Pin Power Research Articles

Development of deep convolutional neural network for prediction of cycle maximum pin power peaking factor in pressurized water reactor

A deep convolutional neural network (DCNN) model based on the SE-ResNet architecture and AlphaGo’s policy architecture is developed to predict the cycle maximum pin power peaking factor (PPPF) for a pressurized water reactor core. The DCNN model is structuralized to predict an output, the PPPF, corresponding to a unique input set by removing the pooling layer in between the blocks and introducing a padding for a quarter core input model. The developed DCNN model is trained on 17,000 randomly generated loading patterns (LPs) of an OPR1000 core and tested on 2,000 LPs. The accuracy of the newly developed model for the cycle maximum PPPF prediction is on average 6.5 times higher against a fully connected neural network model having the equivalent number of trainable weights.

Enhancing transmutation rates of minor actinides using Zr and Y hydride and deuteride coatings in an LFR

The possibility of transmutation of minor actinides (MAs) in lead-cooled fast reactors provides an excellent opportunity to reduce the overall repository of radioactive waste. The present work seeks to augment the transmutation rates of six minor actinides (237Np, 241Am, 243Am, 243Cm, 244Cm, and 245Cm) by increasing neutron capture through the use of moderator coatings around the fuel elements. The reference design for the fuel assembly was based on the design concept of ALFRED, the European Lead Cooled Demonstrator Reactor, and consists of 127 fuel elements loaded with 5 wt.% of minor actinides (MAs) homogeneously mixed with MOX (mixed oxide) fuel. Four different moderators (ZrH1.6, ZrD1.6, YH2, and YD2) with five different coating thicknesses (0.01 cm, 0.02 cm, 0.03 cm, 0.04 cm, and 0.05 cm) were selected for this study and variation in kinf, transmutation rates and fuel cycle parameters (employing the linear reactivity model or LRM) were determined. Results showed that both the addition of a moderator and the increase in moderator thickness improved transmutation rates but had a negative impact on fuel cycle parameters. ZrH1.6 resulted in the highest increase in transmutation rate, but also incurred the largest penalties in fuel burnup, and cycle length. For ZrH1.6 coatings of 0.01 cm, the obtained transmutation rates were 13.57%/y for 237Np, 12.6%/y for 241Am, and 9.26%/y for 243Am. The concentration of Cm isotopes increased over time due to neutron capture of Am, U, and Pu. Deuteride moderators minimized the decrease in fuel cycle parameters but compromised transmutation rates. The fuel temperature coefficient (FTC), energy dependent neutron flux, and relative pin power distribution of the reference design and the models were also determined in order to analyze the effect of moderator addition on these factors. Finally, a suggestion of the best model was put forth in consideration of both transmutation efficiency and fuel cycle parameters.

Validation and Neutronic Analysis of the CENM TRIGA MARK II SCALE Research Reactor Model: Computations Versus Experiment

This research analysis is mainly devoted to enhancing the safe and optimum use of the Center des Etudes Nucléaires de la Maâmora (CENM) TRIGA MARK II research reactor. To serve this purpose, various integral neutronic responses, such as the effective multiplication factor keff, the effective delayed neutron fraction βeff, the neutron flux distributions at the beam port entrances and the pneumatic transfer system bottom, the pin power peaking factors, the total excess reactivity, the control rod worth, the shutdown margin, and the worth of 11 fuel elements taken from different TRIGA core positions are calculated in order to evaluate the accuracy and the reliability of the developed TRIGA SCALE reactor model. The aim has been fulfilled by comparing the TRIGA SCALE results with those obtained by the MCNP TRIGA model, as well as with some recent experimental measurements from 2021. In general, all the obtained results reveal a good consistency between the SCALE and MCNP TRIGA models studied in this paper. The results analysis indicates also that the B-2 fuel element (Ring B) is the hottest rod among the 101 fuel rods existing in the TRIGA reactor core, which releases a maximum power of 31.67 kW. Furthermore, the total control rod worth, the total core excess reactivity, and the shutdown margin results are also closer to the experimental measurements.

