Coolant Temperature Distribution Research Articles

Development of a liquid metal subchannel code applied to ocean conditions and study on the effect of core geometry parameters and ocean motions on flow and heat transfer characteristics

The Generation IV International Forum (GIF) proposes a type of liquid metal reactor, due to its thermal properties and efficient heat transfer in a relatively small system. This special heat transfer characteristic allows the liquid metal reactor to be modular in design and flexible in installation on a nuclear-powered ship. Accurate prediction of the coolant temperature and flow distribution between subchannels is important to ensure nuclear safety in the complex ocean environment. In this study, an ocean version of subchannel code is developed by adding the motion terms in the momentum equations for the liquid metal reactor core based on the previously land-based subchannel code. Since the research on the characteristics of flow resistance and heat transfer mechanism under ocean conditions is not mature, the original empirical model is still used in the ocean version code. The constitutive models of the liquid metal are analyzed and incorporated into the ocean code. The code has been verified and validated by comparing with Computational Fluid Dynamics (CFD) simulation results, and experimental data. The effect of geometric parameters such as rod Pitch to Diameter ratio (P/D), Rod to Wall gap (RTW) is studied in the liquid metal lead 7 rod bundles. In addition, the effect of heaving start-up and shutdown motions between liquid metal lead and water is discussed in the 5 × 5 bare rod bundles. In general, the modified subchannel analysis code can well complete the calculation of liquid metal reactor under ocean conditions, as well as provide support for the design and safety analysis of a liquid metal cooled fast reactor.

Design of alveolar biomimetic enhanced heat transfer structure for cylindrical lithium battery pack

Aiming to address the issue of thermal runaway in high energy density batteries during high charge and discharge rates, a heat management system for an alveolar-like honeycomb liquid-cooled power battery was designed. This system is inspired by the bionic concept of cell cooling in biological tissues, with cylindrical batteries serving as cells and cooling channels resembling blood vessels. The cooling system for a battery pack consisting of 24 cylindrical cells was designed and implemented. Numerical simulations were conducted to optimize and study the effects of inlet diameter, branch angle, flow rate, and spiral baffle on coolant flow uniformity and temperature distribution within the battery pack. The results indicate that as the inlet diameter of the fractal channel increases from 2 mm to 20 mm, the coolant flow uniformity improves and the pressure drop decreases. However, the influence of inlet diameter diminishes after reaching a certain limit. When the inlet diameter increases from 10 mm to 20 mm, the relative standard deviation of the velocity at the branch outlet decreases only slightly, from 0.9 % to 0.8 %. Increasing the branch angle from 30° to 75° leads to an increase in the relative standard deviation of velocity, from 0.79 % to 1.24 %. Incorporating a spiral baffle into the internal flow channel of honeycomb cooling significantly enhances temperature uniformity within the battery pack while reducing maximum temperature levels considerably. Furthermore, a smaller pitch for the spiral baffle yields better cooling performance for the immersed liquid cooling thermal management system. After optimization, the maximum temperature difference is only 1.27 °C across the surface of the battery pack, achieving a remarkable temperature uniformity level of just 0.11 %.

MC/sub-channel coupling for steady state and transient simulation of Xi’an Pulsed Reactor

