Tristructural Isotropic Research Articles

Verification and Validation of Depletion Calculations for TRISO Fuel Against Data from the Second Advanced Gas Reactor Irradiation Campaign

Verification and validation of the Monte Carlo particle transport code Serpent 2 depletion calculation capabilities in the context of tristructural isotropic (TRISO) particles was performed against available data from the second Advanced Gas Reactor irradiation campaign (AGR-2), which was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory. The AGR-2 test train contained both uranium oxycarbide (UCO) and uranium dioxide (UO2) TRISO particles, which were irradiated for 12 ATR power cycles over a period of 3.5 years. For this study, the AGR-2 test train was reconstructed in full detail using Serpent 2, and depletion calculations replicating the experimental irradiation conditions were performed. The fast neutron fluence and AGR-2 fuel burnup resulting from the depletion calculations for the 12 irradiation cycles are compared with the corresponding reported MCNP-computed data. The total AGR-2 fuel burnup accumulated by the end of the AGR-2 test train irradiation is also compared with available experimental data. The fission product inventory at the end of the AGR-2 irradiation obtained from the Serpent 2 depletion calculations is compared with the reported computational data. The Serpent 2 results for the fast neutron fluence (En > 0.18 MeV) differ on average by 5.10% from the MCNP-computed results. Similarly, the Serpent 2 model underpredicts the AGR-2 fuel burnup at the end of the irradiation by 5.13%, on average for all capsules, when compared with the MCNP-computed results and by 2.01% when compared with the experimental results. Last, the fission product inventory calculated by Serpent 2 is underpredicted by 5.27%, on average for all capsules, in comparison with the reference MCNP data. Serpent 2 is generally in satisfactory agreement with the AGR-2 reported values. These results provide confidence in the performance of Serpent 2 depletion calculations for TRISO fuel particles.

Whole-Core 3D Multiphysics Transient Modeling of a Prismatic Micro High-Temperature Gas-Cooled Reactor

A three-dimensional whole-core transient coupled thermal-hydraulic and neutronics code system for modeling prismatic high-temperature gas-cooled reactors (HTGRs) is presented. The discrete ordinates method code PHANTOM-SN was used to solve the multigroup neutron transport problem with cross sections generated with Serpent. The new finite element code OPERA was developed to solve the heat equation in the core and includes simplified subcodes for the coolant, reactor pressure vessel, and concrete containment building, as well as the power conversion cycle. Core graphite thermal conductivity degradation is included as a function of temperature and irradiation temperature. A 20-MW(thermal) HTGR design was modeled using the coupled multiphysics code to prove inherent safety. We simulated steady state, a depressurized loss of forced cooling (DLOFC), a partial blockage, and a reactivity insertion incident. We show that the DLOFC is not the most severe scenario for the fuel temperature in this prismatic micro HTGR. Upon a DLOFC, the peak fuel temperature remains well below the tri-structural isotropic (TRISO) fuel limits, even when the power is increased to 40 MW(thermal). However, during a partial blockage incident of one fuel assembly stack, the maximum fuel temperature reaches 2300°C, severely exceeding the limits. We furthermore contend that the graphite thermal conductivity values used in modeling should always be made explicit and that the temperature of irradiation should be included as a parameter since it can cause a sharp decrease (up to 97%) in the conductivity. We show that using unirradiated graphite parameters leads to an underestimation in peak temperature of 165°C while using a relatively low power density compared to other HTGRs. Finally, we argue that for prismatic HTGRs with a central reflector, bypass flow may lower the maximum fuel temperature.

