique question of how unruptured, activated TRISO particles will interact with humans. TRISO particles are 900 μm in size, and that particle size restricts internal dose assessment to the ingestion pathway. Activity of the TRISO particle was established by High Temperature Engineering Test Reactor simulations. The TRISO particle encapsulation was assumed to be perfect; exploration of internal dose contribution from radionuclides released from encapsulation was not included. The TRISO particle was assumed to be mixed actively within each alimentary tract compartment such that homogenous distribution could be assumed according to the International Commission on Radiological Protection publication 133. The dose assessment results indicate that the rectosigmoid colon had the highest internal organ dose for both reference male (2.1 Sv) and female (2.3 Sv). The internal dose from ingestion of the scenario-specific TRISO particle was 0.25 Sv for the reference male and 0.29 Sv for the reference female, which exceeds the annual occupational effective dose limit of 0.05 Sv in the Code of Federal Regulations, 10 CFR Part 20 Subpart C. Similarly, the annual occupational limit of 0.5 Sv to any one organ would be exceeded for the left colon, right colon, and rectosigmoid colon for both the reference male and female.

Redistribution of radionuclides in irradiated AGR-1 UCO TRISO fuel after 1800 °C safety testing

Release of radionuclides from intact tristructural-isotropic (TRISO) coated particle fuel at normal and accident conditions is a primary metric of fuel performance. The distribution of fission products and actinides in the TRISO layers of individual particles provides insight on radionuclide transport and release behavior and was determined using scanning electron microscopy analysis. Particles were isolated from an irradiated fuel compact (AGR-1 Compact 4-4-2) and analyzed as-irradiated or after individual particle safety-testing at 1800 °C for 650 h. Particles were selected for comparison based on their remaining 110mAg fission product inventory. These comparisons corroborated the observation that the 110mAg inventory is a marker for relative irradiation temperature based on observed radionuclide distribution in the SiC layer. The comparison also indicated that the in-pile behavior influences the fission product and actinide species interactions with the TRISO layers during high temperature exposure after irradiation. The analysis confirms both palladium and uranium diffusion, as well as other species, are active in the UCO TRISO fuel system at 1800 °C and that palladium transport is active at lower temperatures relative to uranium. While diffusion across the SiC layer was observed, the intact nature of the SiC layer after the 1800 °C, 650-h exposure indicates the SiC layer maintained its functionality as a fission product barrier by mitigating release of radionuclides at beyond accident margin temperatures.

Pressureless sintering of fully ceramic microencapsulated fuels

A new strategy was introduced to achieve high volume fraction of tristructural isotropic (TRISO) particles (> 35 vol%) in fully ceramic microencapsulated (FCM) fuels. The proposed strategy requires (1) applying a controlled coating of a SiC matrix on the TRISO particles, (2) forming the coated TRISO particles using cold isostatic pressing, and (3) sintering the formed sample without applied pressure. The strategy was very effective for preventing both the rupture of TRISO particles and matrix cracking during sintering. The thinner the coating layer, the higher the volume fraction of the TRISO particles obtained in the FCM pellets. However, when the coating thickness was extremely thin (≤ 133 μm), radial cracks were observed near the TRISO particles in the SiC matrix after sintering. The maximum TRISO volume fraction (∼35.3 %) was obtained when the coating thickness was ∼215 μm and the TRISO pellets had no cracks in the SiC matrix.

Method for measurement of TRISO kernel and layer volumes by X-ray computed tomography

Layer dimensions are key parameters for as-fabricated tristructural-isotropic (TRISO) particle fuel as well as for post-irradiation examination of particle performance. Layer thicknesses are typically measured by optical microscopy of the particle cross section near mid-plane, while layer volumes are estimated with serial sectioning and microscopy. This method for measuring layer volumes is limited due to the resolution limit imposed by slice thickness and the effect of the polishing process on delicate irradiated particle microstructures. Image processing software has been developed to segregate the three-dimensional (3D) TRISO particle images provided by x-ray computed tomography (XCT) into high-resolution volumetric data for the kernel, all particle layers, and any internal voids introduced during irradiation. These data can be used to analyze key in-reactor behaviors of TRISO particles such as kernel swelling or buffer shrinkage, which are important inputs for fuel performance modeling.