The Monte Carlo (MC) code RMC and sub-channel code CTF are coupled for Xi’an Pulsed Reactor (XAPR) pulse and depletion simulation. A python script is developed to handle data exchange through files. $3.45 pulse and $2 pulse are simulated with the pulsed state core. The pulse peak power and the Full Width at Half-Maximum (FWHM) results are compared with experiments as a validation and good agreement is achieved. Detailed 3-D power and temperature distributions are also obtained. Results show that the core power peak is close to the pulse rod where the reactivity is introduced. The radial power peak of each rod is at the boundary because of the self-shielding effect. The rod temperature distribution follows the same trend with the power, and the coolant temperature is not changed during the pulse period of about 0.12 s, which suggests that the heat transfer plays a negligible role in such a short time. For the steady state core, depletion simulation is carried out for the lifetime of the first fuel cycle of 120 Effective Full Power Days (EFPD). Validation is done by calculating the temperature distribution and the differential worth of the regulating rod at 0 EFPD, which both agree well with experiments. Results show that the power distribution is almost unchanged over time, only slightly more flat, indicating the material is not greatly depleted. The temperature distribution of the core generally agrees with the power distribution, except the radial temperature peak of each rod is at the center because heat is conducted outwards at steady state. Detailed coolant temperature distribution is obtained thanks to the utilization of CTF. Temperatures at the assembly boundary and near the central water chamber is noticeably lower than the other part, which is not shown in the parallel multi-channel models.

Thermal and hydraulic characteristics analysis of horizontal lead-bismuth reactor assembly based on sub-channel analysis code

In this paper, the self-developed subchannel analysis program SACOS-PB was used to analyze the thermal hydraulic characteristics of 127 rods in horizontal lead-bismuth reactor under steady state and typical ocean transient conditions. The results of the steady-state analysis showed that the coolant flow and temperature of the vertical assembly were symmetrically distributed, the coolant flow of the horizontal assembly gradually increased along the gravity direction, the temperature rose first and then fell, and the peak coolant temperature of the cross-sectional channel was about 365 °C. Three typical ocean transient operating conditions, inclined, undulating and swing, were also carried out. The results showed that the maximum temperature of the coolant at the exit of the assembly increased with the increase of the transverse heeling angle and the decrease of the trim angle after the incline motion changes the gravity component. Under the undulating and swing conditions, the coolant hydrothermal parameters fluctuated periodically, and the parameter fluctuation period is 10s in line with the oceanic conditions. In the above transient conditions, the peak coolant temperature at the assembly outlet increases and exceeds 365 °C. Compared with the pitching motion on the Y-axis, the rolling and tumbling motion on the X-axis had a significant effect on the coolant flow rate and temperature distribution. The calculation results can provide support for the analysis study of the safety characteristics of lead-bismuth power units under marine conditions.

PWR natural circulation at a small break LOCA

In this paper, a one-dimensional quasi-steady model has been developed to analyse the potential impacts of reduced primary coolant inventory on natural circulation. This model focuses on the development of expressions for the coolant inventory, temperature distributions, and mass flow rate. The model encompasses a range of fluid flow scenarios, including single-phase, combined single-and-two-phase, and two-phase natural circulation, as well as the phenomenon of reflux condensation. The model is also utilized to examine the impact of reactor core power on natural circulation parameters. Comparisons with experimental findings and earlier research are made, highlighting a satisfactory level of agreement with the analytical findings. Based on the findings of this study, it has been observed that within the range of primary coolant inventory of 100–60% and core decay power of 1.5–5% of full power, natural circulation in a PWR type system has the potential to serve as an effective mechanism for transferring core decay heat to the secondary loop. The flow rates of natural circulation in pressurized water reactors (PWRs) are observed to have a significant dependence onthe amount ofliquid inventory, whereas their relationship with power level is comparatively less pronounced. The flow rate of a pressurized water reactor (PWR) system tends to exhibit its highest value within a specific range of inventory, typically falling between 60% and 80%. The transition between the different modes of natural circulation is observed to exhibit a smooth transition, with varying system inventories.