A MOOSE-Based Model for Fission Product Transport and Source Term Estimation for High-Temperature Gas-Cooled Reactors

Thanks to fuel elements containing tristructural isotropic (TRISO) particles combined with a low core power density and passive feedback mechanisms leading to modest temperature rises in the event of accidental events, high-temperature gas-cooled reactors (HTGRs) offer a high degree of reliability in terms of fission product retention. While the anticipated source term for HTGRs is expected to be very low, it is important to provide a quantitative estimate of radiological releases during nominal and accidental conditions. We propose a computationally efficient mechanistic source term methodology relying on the Multiphysics Object Oriented Simulation Environment (MOOSE) for tracking fission product transport from TRISO particles up to the coolant pressure boundary, as well as modeling the transport and potential deposition of these nuclides inside the reactor coolant loop. The proposed computational scheme is applied to estimate source term inventories for a representative 10-MW(thermal) prismatic high-temperature microreactor and is qualitatively compared against known release fractions. In addition to providing an alternate analysis tool, this MOOSE model can help reactor designers quantify the influence of key design parameters relevant for studies of radiological dose consequences.

Availability of Critical Benchmark Experiments for the Pebble Tanker Transportation Model for Nuclear Criticality Safety Validation of TRISO Pebbles

This study addresses the need for comprehensive investigations into TRi-structural ISOtropic (TRISO) fuel pebble transportation validation. In this work, an exploratory model, the pebble tanker(PT), was developed with the aim of facilitating the validation of nuclear criticality safety calculations in the context of industrial-scale transportation of TRISO fuel. The PT model was designed to investigate the availability and applicability of critical benchmark experiments crucial for assessing the transportation of these pebbles. This work incorporated sensitivity/uncertainty (S/U) similarity studies to quantify the applicability of critical benchmark experiments and to address nuclear data uncertainties in the context of TRISO transportation. Two container models were investigated: one for the Hermes-type pebble and one for the Pebble Bed Modular Reactor (PBMR)–type pebble. The models were simplified, considering fuel, containment, and either water or air, to enable a focus on the underlying physics of applications involving TRISO fuel pebbles using the PT model. A crucial aspect under consideration was the capacity of the transport package to hold pebbles while ensuring subcriticality in the flooded state. An approach in the criticality validation process involves assessing the similarity between systems through an integral index parameter evaluation. This involves calculating a correlation coefficient (referred to as ck) based on shared nuclear data–induced uncertainty between a benchmark experiment and the application of the PT model. To facilitate this analysis, the SCALE tools, particularly the CSAS6-Shift, TSUNAMI-3D-Shift, and TSUNAMI-IP sequences, were employed for comprehensive studies in neutronics and S/U analysis. Our findings showed that there are sufficient critical experimental benchmarks to perform this validation of the PT model in the most reactive state, i.e. when the tanker is flooded. This paper provides valuable insights into validating a transport package for Generation IV TRISO fuel pebbles.

TRISO Fuel’s Safety Functions, Contributions to Reactor Safety, and Necessary Safety Limits

Safety functions are the actions, passive or active, that structures, systems, and components of a nuclear facility contribute to the safety of the workers, the public, or the environment. Well-defined safety functions are the foundation of a solid safety case for a reactor. For reactors that use tri-structural isotropic (TRISO)-coated particles, the TRISO fuel plays an important part in the safety case because of its ability to contain radionuclides in the fuel itself. This ability enables the use of a functional containment strategy for the reactor where radionuclide retention is the primary safety function supported by the safety functions of controlling reactivity control and controlling heat rejection. This paper establishes at a deeper level the role that TRISO fuel plays in each of these safety functions and the associated quality assurance and testing requirements for TRISO particle manufacturing to ensure these safety functions. Safety limits necessary to protect these safety functions include manufacturing specifications, operational limits, time-at-temperature limits, and fission gas release activity limits. This approach demonstrates the role that specific aspects of TRISO fuel play in protecting the safety of workers, the public, and the environment.