Modeling and simulation of VERA core physics benchmark using OpenMC code

Detailed analysis of the neutron pathway through matter inside the nuclear reactor core is exceedingly needed for safety and economic considerations. Due to the constant development of high-performance computing technologies, neutronics analysis using computer codes became more effective and efficient to perform sophisticated neutronics calculations. In this work, a commercial pressurized water reactor (PWR) presented by Virtual Environment for Reactor Applications (VERA) Core Physics Benchmark are modeled and simulated using a high-fidelity simulation of OpenMC code in terms of criticality and fuel pin power distribution. Various problems have been selected from VERA benchmark ranging from a simple two-dimension (2D) pin cell problem to a complex three dimension (3D) full core problem. The development of the code capabilities for reactor physics methods has been implemented to investigate the accuracy and performance of the OpenMC code against VERA SCALE codes. The results of OpenMC code exhibit excellent agreement with VERA results with maximum Root Mean Square Error (RMSE) values of less than 0.04% and 1.3% for the criticality eigenvalues and pin power distributions, respectively. This demonstrates the successful utilization of the OpenMC code as a simulation tool for a whole core analysis. Further works are undergoing on the accuracy of OpenMC simulations for the impact of different fuel types and burnup levels and the analysis of the transient behavior and coupled thermal hydraulic feedback.

Neighbor-informed burnup correction for gadolinia depletion in pin-homogenized core calculation

The neighbor-informed burnup correction (NIBC) method is proposed to improve the accuracy of gadolinia depletion in pin-wise two-step calculations. The inability to capture severe heterogeneous depletion behaviors inside the gadolinia absorber with pin-homogenized quantities leads to inaccurate prediction of the burnup of the gadolinia pins, which is worsened throughout depletion by positive feedback between burnup and power. As a practical means to resolve this problem, the NIBC method represents the burnup of a gadolinia fuel pin using the average burnup of its neighboring ordinary fuel pins. This method is implemented in a GPU-based pin-wise two-step nodal code VANGARD. Verification results demonstrate significant improvements in accuracy. In the APR1400 two-dimensional core depletion problem, the maximum and RMS pin power errors are reduced from 8% and 2.1% to 2.7% and 0.7%, respectively. In a small modular reactor core depletion problem, those are reduced from 6% and 2% to 3% and 1%, respectively.

Electrically active traps in 4H-silicon carbide (4H-SiC) PiN power diodes

Electrically active traps in 4H-silicon carbide (4H-SiC) PiN power diodes

A Neutronics Study of the Initial Fuel Cycle Extension in APR-1400 Reactors: Examining Homogeneous and Heterogeneous Enrichment Design

AbstractIn this study, we conducted a neutronics analysis of a soluble-free-boron APR-1400 reactor core. Our goal was to explore the possibility of extending the fuel life cycle from 18 to 24 months. To achieve this, we examined both homogeneous and heterogeneous fuel enrichment designs while maintaining the original fuel geometries of the reactor. The proposed fuel enrichments for the homogeneous reactor core were 3.0%, 3.5%, 4.0%, 4.5%, and 4.95%. For the heterogeneous reactor core, they were (3.6%, 4.0%, 4.5%, and 5.0%) and (3.4%, 3.8%, 4.5%, and 5.0%). To suppress excess reactivity at the beginning of the fuel cycle (BOC), we applied an Integral Fuel Burnable Absorber as a thin coating layer on the outer surface of the fuel pellets. The coating was distributed uniformly throughout the core. In addition, we also studied the use of selected cladding materials as a replacement for the conventional zircaloy used in the fuel rods. The neutronics calculations for the modified APR-1400 core configuration were performed using the Serpent 2.1.31 Monte Carlo reactor physics code. We evaluated the behavior of the APR-1400 core by analyzing the effective multiplication factor, flux spectrum, pin power distribution, and radial power profile. Both the homogeneous and heterogeneous cores were compared with the reference APR-1400 core configuration. Our results indicate that it is possible to extend the fuel cycle to up to 24 months in both the homogeneous and heterogeneous cores. We also conducted a thorough analysis of the initial cycle for heterogeneous cores to consider more realistic scenarios.