Analysis of fission product distribution and composition in the TRISO layers of AGR-2 fuel

The Advanced Gas Reactor Fuel Qualification and Development Program's second irradiation experiment focuses on the performance of fuel particles produced in an industrial-scale coating system. Ongoing post-irradiation examination efforts include exploration of tristructural-isotropic (TRISO)-coated particle fuel in the as-irradiated state and subjected to elevated temperatures to simulate accident conditions (safety testing). Scanning electron microscopy (SEM) analysis with energy dispersive x-ray spectroscopy (EDS) was used to show the distribution of fission products within the TRISO layers. The SEM analysis showed a variation of fission product distribution within the TRISO layers based on irradiation temperature, burnup, individual particle fission product inventory retention, safety testing history and kernel composition. A key observation is that particles which likely experienced higher temperature showed a greater diversity and distribution of fission products and actinides in the IPyC and SiC layers, implying that temperature is a dominant variable influencing the transport of radionuclides in TRISO particles.

Role of microstructure on CO corrosion of SiC layer in UO2-TRISO fuel

The Advanced Gas Reactor Fuel Qualification and Development (AGR) program has focused on qualification of UCO kernel tristructural-isotropic (TRISO) particle fuel relative to UO2 kernel TRISO particle fuel. However, a UO2 kernel variant was included in the second AGR irradiation experiment (AGR-2) for comparison and to connect to historic fuel irradiation data. The development of a multiscale, post-irradiation examination (PIE) analysis approach through the AGR Program has allowed for a comprehensive understanding of individual particle failure. This approach has been applied to gain an understanding of SiC layer failure in UO2 kernel TRISO fuel from AGR-2 after safety testing at 1600–1700 °C. Particle failure by intergranular CO corrosion, facilitated by a compromised inner pyrolytic carbon layer, has been confirmed through the combined application of modern x-ray computed tomography and electron microscopy techniques. Additionally, a relationship between grain boundary character and CO corrosion has been identified. This finding provides an opportunity to develop mitigating strategies to improve the resilience of the SiC layer to CO corrosion in UO2 TRISO fuel.

Innovative design and fabrication of generation IV nuclear fuel embedded with carbon nanotube

The generation IV nuclear reactors Tri-structural isotropic (TRISO) fuel suffer failure due to internal pressure generation, corrosion by fission products and kernel migration under temperature gradient. An innovative fuel particle design was conceptualized and realized for the first time in literature by embedding multi-wall carbon nanotube (MWCNT) into porous buffer pyrolytic carbon layer (B.PyC) of TRISO particle to address above issues. MWCNT embedded in B.PyC was obtained by sequential coating using novel chemical vapor deposition reactor. Scanning electron microscopy shows ~15 μm uniformly dispersed MWCNT into B.PyC layer. Transmission electron microscopy and Raman Spectroscopy analysis of the deposited CNT revealed to be multi walled with good crystalline. Results of small angle scattering technique of the composite layer using X-ray and neutrons (SAXS/SANS) reveal better pore structure and could pave new vistas for TRISO fuel development. Intrinsic MWCNT features offer immense potential to mitigate various shortcoming of TRISO particle during actual Generation IV nuclear reactor operation.