Optimization Turbine Blade Cooling by Applying Jet Impingement Cooling Channels

Abstract The aim of this paper is to optimize turbine blade cooling channels by applying jet impingement Method. The selection of experiment data for NASA 3CX turbine blade, and 3D model using solidworks software and create computational fluid dynamics (CFD) simulations used to model the coolant flow and temperature distribution in the vane, while experimental testing can validate the CFD results and provide additional insights into the cooling system's performance., ANSYS FLUENT code was used as a CFD solver, and ANSYS ICEM-CFD was used for mesh generation. MATLAB code is used for calculation using experiment data and this was helpful for simulations. Heat transfer conjugation analysis bases SST shear stress analyses K-ω turbulent model. The results conclude that providing additional information about the cooling channels and how they differ in the studies being compared. The results demonstrate that the cooling channels' hydraulic diameter decreases by a significant percentage (up to 49.70%–69.55%) as they are drawn to the trailing edge of the blade. This can have a significant impact on the heat transfer coefficients and the performance of the cooling system. The pressure side of the turbine blade is observed to follow the Hylton Model, while the current study predicts a large over-anticipated heat transfer coefficient around the Turbine blade head and on the bulk of the suction side. In terms of average heat transfer coefficient, the two models differ by 23.36%. The authors found that the cooling effectiveness for the Optimized jet impingement model is 0.4892 for whole blade and compared it with the cooling effectiveness for the optimized jet impingement model, which is 0.6936, The results of the comparison between the base model and the optimized jet impingement model suggest that the optimized model has a significantly higher cooling effectiveness. The increase in cooling effectiveness of 29.46% for the whole blade and 28.823% for the trailing edge indicates that the optimized jet impingement design provides improved cooling performance. These results highlight the importance of considering optimized cooling designs for turbine blades to maintain efficient and safe operation.

Investigation on Applicability of Subchannel Analysis Code ASFRE to Thermal Hydraulics Analysis in Fuel Assembly With Inner Duct Structure of Sodium Cooled Fast Reactor

Abstract In the design study of an advanced sodium-cooled fast reactor (advanced-SFR) in Japan Atomic Energy Agency (JAEA), the use of a specific fuel assembly (FA) with an inner duct structure called fuel assembly with an inner duct structure (FAIDUS) has been investigated to enhance safety of Advanced-SFR. Due to the asymmetric layout of fuel rods by the inner duct, it is necessary to estimate the coolant temperature distribution to confirm feasibility of FAIDUS. In JAEA, an in-house subchannel analysis code named thermal-hydraulic analysis of asymmetrical flow in reactor elements (ASFRE) has been developed as a FA design tool. For the typical FAs, the numerical results of ASFRE had been validated by comparisons with experimental data, in the previous study. As for the FAIDUS, confirmation of validity of the numerical results by ASFRE was not enough because the reference data on the thermal hydraulics in FAIDUS have not been obtained by the mockup experiment, yet. In this paper, therefore, the code-to-code comparisons with numerical results of ASFRE and those of an in-house computational fluid dynamics (CFD) code named single-phase thermal-hydraulic analysis in an irregular rod array layout (SPIRAL) were applied to make further discussion on applicability of ASFRE to the thermal hydraulics analysis in FAIDUS. Thermal hydraulic analyses of a typical FA and FAIDUS at high and low flowrate conditions were conducted. The applicability of ASFRE was indicated through the confirmation of the consistency of mechanism on appearance of the specific temperature distributions between the numerical results by ASFRE and those by SPIRAL. In addition, the necessity of modification on the empirical constants in numerical model of ASFRE to improve the predictive accuracy was indicated.

Coolant temperature distribution in the model of the fuel assembly peripheral area

To evaluate the interchannel temperature transfer, the test section was created, the working section of which is a model of two neighboring cells of the peripheral region of the fuel assembly. The experimental setup is equipped with three fuel rod simulators 500 mm long. During the work, the fuel rods were connected in turn to the power source. This made it possible to determine the features of the temperature distribution in the model with uneven energy release. The power of each fuel rod simulator was up to 2000 W. A movable thermocouple and an IR camera were used for measurements.

