Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “TRISO”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 271 records · Page 15

Fuel Design and Fabrication: TRISO Particle Fuel

Tristructural-isotropic (TRISO) particle fuel is an advanced, mature technology for nuclear energy generation. Inherent containment of radionuclides within the fuel obviates the need for expensive reactor containment buildings; however, this benefit comes at the cost of increased complexity in fabrication relative to other nuclear fuel forms. In addition, rigorous fuel characterization and quality control efforts are required to ensure that the credited radionuclide containment within the fuel is realized.

Helmreich, Grant↗

Effect of PyC Irradiation-Induced Creep on the Stress of TRISO-Coated Fuel Particles - 18474

Irradiation-induced Pyrocarbon (PyC) creep affects the stresses of a TRISO-coated particle. An analytical solution with a 4-parameter linear viscoelastic model was used to investigate the effects of PyC creep on the particle stresses. PyC transient and steady-state creep effects on particle stresses were analyzed. PyC creep parameter effects on the particle stresses were studied, including PyC creep coefficient, creep Poisson s ratio in the steady-state component, and Young s modulus in the transient component. The results showed that PyC irradiation-induced creep decreased the particle stresses. Transient creep affected particle stresses only at the irradiation early stage. Lower Young s modulus and higher creep coefficient in the transient component resulted in a larger stresses decrease, and large Poisson s ratio and creep coefficient in the steady-state creep decreased particle stresses. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Micro/nano-structural examination and fission product identification in neutron irradiated AGR-1 Triso fuel - 18529

Advanced electron microscopic and micro-analysis techniques were developed and applied to study irradiation effects and fission-product behavior in selected low-enriched uranium-oxide/uranium-carbide tri-structural-isotropic (TRISO)-coated particles from fuel compacts in four capsules irradiated to burnups of 11.2 to 19.6% fissions per initial metal atom (FIMA) consisting of Baseline, Variant 1, and Variant 3 fuel types. Trend analysis shows precipitates were mostly random in their distribution along the perimeter of the inner pyrolytic carbon-silicon carbide (IPyC-SiC) interlayer with only weak association with kernel protrusion and buffer fractures. Pd is dominantly found in most precipitates in both intra and intergranular locations. Nano-sized Ag is predominantly found in grain boundaries and triple points with only two findings of Ag inside a SiC grain in two different compacts (Baseline and Variant 3 fueled compacts). Generally, more element combinations exist for precipitates from particles with relatively low Ag retention compared to particles with relatively high Ag-retention irrespective of fuel type. This study shows the presence of Cs in particles from all compacts evaluated. From this work, no single fission product mechanism hypothesis can be reported. The complexity of mechanisms is further highlighted by the multiple variations of elemental combinations found in the more than 700 fission product precipitates examined. It seems that movement of Ag is not assisted by a specific element in all cases. Therefore, it is not necessarily true that a chemical-assisted transport mechanism is dominant. The presence of Ag predominantly on grain boundaries suggests that a grain boundary transport mechanism may be prominent. Studies to determine the effect of neutron damage are recommended for future work. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

An Approach to Improve the Retention of the Fission Product Ag in TRISO Fuel Particles - 18596

The silicon carbide (SiC) coating included in TRISO particles acts as the main barrier against the radioactive fission products release into the coolant circuit. However, radioactive silver (Ag{sup 110m}) was identified outside the intact particles with a very high release fraction, which could be problematic in the licensing process for modern HTR fuel. Our previous work proposed that dissolution/recrystallization of SiC by Ag could help Ag migrate at 1800 deg. C. The aim of this work is to study whether dissolution of SiC by Ag could take place and contribute to the Ag migration at operation temperature (from 1000 to 1300 deg. C). Both, experiment and thermodynamic analysis, showed that SiC dissolution by Ag was possible above 890 deg. C. But the dissolution from 1000 to 1300 deg. C is far less effective in moving Ag due to the insufficient driving force for activating self-penetration into the grain boundaries, compared with the previous work conducted at 1800 deg. C. Compared with silver alone, the silver-palladium mixture was more permeable in SiC and showed a diffusive behavior. The diffusion front was Ag rich while most palladium remained in reaction zone rather than in the diffusion front. Furthermore, thermodynamic analysis suggested that non-stoichiometric SiC with additional free silicon could impede such a reaction. Hence it is conceivable that a sacrificial layer of SiC+Si on the inner side of the main SiC coating could improve fission product retention overall without a significant increase in cost. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Electron microscopy study of Pd, Ag, and Cs in carbon areas in the locally corroded SiC layer in a neutron-irradiated TRISO fuel particle