Study on the Effects of Enrichment and Fraction of Coated Fuel Particles on Fissile Utilization of 100 MWt Prismatic-type of High Temperature Gas Reactor

High Temperature Gas-cooled Reactor (HTGR) is one of Generation-IV reactor technology that has graphite moderate, Helium gas-cooled, and Coated Fuel Particle (CFP) layered by Tristructural-Isotropic (TRISO). The HTGR has an outlet temperature of around 1000°C that can be utilized for many co-generation processes other than to generate electricity. Due to characteristics of the CFP and TRISO, utilization of fissile material during reactor operation becomes important. This study aims to analyze the effects of enrichment and fraction of CFP on the fissile utilization of 100 MWt HTGR for two different fuels; UO2 and (Th-U)O2. Fissile enrichment is analyzed from 1-20% while the fraction of CFP from 10-60%. Using SRAC2006 code and JENDL-4.0 as nuclear data library, calculation and analysis are performed to find the optimal values of several important neutronic parameters.

Investigation of Zirconium Carbide Ceramic Coating on Surrogate Kernels Deposited by Plasma-Pulsed Laser Deposition: A Preliminary Study

High Temperature Gas-cooled Reactor (HTGR) is one of advanced nuclear reactors with high safety system feature. Indonesia has a plan to build HTGR as an Experimental Power Reactor (in Indonesian: RDE - Reaktor Daya Eksperimental). One of the safety system feature is its fuel system. There are two types of fuel of HTGR i.e. pebble bed type and prismatic type which all of the types are containing TRISO (Tri-Structural Isotropic) which consist of Inner Pyrolitic Carbon, SiC (Silicon Carbide) dan Outer Pyrolitic Carbon. One of the issues of the system is silver (Ag) and palladium (Pd) as the fission product could interact with SiC layer and cause corrosion. One of the candidate to resolve the issue is to replace SiC with ZrC (Zirconium Carbide). In this preliminary study, Zirconium Carbide ceramic has been deposited on a surrogate (uranium-free) fuel microkernels using PLD Plasma-Pulsed Laser Deposition. The parameter during ZrC deposition were temperature of 850°C for the substrate, the number of laser shots of 90,000 and the oxygen background gas with 40 sccm (standard cubic centimeters per minute) with the chamber pressure of 235 mTorr. Thereafter, the samples were analyzed using OM - Microscope Optic, SEM-EDS - Scanning Electron Microscope – Energy Dispersive X-ray Spectroscope, XRD – X-Ray Diffractometer and AFM – Atomic Force Microscope. The results exhibited that the ceramic coating could deposited on the surface of the substrate surface with homogeneously and sticky. The surface was very smooth within the nano-meter range order of ∼80-120 nm of surface roughness. Nevertheless, the results indicated that some of ZrC was react with oxygen at high temperature on the surface of the sample and deposited on the surface of surrogate microkernels.

Quantification of silicon carbide grain structure in TRISO fuel by BSE image analysis

The grain structure of the silicon carbide (SiC) layer in tristructural-isotropic (TRISO) fuels has commonly been qualified in fuel specifications based on a qualitative comparison to visual standards to ensure undesirable microstructure are not produced. This approach is inherently dependent on subjective judgement and provides limited data for analysis or comparison. An alternative method to rapidly quantify (i.e., provide a numeric measure of) the grain structure of the SiC layer in TRISO fuels has been developed. This method relies on the identification of silicon carbide grain boundaries based on brightness variation in backscattered electron (BSE) images through the application of automated image processing algorithms. The speed with which the requisite BSE imaging and analysis may be completed enables the practical use of this method on an industrial scale as a part of qualifying TRISO fuel to meet a given specification on silicon carbide grain structure, which is an important parameter for TRISO fuel performance. The method is benchmarked by comparison to prior Electron-Backscattered Diffraction (EBSD) results for TRISO particles from the Advanced Gas Reactor (AGR) program.