Numerical study on temperature distribution and mixing effect by spacer grid in 2×1 subchannel

A major function of Mixing Vane Grids (MVGs) in Light Water Reactors (LWRs) is mixing. Strong mixing results in more uniform coolant temperature distribution, which often leads to improvement in Critical Heat Flux (CHF). In subchannel analysis, a typical mixing coefficient β is often used to represent MVG mixing performance. Despite of this, due to the periodical mixing caused by MVGs with complex geometry under a wide range of operating conditions, a simple constant mixing coefficient cannot accurately reflect the true characteristics of MVG. In order to explore the fundamental mixing characteristics introduced by mixing vanes, a 2×1 subchannel system was selected to study the mixing mechanism. Due to the wall effect, different boundary settings in the 2×1 subchannel gaps were analyzed and compared with 5×5 simulation results to get the most similar simulation results. To evaluate the effect of grids on mixing, the temperature distributions and thermal non-uniformity (Et) between bare rod bundles and grid-enclosed rod bundles were compared. A new concept of local mixing and subchannel mixing was proposed, by comparing the local Et (defined based on mesh size used in CFD calculation), the subchannel Et (defined based on the subchannel area) and temperature contours. In subchannel analysis, lumped parameters could not accurately represent the true mixing caused by mixing vanes. They overestimated the mixing effect and neglected the important near wall mixing phenomenon.

Investigation of hybrid nanofluids effects on heat transfer characteristics in VVER-1000 nuclear reactor

In this study, the analysis of heat transfer characteristics of VVER-1000 nuclear reactor by using hybrid nanofluids as a coolants in subchannels in hot fuel assembly have been performed. Firstly, the criticality (keff) and fuel assembly power distributions of the VVER-1000 reactor by using Monte Carlo particle transport (MCNP) code to determine of the heat transfer characteristics with COBRA IV PC code have been examined. Secondly, the axial coolant temperature distributions and Departure from Nucleate Boiling Ratio (DNBR) values in hottest channel for Cu, CuO and TiO2 nanoparticles of the different volume fractions (0.05%, 0.01% and 0.1%) mixed with 0.1 vol % Al2O3 hybrid nanofluids have been analyzed. The obtained results by using hybrid nanofluids have been compared with the pure water results. The keff decreases with the increase of volumetric fractions of hybrid nanoparticles Cu, CuO and TiO2 mixed with 0.1% Al2O3 in the coolant. The maximum temperature has been reached by the püre water coolant at the end of the channel as 612.68 K whereas, the highest temperature change has been obtained in the case of 0.1% Al2O3 + 0.1% TiO2 as 617.35 K when evaluated among nanoparticle themselves. The results of analysis show that the heat transfer enhances with increasing hybrid nanoparticles volume fractions in the subchannel geometry. Also, the obtained minimum DNBR values using hybrid nanofluid haven't been exceed the safe operational limits; therefore the reactor operation would be safer.

Neural Network Modeling of the Solar Photovoltaic Thermal Module Performance

The design of photovoltaic thermal modules (PVT modules) features a nonuniform distribution of coolant temperature in the channel, and the solar cells (SC) that are in thermal contact with the PVT module channel are under different temperature conditions. The nonuniform distribution of the SC temperature causes undesirable effects that are complex and nonlinear, such as a drop of generated power and SC damage resulting from the occurrence of hot spots. The aim of the work is to develop a mathematical model for modeling the PVT module thermal and electrical performance based on a feedforward artificial neural network (FNN). A two-layer FNN with sigmoid hidden neurons and linear output neurons has been developed. The input layer is made up of the ambient temperature, coolant flowrate and environmental variables (the total insolation). The output layer represents the PVT module thermal and electrical efficiencies. The developed FNN was trained and adapted on the basis of modeling and an experimental database on the PVT module input and output parameters. The developed FNN was trained using the Levenberg–Marquardt algorithm. The mean absolute error achieved during the training is in the range from –0.319 to 0.448 for electrical efficiency and from –0.129 to 0.198 for thermal efficiency. The r.m.s error is 0.0678 for electrical efficiency and 0.0247 for thermal efficiency; the training time is 15 s or longer. An effective model has been developed that implements a new approach to modeling the performance of PVT modules based on artificial neural network algorithms with fairly close values of thermal and electrical efficiencies.