Sensitivity neutronic analysis of accident tolerant fuel concepts in APR1400

Zirconium-based cladding materials used in nuclear power plants have been shown to pose safety concerns due to their susceptibility to oxidation and hydrogen production during high-temperature conditions, such as loss-of-coolant accidents. In order to address these concerns, alternative cladding materials that are more resistant to these conditions are being explored. This study investigates the potential of using silicon carbide (SiC) and iron-chromium-aluminum (FeCrAl) cladding materials to improve the safety and efficiency of nuclear power plants, particularly in the context of accident tolerant fuel (ATF) technology concepts. The study utilized Serpent Monte Carlo 2.31 to conduct neutronics analyses on the APR-1400 reactor, assessing the impact of SiC and FeCrAl cladding materials on fuel reactivity, neutron spectrum, Pu-239 inventory, and pin power distribution. A sensitivity analysis was conducted by adjusting parameters such as fuel enrichment, cladding thickness, and pellet diameter to determine the effects of the new materials on design parameters. Results of the analysis revealed that, by adjusting design parameters, SiC and FeCrAl materials could be used to meet the cycle length of the original fuel-cladding (zircalloy). Furthermore, using SiC cladding resulted in a hardening of the neutron spectrum, which increased the production of actinides such as plutonium. By analyzing the effects of changing parameters, including fuel enrichment, fuel geometry, and cladding material, the study concludes that SiC is a feasible substitute for conventional zirconium cladding. The findings of this study have practical implications for the safety and efficiency of nuclear power plants. The use of SiC and FeCrAl cladding materials could significantly reduce the risks associated with loss-of-coolant accidents and other high-temperature conditions, thereby improving the safety and reliability of nuclear power plants.

Conceptual design of the double tube burnable poison for the next generation small modular reactor

In this study, a new application of an innovative burnable poison (BP), which is called Double Tube Burnable Poison (DTBP), is introduced. Two design cincepts of the DTBP were tested in a small modular reactor (SMR) to achieve a longer operating cycle with low soluble boron concentration. These two concepts can be loaded in different locations in the fuel assembly, which increases the degree of freedom of nuclear design. The flexibility of this design allows it to increase its poison effect with the ability to control the depletion speed of its absorber materials. The DTBP pins were combined with the Erbia pins to achieve a flat k-infinite letdown curve with a significant reduction of the excess reactivity. The assembly and core calculations were performed by DeCART2D and MASTER codes. The results showed that the combination of DTBP and Erbia core achieved a cycle length of more than 3.5 Effective Full Power Years. Also, the highest value of Critical Boron Concentration was reduced to 309 ppm within the design limitations of the moderator temperature coefficient and the maximum pin power peaking. The performance of this option is the best among the other options including the conventional BP designs.

Uncertainty Propagation in UAM Time-Dependent Neutronics PWR Studies

A subset of pressurized water reactor (PWR) studies from Exercise II-2 of the Uncertainty Analysis in Modeling for Light Water Reactors (UAM-LWR) benchmark study has been performed to quantify the importance of both nuclear data uncertainties and manufacturing uncertainties in an assembly depletion calculation and a mini-core rod movement transient. The depletion study of the 15 × 15 PWR assembly using the SAMPLER and TRITON modules of the SCALE code system revealed a maximum uncertainty in keff of 0.49% for fresh fuel, decreasing to 0.38% at the midpoint of the fuel cycle and rising back to 0.48% at the end of the fuel cycle. Uncertainties and correlations of various homogenized cross sections and other group constant data, such as keff, have been determined, and the effect of randomly applied manufacturing uncertainties was found to be largely negligible relative to nuclear data uncertainties for bulk lattice parameters. However, for local parameters, such as the pin power factors, assembly discontinuity factors, and diffusion coefficients, the effects from manufacturing uncertainties were appreciable and sometimes dominant. Nuclear data uncertainties were found to be the dominant contributors to uncertainty in the isotopic composition of the overall assembly, with the exception of very early in the fuel cycle, where manufacturing uncertainties such as perturbations to the fuel density and pin radius made nonnegligible contributions to total uncertainty. The contribution of manufacturing uncertainties to isotopic uncertainties was nonnegligible at a pin-by-pin level, but still smaller than the contributions from nuclear data uncertainty. Studies of the PWR mini-core rod movement transient using homogenized data from the SCALE models in the PARCS diffusion code showed little difference between the tested modeling approaches and demonstrated that nuclear data uncertainties dominated the manufacturing uncertainties in the global figures of merit considered, such as the equilibrium core boron concentration, the maximum core power factor, and the maximum reactivity insertion. For local effects, such as maximum pin power during the transient, the randomly applied manufacturing uncertainties were dominant. It was found in general that for global system properties, nuclear data uncertainties made significantly larger contributions to total uncertainty, whereas for local parameters the impact of manufacturing uncertainties was at least nonnegligible, and for some parameters, dominant.