Effect of Structure on the Thermal-Mechanical Performance of Fully Ceramic Microencapsulated Fuel

The effect of non-fuel part size on the thermal-mechanical performance of fully ceramic microencapsulated (FCMTM) Fuel was investigated, and the non-fuel part size was selected according to integrity maintaining of non-fuel part and silicon carbide (SiC) layers. The non-fuel part size can affect the FCMTM temperature and stress distribution greatly by changing the distance between tristructural isotropic (TRISO) particles. The maximum temperature of SiC matrix increased from 1220 K to 1450 K with the non-fuel part size increasing from 100 μm to 500 μm, and the matrix temperature of all the samples was lower than the decomposition point of SiC ceramics. The maximum hoop stress decreased with non-fuel part size, but the inner part exhibiteda crosscurrent trend. The inner part of the SiC matrix lost structure integrity because of the large hoop stress caused by the deformation of TRISO particles, however, the non-fuel parts of FCMTM pellet may maintain their integrity when the non-fuel part size was larger than 300 μm. SiC layers hoop stress increased with non-fuel part size, and the failure probability of SiC layer was lower than 2.2 × 10−4 for the FCMTM pellet with small non-fuel part size. The integrity of non-fuel and SiC layers can be maintained for the FCMTM pellet with the non-fuel part size from 300 μm to 400 μm.

Investigation of the structure and chemical nature of Pd fission product agglomerations in irradiated TRISO particle SiC

Tristructural-isotropic (TRISO)-coated fuel particles are used in high-temperature gas-cooled nuclear reactors. Although the polycrystalline 3C–SiC layer acts as the main barrier to fission product release, post-irradiation examinations have shown that certain metallic fission products are found on the outside of the TRISO coated particle, with no observable evidence of micro-cracks or mechanical failure. In this study, an atomic resolution transmission electron microscopy investigation of a SiC layer from a neutron irradiated (19.38% fissions per initial metal atom average burnup; 1072 °C time-averaged temperature) TRISO-coated particle from the first fuel irradiation experiment as part of the US DOE Advanced Gas Reactor Fuel Qualification and Development (AGR) Program, was conducted. The fission product Pd was found to be present at dislocation cores associated with twins, stacking faults, and their intersections as atomic scale silicide inclusions. EELS elemental quantification of the silicide region showed a reduction of C and increase of Si in the Pd containing inclusion. Pd was also observed as distributed along grain boundaries down to the atomic scale. The local corrosion and dissociation of C from Pd containing grain boundary agglomerations was observed at the atomic level. Elemental quantification of atomically dispersed Pd grain boundary regions using EELS spectrum imaging similarly showed a measurable decrease in C concentration at Pd regions and an increase in Si concentration.

Assembly design of a fluoride salt-cooled high temperature commercial-scale reactor: Neutronics evaluation and parametric analysis

In this study, a novel assembly design is proposed for a fluoride salt-cooled high-temperature commercial-scale (FHCR) reactor. It employs tristructural isotropic (TRISO) fuel particles nestled within removable cylindrical beryllium carbide (Be2C) moderator blocks, which are further contained within prismatic graphite blocks. As the name implies, the FHCR is a preconception of a 3400 MW(t) commercial power reactor that uses FLiBe (LiF-BeF2) as the primary coolant of choice. This paper uses the SERPENT 2 code to conduct a parametric neutronics study on the two-dimensional lattice assembly design of the FHCR. The calculations examine the effects of various fuel enrichment levels, TRISO particle packing fractions, fuel compacts’ pitch sizes, and moderating materials on the cycle length, while also determining the neutron spectrum and various nuclide inventories. Finally, a preliminary core is modeled based on the study conducted on the assembly design. Based on the negative values of the fuel temperature coefficients (FTC), moderator temperature coefficients (MTC), and coolant temperature coefficients (CTC) obtained, the design is determined to be safe. This new assembly design is also able to achieve keff>1 for approximately 3.55 years, translating to a burn-up of 160.6 MWd/KgU.