High-Fidelity Steady-State and Transient Simulations of an MTR Research Reactor Using Serpent2/Subchanflow

In order to join efforts to develop high-fidelity multi-physics tools for research reactor analysis, the KIT is conducting studies to modify the coupled multi-physics codes developed for power reactors. The coupled system uses the Monte Carlo Serpent 2 code for neutron analysis and the Subchanflow code for thermo-hydraulic analysis. Serpent treats temperature dependence using the target motion sampling method and Subchanflow was previously extended and validated with experimental data for plate-type reactor analysis. This work present for the first time the steady-state and transient neutron and thermo-hydraulic analysis of an MTR core defined in the IAEA 10 MW benchmark using Serpent2/Subchanflow. Important global and local parameters for nominal steady-state conditions were obtained, e.g., the lowest and highest core plate/channel power/temperature, the radial and axial core power profile at the plate level, and the core coolant temperature distribution at the subchannel level. The capabilities of Serpent2/Subchanflow to perform transient analysis with on-the-fly motion of the control plates were tested, namely with fast and slow reactivity insertion. Based on the unique results obtained for the first time at the subchannel and plate level, it can be stated that the coupled Serpent2/Subchanflow code is a very promising tool for research reactor safety-related investigations.

Neutronics and both analytical and numerical solutions for the rod centered subchannel thermal-hydraulic model

Abstract This work investigates the capacity to detect safety levels in pressurized water reactors using neutronic and thermal-hydraulic calculations (PWR). The main neutronic parameters, such as criticality calculations, reactivity calculations, burn-up calculations, and power distribution calculations, have been calculated by using MCNPX based on the Monte Carlo method. The hot channel was determined from the power distribution. On the hot channel, the rod-centered sub-channel model was used to investigate the main thermal-hydraulic parameters such as the fuel, cladding, and coolant temperature distribution in both axial and radial directions, the coolant pressure drop, and the departure from nucleate boiling ratio (DNBR). Two techniques were used to solve the rod-centered sub-channel model. MATLAB code was used for the analytical methodology, and COMSOL-Multiphysics computer software was used for the numerical methodology. The numerical technique has demonstrated high accuracy in determining the reactor safety level consistent with the analytical approach.

Thermal modelling and characteristic evaluation of electric vehicle battery system

Thermal modelling of Li-ion battery system used in electric/hybrid electric vehicles is carried out. The main highlight of this study is the numerical modelling of Li-ion battery using finite volume method with the adoption of realistic coupled heat and fluid flow process. Thermal characters of the battery system are analyzed with the change in parameters like flow Reynolds number, battery internal heat generation, the length to width ratio, and the conduction-convection related parameter at the interface of battery and air. In detail analysis of maximum temperature, heat flux variations, Nusselt number, friction coefficient, coolant temperature, and velocity distributions are carried out. Finally, the effect of channel width formed between two parallel placed battery cells is also carried. The numerical analysis performed reveals that the length to width ratio of battery does not impact the thermal performance of battery. At lower Reynolds number, the conduction-convection parameter plays a significant role in reduction of battery temperature avoiding thermal stresses developed. Channel spacing has a prominent role in variation of battery and coolant temperature. The battery surface temperature is largely affected by parameters considered. However, it is at higher input of conduction-convection parameter, Reynolds number, and channel spacing the thermal variation remains insignificant.

Analysis of the Accelerator-Driven System Fuel Assembly during the Steam Generator Tube Rupture Accident.