Development and validation of multiphysics PWR core simulator KANT

KANT (KAIST Advanced Nuclear Tachygraphy) is a PWR core simulator recently developed at Korea Advance Institute of Science and Technology, which solves three-dimensional steady-state and transient multigroup neutron diffusion equations under Cartesian geometries alongside the incorporation of thermal-hydraulics feedback effect for multi-physics calculation. It utilizes the standard Nodal Expansion Method (NEM) accelerated with various Coarse Mesh Finite Difference (CMFD) methods for neutronics calculation. For thermal-hydraulics (TH) calculation, a single-phase flow model and a one-dimensional cylindrical fuel rod heat conduction model are employed. The time-dependent neutronics and TH calculations are numerically solved through an implicit Euler scheme, where a detailed coupling strategy is presented in this paper alongside a description of nodal equivalence, macroscopic depletion, and pin power reconstruction. For validation of the steady, transient, and depletion calculation with pin power reconstruction capacity of KANT, solutions for various benchmark problems are presented. The IAEA 3-D PWR and 4-group KOEBERG problems were considered for the steady-state reactor benchmark problem. For transient calculations, LMW (Lagenbuch, Maurer and Werner) LWR and NEACRP 3-D PWR benchmarks were solved, where the latter problem includes thermal-hydraulics feedback. For macroscopic depletion with pin power reconstruction, a small PWR problem modified with KAIST benchmark model was solved. For validation of the multi-physics analysis capability of KANT concerning large-sized PWRs, the BEAVRS Cycle1 benchmark has been considered. It was found that KANT solutions are accurate and consistent compared to other published works.

Burnup Characteristics of Transuranic Fuel With Burnable Poison in Dual Cooled Annular Rod of VVER-1000

Abstract The potentiality of Transuranic (TRU) fuel as dual cooled annular fuel rods with different types of burnable absorbers (BAs)- integrated burnable absorbers (IBAs), burnable poison rods (BPRs), coated absorbers, etc. in the hexagonal assembly of VVER-1000 was studied. Annular 7, Annular 8, and Annular 9 models were taken for various combinations of TRU fuel and BAs. Planned models were simulated in Monte Carlo particle simulation code OpenMC and lattice physics deterministic code Dragon Version5. Burnup-dependent multiplication factors (MFs) were simulated for 3000 effective full power days (EFPDs). Reactivity was calculated by taking 3% neutron leakage. The study showed that both the MF and reactivity for TRU fuel are significantly higher than conventional UO2 fuel. Linear reactivity model (LRM) was applied to find out cycle burnup, discharge cycle burnup, and cycle length. High cycle burnup, discharge burnup, and cycle length have been observed for TRU fuel compared to UO2 fuel. Burnup-dependent atomic concentration graphs showed that slight burn of Np-237, constant concentration for Cm-244, a slight increase for Am-242, and linear burnout of BAs- Gd-155, Gd-155, and Er-167. A lower concentration of Xe-135 has been observed for TRU fuel. Pin power distribution and energy-dependent neutron flux for different models and BAs combinations are also included in this study.

High-Performance and High-Fidelity GPU-Based Monte Carlo Solutions to the BEAVRS Benchmark

The BEAVRS benchmark is solved by PRAGMA, the graphics processing unit (GPU)–based continuous-energy Monte Carlo code. The solutions consist of the detailed simulation results for the two cycles that involve the reactivity and pin power distribution information for the zero-power physics tests and depletion. Primary results at hot zero power, such as the critical boron concentration at various rodded conditions, control rod bank worth, isothermal temperature coefficients, and assemblywise detector signal, are compared with the measured data. Core-follow calculations are performed with varied power, and the resulting boron letdown curves are compared with the measured one. Hot full-power depletion is also performed and the resulting pinwise power distributions of cycle 1 are compared with the nTRACER results. The comparison with the measured data and also with the nTRACER results demonstrates the high solution fidelity of PRAGMA. In all the calculations, PRAGMA uses a tremendously large number of histories, ranging from up to hundreds of millions per cycle, that are used to fully exploit the massive parallel computing capacity of GPUs. The execution time of the entire core-follow calculation with about 30 burnup steps takes less than 16 h on a single rack of computing nodes mounted with 24 gaming GPUs, which represents considerably high Monte Carlo core calculation performance.