Development and demonstration of a BISON–Griffin modeling framework for the design of targeted TRISO transient experiments in the Transient Reactor Test Facility

Development and demonstration of a BISON–Griffin modeling framework for the design of targeted TRISO transient experiments in the Transient Reactor Test Facility

Improving the Resolution and Computational Efficiency of Depletion Simulations on a TRISO-Fueled Microreactor

Modeling individual TRi-structural ISOtropic (TRISO) particles is possible using constructive solid geometry in Monte Carlo neutron transport codes (such as MCNP); however, the billions of surfaces required incurs considerable computational expense. Depletion simulations, in conjunction with Monte Carlo transport, are also inherently slow, requiring many individual criticality calculations to approximate time-dependent behavior. Additionally, choosing acceptable time step sizes, spatial zoning, and other depletion parameters involves prior knowledge and/or iteration by the user. The slow and complex nature of depletion calculations greatly increases the cost and duration of the design iteration process. The importance of tracked isotopes, spatial zones, and time stepping during depletion simulations is evaluated globally using neutron multiplication factor, total isotope inventory, and burnup. Different depletion schemes are compared using the TRISO-fueled Snowflake microreactor. Woodcock Delta Tracking and a Reactivity-equivalent Physical Transformation (RPT) are investigated as potential approaches to increase the speed of Monte Carlo neutron transport calculations involving TRISO fuel. The Monte Carlo N-Particle® Code (MCNP®) Version 6.3 is used to obtain the transport solution, and depletion calculations are performed using its internally coupled version of CINDER90. A prototype MCNP delta tracking module under development at Los Alamos National Laboratory is also used. Although a single spatial depletion region is sometimes acceptable, large errors in core lifetime and total 238Pu, 241Pu, and 242Pu mass are possible. The prototype delta tracking module significantly speeds up criticality calculations; however, depletion calculations are not always faster than those performed with surface tracking. The RPT method is effective at reducing CPU times in both criticality and depletion calculations, without significantly impacting neutron multiplication factor and total isotope inventory. The necessary number of time steps to converge total isotope inventory was also explored; the required number and spacing varied greatly depending on the objective of the depletion simulation.

Unravelling the impact of palladium on ruthenium-induced corrosion of SiC coatings in TRISO particles

Unravelling the impact of palladium on ruthenium-induced corrosion of SiC coatings in TRISO particles

Multiscale, mechanistic modeling of irradiation-enhanced silver diffusion in TRISO particles

Multiscale, mechanistic modeling of irradiation-enhanced silver diffusion in TRISO particles

Influence of temperature, oxygen partial pressure, and microstructure on the high-temperature oxidation behavior of the SiC Layer of TRISO particles

Influence of temperature, oxygen partial pressure, and microstructure on the high-temperature oxidation behavior of the SiC Layer of TRISO particles

Understanding Ag liquid migration in SiC through ex-situ and in-situ Ag-Pd/SiC interaction studies

Understanding Ag liquid migration in SiC through ex-situ and in-situ Ag-Pd/SiC interaction studies

Analysis of micro-defect-induced cracking in the IPyC layer of TRISO particle with the XFEM

PurposeWe use the extended finite element method (XFEM) to model the whole process of initiation and propagation of cracks in the inner dense pyrolytic carbon (IPyC) layer of tri-structural isotropic (TRISO) particle induced by the microdefect in an irradiation-induced thermomechanical coupling environment and study the effect of microdefect sizes on the propagation path.Design/methodology/approachThe irradiation-induced thermal–mechanical coupling analysis is first conducted for the representative volume element (RVE) of the TRISO particle by using the conventional finite element method (CFEM) so that the stress distribution is obtained. The stress results are then restored for the enriched elements, and the simulation of crack initiation and propagation is eventually carried out by using the XFEM.Findings1. As a crack initiates in the IPyC layer, it will terminate at the free edge of the RVE TRISO particle in the end. 2. The size of the microdefect has a significant impact on the propagation path.Originality/valueThe ceramic dispersion microencapsulated (CDM) fuel is a good accident-resistant fuel whose safe operation is crucial to the safety and reliability of the whole nuclear reactor. It is of great scientific significance and practical value to study the irradiation-induced thermomechanical coupling stress distribution and cracking behavior in the IPyC layer of TRISO particles for the CDM fuel. Crack initiation and propagation analysis is challengeable for this complex multi-layer structure. This can help understand the failure mechanism of TRISO particles and evaluate the operation safety of the reactor.