China is developing an ADS (Accelerator-Driven System) research device named the China initiative accelerator-driven system (CiADS). When performing a safety analysis of this new proposed design, the core behavior during the steam generator tube rupture (SGTR) accident has to be investigated. The purpose of our research in this paper is to investigate the impact from different heating conditions and inlet steam contents on steam bubble and coolant temperature distributions in ADS fuel assemblies during a postulated SGTR accident by performing necessary computational fluid dynamics (CFD) simulations. In this research, the open source CFD calculation software OpenFOAM, together with the two-phase VOF (Volume of Fluid) model were used to simulate the steam bubble behavior in heavy liquid metal flow. The model was validated with experimental results published in the open literature. Based on our simulation results, it can be noticed that steam bubbles will accumulate at the periphery region of fuel assemblies, and the maximum temperature in fuel assembly will not overwhelm its working limit during the postulated SGTR accident when the steam content at assembly inlet is less than 15%.

Application and Benchmarking of a Novel Coupled Methodology for Simulating the Thermomechanical Evolution of Sodium-cooled Fast Reactors Fuel Subassemblies

We employ a novel methodology to simulate the irradiation of Sodium cooled Fast Reactors (SFR) subassemblies. The proposed approach allows considering the coupling between key thermal–hydraulic and thermomechanical phenomena, which is usually neglected in traditional simulation methods. Comparisons between results obtained from coupled and uncoupled methodologies show that the latter can significantly over predict the stresses and the strains of highly irradiated fuel bundles. Additionally, we explore different options for the pin mechanical modeling. In particular, we show that a relatively simple beam-based finite element model for the fuel pins can be used to efficiently compute swelling and irradiation creep strains averaged over the cross section of the fuel claddings. However, if a prediction of the cladding swelling gradients is needed, the beam model must be replaced by a more computationally expensive 3D fuel pin model. An efficient method to estimate the fuel cladding temperature evolution based on a limited number of Computational Fluid Dynamics (CFD) simulations is also presented in this work, and its performance is evaluated. Finally, a numerical benchmark against the state-of-the-art coupled code system for the simulation of SFR subassemblies is presented. This benchmark shows that the proposed coupled methodology allows obtaining results that are in good agreement with the reference methodology. However, two clear advantages of the proposed methodology were identified. Firstly, the use of detailed CFD simulations in our methodology, compared to the lower resolution thermal-hydraulic model employed in the state-of-the-art approach. Secondly, considering the reduction of the coolant mass flow rate caused by the fuel bundle deformation, neglected in the pre-existing approach, which is shown in this work to have a potentially large impact on the coolant temperature distribution.

Simulation of an unprotected transient of the ALLEGRO reactor using the coupled neutronics/thermal-hydraulics system code KIKO3DMG/ATHLET3.0

The KIKO3DMG nodal reactor physics code and the ATHLET3.0 thermal-hydraulics system code have been extended in recent years for the analyses of fast reactors. This paper focuses on the capabilities of the coupled KIKO3DMG/ATHLET3.0 code in the case of an unprotected transient of the ALLEGRO demonstrator gas-cooled fast reactor. We show that the coupled code can simulate an asymmetrical control assembly withdrawal transient. The code provides adequate results by using 6-group constants instead of the widely used 24-group constants, and with this, the computation time of the transient is reduced by 95%. Moreover, we drove attention to the applicability of the point-kinetics methodology, which can be limited because of the approximate manner of the derivation of the corresponding parameters. Finally, we present a methodology to flatten the distribution of the outlet coolant temperature by applying flow restrictors.