Here, a detailed electron microscopy study was performed on focused ion beam-prepared lamellae from different locations relative to a crack across the inner pyrolytic carbon layer of a neutron-irradiated tristructural isotropic-coated particle. The distribution and composition of fission products across the inner pyrolytic carbon and silicon carbide (SiC) layers were studied. Previously, this crack was identified in the particle that released significant inventory fractions of cesium and silver during irradiation and displayed localized palladium pileup with SiC degradation. In this study, carbon areas were found in the SiC layer close to the crack tip and they had precipitates that consisted mostly of palladium silicides or palladium, with silver and/or cadmium frequently identified. Results confirmed that areas in the SiC layer close to the crack tip with localized accumulation of palladium were corroded by palladium, forming pure carbon areas and palladium silicide that provided pathways for silver, cadmium and cesium migration.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Radial Deconsolidation and Leach-Burn-Leach of Eight As-Irradiated AGR-3/4 TRISO Fuel Compacts

Eight as-irradiated AGR-3/4 fuel compacts were subjected to destructive post-irradiation examination via radial-deconsolidation-leach-burn-leach (RDLBL) at INL. The RDLBL process deconsolidated the compacts in multiple, radial steps, followed by a final, single-step axial deconsolidation. The samples generated at each step were analyzed for isotopes of key fission products and actinides. After each deconsolidation step, the compact volume was assessed, and this was used to normalize the measured quantity of nuclides of interest to give a volumetric concentration as a function of radial position within the compact. The total inventories of measured fission products and actinides and the radial concentration profiles were compared among the eight compacts deconsolidated at INL and four other as-irradiated compacts examined at ORNL. The results were analyzed for the effects of irradiation temperature. These results will be used for comparisons with fission product transport models and as input from which fission product diffusivities can be calculated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

AGR TRISO Fuel Performance Modeling in FY-25

This report summarizes the activities performed in FY-25 to support fuel performance modeling for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. This includes implementation of a new uranium oxycarbide (UCO) kernel swelling rate model, inner pyrolytic carbon (IPyC) cracking behavior and failure predictions, development of new UCO and silicon carbide (SiC) cesium diffusion parameters, and the thermomechanical behavior of particle layers using experimental data from micro-tensile strength testing.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

An advanced evaluation of spouted beds scale-up for coating TRISO nuclear fuel particles using Radioactive Particle Tracking (RPT)

A set of dimensionless groups has been proposed in the literature by He et al. (1997) to scale-up gas-solid spouted beds while maintaining their hydrodynamics similarity. The literature reported studies do not provide conclusive assessments about this methodology. Therefore, in this work, we have applied an advanced non-invasive Radioactive Particle Tracking (RPT) technique for the first time to evaluate such scale-up methodology by measuring the local solids velocity, normal and shear stresses and the turbulent kinetic energy. The axial and azimuthal averaged radial profiles of solids velocity, normal stresses, shear stresses, and turbulent kinetic energy illustrate that the similarity of the hydrodynamics has not been attained when the proposed set of dimensionless groups has been matched using two sizes of spouted beds of 0.076 m and 0.152 m and sets of operating conditions. The conclusion is consistent with the recent reported findings by measuring cross sectional distribution and radial profiles of solids and gas holdups along the bed height using gamma-ray computed tomography and by the limited point measurements of solids velocity and holdup using optical fiber probe. It is clear that local measurements of hydrodynamic parameters are essential for detailed assessment of scale-up methodologies. Here, the presented results of our work in terms of the components profiles of the particles radial velocities and turbulent parameters are also valuable for benchmarking computational fluid dynamics codes and models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Distribution of fission products palladium, silver, cerium and cesium in the un-corroded areas of the locally corroded SiC layer of a neutron irradiated TRISO fuel particle