Sensitivity analysis and fuel performance of a control rod withdrawal in the generic fluoride-salt-cooled high temperature reactor

Sensitivity analysis and fuel performance of a control rod withdrawal in the generic fluoride-salt-cooled high temperature reactor

Modular gas-cooled reactor core optimal design and additive manufacturing performance based on SiC matrix dispersed with TRi-structural ISOtropic fuel

Modular gas-cooled reactor core optimal design and additive manufacturing performance based on SiC matrix dispersed with TRi-structural ISOtropic fuel

X-ray Computed Tomography of Tristructural Isotropic (TRISO) Fuel from the AGR-5/6/7 Irradiation Tests

X-ray Computed Tomography of Tristructural Isotropic (TRISO) Fuel from the AGR-5/6/7 Irradiation Tests

Thermal Model of the AGR-5/6/7 Experiment with Offset Gas Gaps

Because of their high impact on calculated temperatures, the offset gas gaps between the graphite holder and capsule wall are evaluated for the thermal model of the Advanced Gas Reactor 5/6/7 (AGR-5/6/7) irradiation experiment. Fuel compact temperatures are a crucial factor in assessing the irradiation performance of tristructural isotropic (TRISO)-coated-particle fuel for use in high-temperature gas-cooled reactors. TRISO fuel irradiation performance is a core tenet of the U.S. Department of Energy’s Advanced Gas Reactor Fuel Development and Qualification Program, conducted in the Advanced Test Reactor at Idaho National Laboratory. Specifically, the irradiation experiments were intended to provide irradiation performance data to support fuel process development, qualify fuel for normal operating conditions, support the development of fuel performance models and codes, and provide irradiated fuel and materials for postirradiation examination and safety testing. The thermal model used to predict the fuel compact temperatures was a three-dimensional finite element model that was subject to uncertainty. The most dominant factor in the uncertainty pertaining to calculated fuel temperatures is the gas gap uncertainty, which is caused by the graphite holder nub-to-capsule wall clearance. In AGR-5/6/7 capsules, the graphite holders can be shifted away from being perfectly centered inside the stainless steel capsule walls. This is due to irradiation-induced graphite shrinkage or to insufficient nubs on the graphite holder leading to azimuthally varying gas gaps during irradiation. A method was developed to find the best offset magnitude and direction at the top and bottom of the graphite holders in Capsules 1 and 2, which experienced higher temperatures than originally predicted among the five AGR-5/6/7 axially dispersed capsules irradiated between 2018 and 2020. The differences between the calculated and the measured temperatures of thermocouples (TCs) placed in the graphite holder near the fuel compacts were used as the basis for the holder offset determination. The best-fit offset assumes that the root-mean-square error between the measured and calculated TC temperatures is a minimum value. Typical offsets in the range of 0.002 to 0.004 in. increase the fuel compact temperatures in the range of 100°C to 200°C when compared to the zero offset for Capsules 1 and 2. Additionally, because of irradiation-induced changes in the capsule geometries, the holder offsets change throughout the irradiation period.