Simulation of Heat and Mass Transfer in Wire-Wrapped Fuel Assemblies in the Anisotropic Porous Body Approximation

Results of the simulation of heat and mass transfer in wire-wrapped fuel assemblies in the anisotropic porous body approximation using the developed APMod software package are presented. The modifications introduced into the porous body model to make it suitable for wire-wrapped fuel assemblies are described. The predictions of thermal and hydraulic characteristics in the liquid-metal cooled experimental and model fuel assemblies according to this updated model are presented. An isothermal sodium flow in a Bundle 2A experimental 19-rod wire-wrapped assembly and uniform or nonuniform heating of the rods was studied. The predictions were compared with the experiments using the pressure difference across the assembly versus the coolant flowrate and the coolant temperature distribution in the assembly’s outlet section. The thermal–hydraulic characteristics in the BN-1200 reactor fuel assembly’s heated 19-rod fragment with its flow cross-section locally blocked in the central part calculated by the porous body model were compared with the predictions by the CONV-3D DNS code. Before their comparison, the distributions of local velocities, pressure, and temperature in an assembly cross-section calculated by the CONV-3D code were averaged over the averaging cells in the APMod software package. It is demonstrated that the APMod software package may be used to calculate parameters, which are averaged over a representative averaging cell, in a liquid-metal coolant flow in wire-wrapped fuel assemblies with an accuracy adequate for engineering applications.

Numerical simulation of temperature heterogeneity inside the AP1000 upper plenum and hot leg

The high-fidelity simulation for nuclear reactor has attracted more attentions recent years and it could provide much deeper understanding of the thermal hydraulic features in the reactor pressure vessel, promoting the reactor design optimization or operation guidelines making. The temperature heterogeneity in hot legs induces the deviation of reactor core power level estimation. In order to investigate the lateral flow at the reactor upper plenum and temperature heterogeneity in the hot legs of AP1000, a multi-scale coupling method was proposed with the computational fluid dynamics (CFD) and sub-channel analysis in this paper. The sub-channel analysis code, COBRA-EN, was used to calculate the coolant temperature distribution at core outlet and provide the accurate boundary conditions for CFD simulation. A detailed CFD analysis model from the core outlet to hot leg outlet was established. The main internal components including control guide tubes with eight large opening windows on it and support columns were kept, while other little components including control rod assemblies were omitted to decrease the total mesh quantity. Compared with the temperature field at core outlet, the maximum temperature difference at the entrance of the hot leg drops 10.0 K after the strongly mixing in the upper plenum. The hotter coolant from central fuel assemblies remains at the upper part of the hot leg, while the cooler coolant from peripheral fuel assemblies stays in the lower part of the hot leg. The temperature heterogeneity and its evolvement along the hot leg were analyzed.

Fuel performance analysis of BEAVRS benchmark Cycle 1 depletion with MCS/FRAPCON coupled system

A fuel performance (FP) analysis of the BEAVRS (Benchmark for Evaluation and Validation of Reactor Simulations) benchmark Cycle 1 depletion is performed using the MCS/FRAPCON coupled code system. MCS/FRAPCON is a cycle-wise Picard-iteration inner-coupling code system. It is based on the Monte Carlo neutron-transport code MCS and employs the steady-state fuel behavior prediction code FRAPCON as a thermal-hydraulic (T/H) and FP solver. MCS is developed by the Computational Reactor Physics and Experiment Lab of Ulsan National Institute of Science and Technology for the high-fidelity full-core analysis of large-scale commercial light water reactors. Results of power, fuel temperature, coolant temperature and coolant density distributions are presented and analyzed for a quarter-core pin-wise depletion simulation of the BEAVRS Cycle 1 benchmark with T/H and FP feedback (10 axial meshes per pin, 180,870 depletion cells in total). For code-code comparison purpose, the depletion simulation is also conducted with the MCS/TH1D (internal one-dimensional T/H solver) and MCS/CTF (external sub-channel 3D T/H solver) coupled code systems. The dependence to the burnup of the power, fuel temperature, coolant temperature, and coolant density distributions is analyzed by comparison between the three coupled systems. Validation is performed against BEAVRS measured data for the calculated boron letdown curve and calculated distributions of axially-integrated assembly-wise detector signal. Finally, unique distributions of parameters that can only be obtained by FP analysis are examined to illustrate the advanced analysis capability of MCS/FRAPCON.