Geometry Extension and Assemblywise Domain Decomposition of nTRACER for Direct Whole-Core Calculation of VVERs

The capability and performance of the hexagonal version of the nTRACER direct whole-core calculation code are enhanced for VVER applications by extending the geometry-handling features and also by implementing assemblywise parallelization of the planar method of characteristics (MOC) calculation with higher-order scattering. The geometry-handling methods for the VVER hexagonal geometry having various special constituents are presented with detailed illustrations. The assemblywise domain decomposition (ADD) scheme is established under the hexagonal coarse-mesh finite difference formulation, which is exploited to update the incoming angular flux needed for the ADD parallelization. The solution accuracy and parallel performance are assessed for various hexagonal core problems, including the VVER benchmarks. It is shown that the hexagonal geometry solutions of nTRACER match with the reference Monte Carlo solutions within about 50 pcm in reactivity and 1% in pin power distribution and that the hexagonal ADD can reduce the computing time of the planar MOC calculation by up to 53% when compared to the anglewise parallelization.

Preliminary design of soluble boron free small modular reactor core based on hexagonal lattice

This study made an effort to design a novel high-power density SMR core based on a hexagonal lattice for the soluble boron free environment. The VVER-1000 reactor core was converted to an SMR core with a “halved scale”. The core design process is divided into two parts; Part I: assembly-level analysis and Part II: whole-core analysis. Important design parameters such as core dimensions, number, and arrangement of fuel assemblies, fuel enrichment, appropriate burnable poison, and placement of burnable poison pins have been studied in Part I. Also, by designing an optimized burnable poison pattern, the excess reactivity and acceptable radial pin power distribution were treated without soluble boron. WIMS, and PARCS codes have been investigated for fuel assembly and whole core analysis. Finding the most appropriate fuel-loading pattern involved an iterative manual optimization between the core loading pattern and FA design to assess the acceptable neutronics and thermal-hydraulics core features and satisfy the safety margins.The design of high-power density core requires accurate prediction of the thermal–hydraulic parameters to ensure safe operation. In Part II coupled 3-D neutronic/thermal–hydraulic analysis has been done to identify the main safety challenges of the developed core. To access the proper discharge burn-up, an initial enrichment of 6.6% U235 was determined for the developed core. The developed core maximum discharge burn-up is equal to ∼49 Mwd/KgU at full power operation. Radial-zoning burnable poison is investigated in the seek of a proper pattern to limit power peaking.

Pin power reconstruction for hexagonal geometry in nodal neutronics program Ants

The pin power reconstruction methodology for hexagonal geometry in the multigroup nodal neutronics program Ants is described and verified against full core Serpent Monte Carlo solutions in four VVER benchmarks. In order to provide the best possible reference solution, Serpent uses the same nuclear data libraries and simulation options both to generate the required group constants for Ants and to solve the full core reference solution. The root mean square of the relative differences in 2D pin powers between Ants and Serpent was between 0.83% and 1.11% in the four benchmarks, while the root mean square of the relative differences in the 2D intra-assembly peaking factor (Kk) was between 0.54% and 0.71%. Increasing the number of energy groups in the nodal solution tended improve the accuracy of the nodal solution as did the use of critical spectrum condensed group constant data instead of infinite spectrum group constants.

Verification and validation of the high-resolution code nTF with VVER-1000 hot zero power monte carlo calculations and experimental data

Novel codes allow the prediction of parameters with increased resolution, in comparison to the conventional approach, sometimes using meshing smaller than the size of the fuel pin. Optimally, complex geometric structures, like the ones included in the VVER core e.g. corner stiffeners, would be modeled explicitly by novel neutronic codes. In this paper, the performance of the deterministic neutron transport code nTF is studied for VVER geometries. The code is verified standalone with Monte Carlo reference solutions for a range of Hot Zero Power (HZP) models, from 2D pincells to the 3D full core of a VVER-1000, described in the X2 benchmark. Several modeling options and approximations are studied to optimize the nTF X2 3D core model. The updated version of nTF, including all necessary modifications to improve its VVER modeling capabilities, presents good agreement with the reference solutions for the simple VVER configurations and the critical state of the X2 core. Specifically for the full core critical state, the RMS of the pin power difference is 1.5% and the eigenvalue difference 74 pcm with respect to a Monte Carlo reference solution. Any resulting discrepancies are analyzed, the more significant ones with the use of the lattice code CASMO5-VVER that employs similar methods as nTF. The optimized nTF X2 full core model is validated with the experimental data included in the X2 benchmark, produced by HZP start-up tests. The differences of the nTF results with the measured and simulated data of the X2 benchmark remain within the accuracy limits set by the utilities and the literature of deterministic code verification for most cases. The related discrepancies are discussed in detail.