Identification of TRISO pebbles at arbitrary orientation using pairs of X-ray radiographs

Identification of TRISO pebbles at arbitrary orientation using pairs of X-ray radiographs

Leveraging Optimal Sparse Sensor Placement to Aggregate a Network of Digital Twins for Nuclear Subsystems

Nuclear power plants (NPPs) require continuous monitoring of various systems, structures, and components to ensure safe and efficient operations. The critical safety testing of new fuel compositions and the analysis of the effects of power transients on core temperatures can be achieved through modeling and simulations. They capture the dynamics of the physical phenomenon associated with failure modes and facilitate the creation of digital twins (DTs). Accurate reconstruction of fields of interest (e.g., temperature, pressure, velocity) from sensor measurements is crucial to establish a two-way communication between physical experiments and models. Sensor placement is highly constrained in most nuclear subsystems due to challenging operating conditions and inherent spatial limitations. This study develops optimized data-driven sensor placements for full-field reconstruction within reactor and steam generator subsystems of NPPs. Optimized constrained sensors reconstruct field of interest within a tri-structural isotropic (TRISO) fuel irradiation experiment, a lumped parameter model of a nuclear fuel test rod and a steam generator. The optimization procedure leverages reduced-order models of flow physics to provide a highly accurate full-field reconstruction of responses of interest, noise-induced uncertainty quantification and physically feasible sensor locations. Accurate sensor-based reconstructions establish a foundation for the digital twinning of subsystems, culminating in a comprehensive DT aggregate of an NPP.

Shutdown dose rate calculations in high-temperature gas-cooled reactors using the MCNP-ORIGEN activation automation tool

Background High-temperature gas-cooled reactors (HTGRs) have many distinct features from the current light water reactor (LWR) fleet, and potential decommissioning strategies should take them into consideration. The proper characterization of the shutdown dose rates can establish a suitable strategy. Methods This article introduces a new shutdown dose rate calculation capability that relies on the MCNP-ORIGEN activation automation tool and the MCNP repeated structures to explicitly model the TRistructural ISOtropic (TRISO) particles as decay radiation sources. Results Three exercises are conducted with this capability, and their results discussed. The first two exercises verify and demonstrate the workflow. The third exercise proposes a decommissioning strategy for a high-temperature gas-cooled microreactor and studies its feasibility from the shutdown dose rate standpoint. The calculations yield a dose rate map above at the reactor citadel after a three month cool-down, showing values above 40 mSv/h. Conclusions The findings imply that the decommissioning strategy based on placing the reactor in a safe shutdown state and allowing it to cool down for three months before removing various components and extracting the fuel assemblies, may not be sufficient to minimize unnecessary exposure of personnel.

The Use of High-Density UN Fuel in Heat-Pipe Microreactors

Heat-pipe microreactors (HPMRs) are very small-scale nuclear reactors that employ heat pipes (HPs) for heat removal. HPMRs can be easily integrated with other forms of renewable energies, can be used for emergency responses to disaster relief zones, can be deployed in remote locations not connected to the grid, and can be removed from sites and replaced by new ones. HPMRs can also be used for space missions as HPs do not rely on gravity for heat transfer. Conventional fuel materials, such as uranium oxide (UO2) and uranium oxycarbide (UCO), are currently considered in most existing HPMR designs, but ceramic uranium nitride (UN) fuel that has high uranium density, high thermal conductivity, and high melting point may become a better fuel candidate. Through neutronics calculations, this paper assesses the impact of using UN fuel in HPMRs with two different neutron spectra (fast and thermal) and two different fuel forms [traditional solid fuel pellets and TRi-structural-ISOtropic (TRISO) fuel compacts]. It was concluded that retrofitting HPMRs with UN fuel has the potential to reduce the initial 235U enrichment requirement by ~3 wt% (to keep the same cycle length) or increase the cycle length (by keeping the same initial 235U enrichment), which enables more compact and transportable HPMR core designs. However, using UN fuel decreases the control element worth [by up to 20% for the Special Purpose Reactor (SPR) and 5% for HP-MR] and is up to 80% more costly. Increasing 15N enrichment can further decrease the initial 235U enrichment requirement and increase the control element worth but is more costly. Compared to fast-spectrum HPMRs fueled with solid pellet fuels, retrofitting UN fuel is more suitable for thermal-spectrum HPMRs fueled with TRISO fuel compacts, where the neutron spectrum hardening caused by using UN is less significant.