Here, detailed electron microscopy studies were performed to investigate the distribution and composition of fission products in the SiC layer of a tristructural isotropic coated particle exhibiting localized corrosion. Previous studies on this particle indicated that pure carbon areas in the SiC layer, resulting from localized corrosion of SiC by Pd, provide pathways for Ag, Cd and Cs migration. This study reveals the presence of Ag- and/or Cd-containing precipitates in un-corroded SiC areas. Ag/Cd may exist by themselves or coexist with Pd. Ag/Cd mainly transport along SiC grain boundaries. An Ag-Pd-Cd precipitate was identified at a stacking fault inside a SiC grain, suggesting that intragranular transport of Ag/Cd is possible. Ce is present with Pd or Pd-U in some precipitates >50 nm. U and Ce frequently coexist with each other, whereas Ag/Cd usually does not coexist with U or Ce. No Cs was detected in any precipitates in the areas examined.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Physicochemical evolution of uranium nitride kernel microstructure with varying carbon distribution for advanced TRISO fuel forms

Uranium nitride (UN) has emerged as a fuel candidate for advanced nuclear reactor concepts due to its superior uranium density, thermal conductivity, and high melting temperature. However, the fabrication route for converting UO 2 to UN is complex and difficult to standardize. Although the chemistry of this conversion process is well-studied, more insight into the physicochemical dynamics of this conversion using advanced characterization techniques can help further our understanding of this material system. This work leveraged thermogravimetric analysis (TGA), X-ray diffraction (XRD), and nondestructive 3D X-ray computed tomography (XCT) to characterize dynamic microstructural changes in the UO 2 → UCO → UN fabrication pathway for two kernels with a varying carbon distribution in the starting composition. TGA and XRD were used to quantify changes in the mass, density, and chemical composition of the two kernels, while three-dimensional image processing and segmentation of XCT data were used to quantify the volume, surface area, and spatial distribution of features within each kernel for multiple steps along the fabrication pathway. The analysis indicates distinct differences between the two kernels that are correlated to downstream conversion efficiency. In conclusion, this work is among the first to perform 3D quantification of physicochemical evolution during UN conversion, providing quantitative correlation between processing, properties, and expected fuel performance.

Nuclear fuel↗

Generation IV benchmarking of TRISO fuel performance models under accident conditions: Modeling input data

This document presents the benchmark plan for the calculation of particle fuel performance on safety testing experiments that are representative of operational accidental transients. The benchmark is dedicated to the modeling of fission product release under accident conditions by fuel performance codes from around the world, and the subsequent comparison to post-irradiation experiment (PIE) data from the modeled heating tests. The accident condition benchmark is divided into three parts: the modeling of a simplified benchmark problem to assess potential numerical calculation issues at low fission product release; the modeling of the AGR-1 and HFR-EU1bis safety testing experiments; and, the comparison of the AGR-1 and HFR-EU1bis modeling results with PIE data. The simplified benchmark case, thereafter named NCC (Numerical Calculation Case), is derived from ''Case 5'' of the International Atomic Energy Agency (IAEA) Coordinated Research Program (CRP) on coated particle fuel technology [IAEA 2012]. It is included so participants can evaluate their codes at low fission product release. ''Case 5'' of the IAEA CRP-6 showed large code-to-code discrepancies in the release of fission products, which were attributed to ''effects of the numerical calculation method rather than the physical model''[IAEA 2012]. The NCC is therefore intended to check if these numerical effects subsist. The first two steps imply the involvement of the benchmark participants with a modeling effort following the guidelines and recommendations provided by this document. The third step involves the collection of the modeling results by Idaho National Laboratory (INL) and the comparison of these results with the available PIE data. The objective of this document is to provide all necessary input data to model the benchmark cases, and to give some methodology guidelines and recommendations in order to make all results suitable for comparison with each other. The participants should read this document thoroughly to make sure all the data needed for their calculations is provided in the document. Missing data will be added to a revision of the document if necessary.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Advanced Electron Microscopy and Micro analytical technique development and application for Irradiated TRISO Coated Particles from the AGR-1 Experiment