Modeling reactivity insertion experiments of TRISO particles in NSRR using BISON

This research presents BISON simulations of stress, temperature, and failure probability of tristructural isotropic (TRISO) fuel particles subjected to transient power pulse conditions in the Nuclear Safety Research Reactor (NSRR). By modifying the default elastic properties of the PyC and the UO2 coefficient of thermal expansion correlation, BISON was found to produce suitable agreement with the observed and independently simulated results when appropriate shape and scale Weibull parameters for SiC failure were chosen. The Weibull parameter effects were also explored within the range of SiC failure and were applied using experimental data from hemispherical crush testing.

New method for analysis of X-ray computed tomography scans of TRISO fuel forms

X-ray computed tomography (XCT) is a powerful characterization tool that has been used for examination of both tristructural-isotropic (TRISO) coated particles and fuel forms. Rapid, nondestructive imaging of complete specimens by XCT provides significant characterization advantages in both speed and comprehensiveness over traditional materialographic methods. This paper discusses a method that has been developed to analyze XCT images of pebble fuel forms containing ~18,000 TRISO particles each. The XCT images are three-dimensional with contrast based on relative x-ray transparency. In the case of TRISO fuel forms, XCT image intensity distinguishes between the high-density kernel, the medium-density SiC, and the low-density matrix and carbon layers. Image analysis software has been developed to process XCT data sets to extract spatial distribution information, including particle nearest-neighbor distances, local packing fraction, and fuel-free zone thickness in TRISO fuel forms. Although this type of high-resolution XCT analysis is likely not feasible as a quality control measurement for commercial TRISO fuel, it is highly useful in generating feedback for the development of fuel form fabrication processes.

X-ray tomography study on the crushing strength and irradiation behaviour of dedicated tristructural isotropic nuclear fuel particles at 1000 °C

Two types of dedicated Tristructural isotropic (TRISO) nuclear fuel particles, PyC-1 (Kernel/Buffer/PyC) and PyC-2 (Kernel/Buffer/SiC/PyC) from PYCASSO (Pyrocarbon irradiation for creep and swelling/shrinkage of objects) neutron irradiation experiments, were studied. For unirradiated particles, crushing experiments using a unique hot cell, combined with in situ X-ray computed micro-tomography (XCT) imaging, were conducted at room temperature (RT) and at 1000 °C. Although the SiC layer on the particles is presumed to provide ‘mechanical stability’ to the TRISO particles, results showed a remarkable reduction (~45%) in the crushing strength of the PyC-2 particles at 1000 °C compared to RT. The fracture patterns of the two types of particles, both at the contact zone and on subsequent propagation, differ significantly at RT and 1000 °C. Further, irradiated particles (irradiation temperature: 1000 ± 20 °C; irradiation doses: 1.08–1.23 dpa and 1.49–1.51 dpa) were imaged by XCT; 250 PyC-1 particles and 223 PyC-2 particles were studied in total and the change in radius/layer thickness in each type was examined. It was found that the buffer densification was lower in PyC-1 particles compared to PyC-2 particles, and the PyC layer shrank in the PyC-I particles, whereas it expanded in PyC-2. Results are discussed in terms of how the residual stresses can impact the high-temperature and post-irradiation behavior of these particles.

Microstructure and fission products in the UCO kernel of an AGR-1 TRISO fuel particle after post irradiation safety testing

Tristructural isotropic (TRISO) coated particle fuel, as a key fuel concept for the high-temperature gas-cooled reactors and a candidate accident-tolerant fuel, has been investigated under the US-DOE Advanced Gas Reactor Fuel Development and Qualification Program. Extensive studies have been conducted to evaluate the fission-product release, diffusion of Ag, Pd, and Cd in the SiC layer and the TRISO coating system, and safety-test performance. However, to date there are limited reported results on the fuel kernels’ response to irradiation with or without post irradiation safety testing. To incrementally fill this knowledge gap, extensive studies using transmission electron microscopy (TEM) and atom probe tomography (APT) were conducted on a TRISO fuel-particle kernel with a 19.74% 235U enrichment, irradiated to 18.63% FIMA and subsequently subjected to safety testing at 1600 °C for 300 h. Microstructural characterizations, elemental analysis, and phase identification were conducted using conventional TEM and scanning TEM imaging, energy-dispersive X-ray spectroscopy, selected-area electron diffraction and APT. The following findings were made: (1) significant reconstructions and phase evolutions occurred in the irradiated and post safety-tested fuel kernel, and its microstructure consists of two primary phases, namely a higher-Z (atomic mass) UC phase and a lower-Z UO phase, (2) no fission-gas bubbles are identified within the fuel kernel, that can be attributed to the high-temperature post safety-testing, (3) fission products Zr, Nb, Mo, Ru, Tc and Rh were found to segregate preferentially into UC phase or to form metallic precipitates while the lanthanide fission products tend to stay in the solution of UO phase, and (4) Pd was detected in the rod-shaped precipitates.