An iDTMC-based Monte Carlo depletion of a 3D SMR with intra-pin flux renormalization

Improved deterministic truncation of the Monte Carlo (iDTMC) solution method has been applied to neutronic depletion analyses of a nuclear reactor in this work. The theoretical background on the iDTMC method and material depletion calculation is briefly discussed, and the concept of an intra-pin renormalization scheme is presented to determine the detailed power and flux profiles with the pin-wise homogenized iDTMC solutions. The iDTMC method is applied and evaluated in a 100 MWth 3-dimensional small modular reactor (SMR) problem. Burnup-dependent multiplication factors, pin power distributions, and material densities are estimated by the iDTMC method and compared with the reference and standard Monte Carlo solutions. Numerical performance is investigated in terms of real standard deviation, root-mean-square and relative errors, computing time, and figure-of-merit. The numerical results demonstrate that the iDTMC method is superior to the standard MC method in estimating the burnup-dependent reactor eigenvalue and 3D pin-wise power distribution.

Nuclear and thermal hydraulic calculation of a representative I2S-LWR first core

The Integral Inherently Safe Light Water Reactor (I2S-LWR) concept developed by Georgia Tech is a novel PWR reactor delivering electric power of 1000 MWe while implementing inherent safety features typical for Generation III+ small modular reactors. The main safety feature is based on integral primary circuit configuration, bringing together compact design of the reactor core (121 fuel assembly), control rod drive mechanism (CRDM), 8 primary heat exchangers (PHE), 4 passive decay heat removal systems (DHRS), 8 pumps, and other integral components. A high power density core based on silicide fuel and APMT (FeCrAl) stainless steel cladding is selected to achieve a high thermal power. Initial representative first core nuclear design is proposed by Westinghouse. Full core 3D depletion calculation was performed using PARCS code. The cross section library is prepared using FA2D code and verified using Polaris sequence from SCALE 6.2 beta5. The axial and radial reflectors are assumed to be homogeneous water-APMT mixtures. The axial reflectors are both assumed to be 12 inch (30.48 cm) sections composed of 30% APMT steel by volume. The radial reflector is assumed to be 90% APMT steel by volume. The reflector constants were calculated using SCALE TRITON sequence. The thermal hydraulic part of the model is based on COBRA subchannel code coupled to PARCS code. Initial depletion calculation is based on one thermal hydraulics channel per fuel assembly approach. The hot fuel assembly is determined using separate pin-by-pin COBRA subchannel model and pin power reconstruction data from PARCS. The objective of the paper is demonstration of LWR design methodology applicability to silicide fuel and identification of possible improvements in the first core design.

The neutronics effect of TRISO duplex fuel packing fractions and their comparison with homogeneous thorium‐uranium fuel

To improve the performance of the thorium fuel, the duplex fuel potential for the micro-sized high-temperature reactor (HTR) using TRISO fuel has previously been investigated. The objective of this work is to expand on past research by delving deeper into several crucial components of the microheterogeneous thorium-uranium TRISO fuel compact. The MCNPX simulation was used to compare the neutronic performance of TRISO duplex fuel to that of homogeneous (Th,U)O2 fuel at various packing fractions and 235U enrichment. Variations in duplex fuel packing fraction and 235U enrichment were shown to have a significant effect on the neutronics of TRISO fuel. The fuel rod with 20% TRISO packing fraction has the highest kinf value at BOC, fastest burnup rate, fastest 235U depletion, and longest cycle length. As the packing fraction increases, these values however deteriorate in the opposite direction. Furthermore, the conversion ratio increases and the Doppler effect becomes stronger. The findings of evaluating duplex fuel rods at the fuel block level differed. The 40% packing fraction fuel block model achieves the lowest reactivity swing, the strongest Doppler effect, and the longest cycle length. As its packing fractions reduced to 30% and 20%, the values changed in the opposite direction. In comparison to the duplex fuel design, the homogenous TRISO fuel has a lower achievable discharged burnup and shorter cycle length. However, at the beginning of its cycle, it has a stronger doppler effect and lower pin power peaking. Despite the fact that the duplex's power peaking value drops with increasing burnup and becomes comparable to homogeneous fuel, these are some of the challenges and concerns that should be addressed in future research. Despite this, the overall result indicates that the duplex fuel design has promising potential for application in a micro-sized HTR.