A series of up to seven irradiation experiments are planned for the Advanced Gas Reactor (AGR) Fuel Development and Quantification Program, with irradiation completed at the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) for the first experiment (i.e., AGR-1) in November 2009 for an effective 620 full power days. The objective of the AGR-1 experiment was primarily to provide lessons learned on the multi-capsule test train design and to provide early data on fuel performance for use in fuel fabrication process development and post-irradiation safety testing data at high temperatures. This report describes the advanced microscopy and micro-analysis results on selected AGR-1 coated particles.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

GEN-IV Benchmarking of Triso Fuel Performance Models under accident conditions modeling input data

This document presents the benchmark plan for the calculation of particle fuel performance on safety testing experiments that are representative of operational accidental transients. The benchmark is dedicated to the modeling of fission product release under accident conditions by fuel performance codes from around the world, and the subsequent comparison to post-irradiation experiment (PIE) data from the modeled heating tests. The accident condition benchmark is divided into three parts: • The modeling of a simplified benchmark problem to assess potential numerical calculation issues at low fission product release. • The modeling of the AGR-1 and HFR-EU1bis safety testing experiments. • The comparison of the AGR-1 and HFR-EU1bis modeling results with PIE data. The simplified benchmark case, thereafter named NCC (Numerical Calculation Case), is derived from “Case 5” of the International Atomic Energy Agency (IAEA) Coordinated Research Program (CRP) on coated particle fuel technology [IAEA 2012]. It is included so participants can evaluate their codes at low fission product release. “Case 5” of the IAEA CRP-6 showed large code-to-code discrepancies in the release of fission products, which were attributed to “effects of the numerical calculation method rather than the physical model” [IAEA 2012]. The NCC is therefore intended to check if these numerical effects subsist. The first two steps imply the involvement of the benchmark participants with a modeling effort following the guidelines and recommendations provided by this document. The third step involves the collection of the modeling results by Idaho National Laboratory (INL) and the comparison of these results with the available PIE data. The objective of this document is to provide all necessary input data to model the benchmark cases, and to give some methodology guidelines and recommendations in order to make all results suitable for comparison with each other. The participants should read this document thoroughly to make sure all the data needed for their calculations is provided in the document. Missing data will be added to a revision of the document if necessary. 09/2016: Tables 6 and 8 updated. AGR-2 input data added

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Verification of Triso Fuel Burnup Using Machine Learning Algorithms

Pebble Bed Reactors are fueled with fuel pebbles that are circulated multiple times through the reactor vessel before discharge. During the normal operation of a PBR, ejected pebbles are returned to the reactor or discharged depending on the fuel burnup and physical condition of the pebbles. The burnup measurement is usually based on detected radiation signatures of fission products accumulated in the pebble fuel over burnup. Previous research has shown that height of photopeaks of fission products, such as 134 Cs, 137 Cs, 154 Eu, etc., can be used independently or in combination to infer or predict the level of burnup in the fuel. However, it remains challenging to measure such complex sources due to self-shielding effects, strong radiation background and intervening materials. Another operational challenge is the required high throughput of burnup measurement, which necessitates limited measurement time and thus impacts quality of measured gamma-ray spectra. Hence, advanced spectral analysis methods are needed to analyze the noisy gamma spectra and predict the burnup values. We propose to use machine learning (ML) method to interpret gamma-ray spectra and predict the burnup values of the pebbles. ML has achieved widespread success and adoption across a few domains that require pattern recognition and analysis in varied data types. In this work, we apply three proven ML approaches - multilayer perceptrons, convolutional neural networks, and transformers - to the task of predicting fuel burnup from measured gamma spectra, and compile a dataset of simulated spectra for training and validation of the ML models. In this paper, we will discuss the network architecture of these three ML approaches and compare the performance of the simplest of these (MLP) to a standard linear regression.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Reconstruction of Fission Product Distribution from Tomographic Scans in TRISO Fuel Graphitic Matrix and Nuclear Grade Graphites

An image reconstruction method was developed to rectify shortcomings of earlier methods that became apparent as destructively sampled data became available. This reconstruction method was applied to the tomographic gamma scans of nuclear graphite and graphitic matrix samples from AGR-3/4. There is generally agreement between profiles measured via destructive sampling of these rings and profiles from tomographic reconstruction.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