Microstructure and thermal physical properties of SiC matrix microencapsulated composites at temperature up to 1900 °C

SiC matrix microencapsulated composites with particle-free buffer (PB) layer were produced by spark plasma sintering for the first time. Microstructure and thermal physical performance of the composites from room temperature to 1900 °C were investigated. Dense composites with rational layout of tristructural isotropic (TRISO) particles and PB layer were obtained. The TRISO particles in the composites kept intact spherical shell structure. Thermal physical properties of the composites approached those of the SiC materials. The composites with high thermal conductivity and low coefficient of thermal expansion were suitable for nuclear applications. The effect of oxide additives on comprehensive performance of the composites was reappraised. Dissolution and recrystallization of SiC in the eutectic liquid phase revealed a new high temperature failure mechanism for the composites.

Preliminary Study of a Prismatic-Core Advanced High-Temperature Reactor Fuel Using Homogenization Double-Heterogeneous Method

The double heterogeneity of the tristructural isotropic (TRISO) fuel in the prismatic-core advanced high-temperature reactor should be accurately and correctly modeled and analyzed, especially for a large-scale loaded with the double-heterogeneity effect. The reactivity-equivalent physical transformation method was developed and employed to enable homogenizing TRISO fuel in a high temperature reactor considering the double heterogeneity and taking into account the large problem involved in performing the whole-core burnup calculation using Monte Carlo transport codes with double-heterogeneity problems. In this work, the heterogeneous effects of a collision of probability calculation method were used to represent the effects of scattering anisotropy on the leakage rates and the isotropic streaming effects due to low optical density in the model. The WIMS and DRAGON codes have been used to perform the calculations of double heterogeneity for the TRISO fuel, fuel compact, and fuel element and the results are compared with the SERPENT Monte Carlo code.

Reactivity analysis of a direct nuclear heated gas turbine with nitrogen coolants

This paper investigates the effects on core reactivity of using a nitrogen coolant within a prismatic High Temperature Gas Reactors (HTGR). The intent is that the nitrogen coolant will be directly compatible with existing technologically mature gas turbine rotating machinery and therefore simplify the design path to a direct nuclear heated closed cycle gas turbine. The 14N isotope of nitrogen makes up the majority (99.6%) of natural nitrogen and has a reasonably significant absorption cross-section for thermal neutrons. An understanding is therefore required of any implications for reactivity insertions during coolant pressure and temperature changes. The analysis was undertaken using the Serpent continuous-energy Monte Carlo particle transport code modelling an infinite lattice core of fuel blocks containing tristructural-isotropic (TRISO) coated fuel particles. The results of the analysis show that for low pressure ratio turbo machinery (10:1) with atmospheric compressor inlet pressure the reactivity swings are small (<60pcm). Equally the effects of temperature transients were also shown to provide only a small positive reactivity coefficient (∼0.09pcm/K) which inversely diminishes with absolute temperature. The conclusion of this paper is that the effect of nitrogen on core reactivity does not prevent its utilisation as a potential reactor coolant for small prismatic HTGRs, potentially enabling direct coupling of existing air derived turbomachinery.