Next Generation Nuclear Research Articles

Construction Strategy for the Future K-DEMO Superconductor Test Facility

The 16 T superconducting magnet for the next generation fusion reactor, called K-DEMO would be a great engineering challenge with current technology. The project on the test facility, named SUCCEX (Super Conducting Conductor Experiment) is prerequisite for the development of K-DEMO superconducting magnet. It would generate an external magnetic field of 16 T for developing and testing K-DEMO superconductors. Also, this project will give a guideline to fabricate future superconducting conductors for the next generation nuclear fusion reactor construction. It will be an essential facility for the production of superconducting magnets for K-DEMO. In 2021, the Korean government has approved this project and the project is in its early stages of construction. The construction strategy and the current progress are discussed in this paper.

Membrane reactors for hydrogen generation: From single stage to integrated systems

The excessive production of carbon dioxide, the greenhouse effect, and the resulting global warming due to the use of fossil fuels have prompted governments, industry, and researchers toward renewable and clean energy sources. Hydrogen has emerged as a novel carbon neutral energy carrier with the highest specific energy content. Production, storage, distribution, and utilization of hydrogen are rapidly increasing. In this context, the use of hydrogen in power plants, fuel cells and internal combustion engines is constantly being developed. However, widespread use of hydrogen requires its sustainable production at low cost and under high purity conditions, which is the subject of further research. Hydrogen separation using membrane reactors is not a new concept; however, it has gained special attention in the last decades and represents a relevant and promising option to potentially overcome the above mentioned problems. This review examines advances in industry and research in hydrogen utilization, hydrogen production using membrane reactors, the nature and physicochemical properties of membranes, and the various configurations of membrane reactors. The effects of membrane reactors on the interplay of temperature, pressure, reactant conversion, hydrogen production, and thermodynamic equilibrium are explained, and the results are compared with those of conventional reactors. Finally, fuel cells, solar collectors, and their integrability into membrane reactors are discussed.

Lattice Boltzmann solidification modeling of forced convection internal flows applied to Gen-IV nuclear reactor coolants

Generation-IV nuclear reactors involve novel coolants such as molten salts and lead. One of the issues that must be assessed with these coolants is the risk of solidification during reactor operation owing to their high melting temperature. However, the numerical simulation of solidification problems is computationally demanding. In this work, we assess the applicability of Lattice Boltzmann Method (LBM) to nuclear coolants solidification problems under laminar internal forced flow conditions. The double distribution Total Enthalpy-Partially Saturated Method is implemented in the open-source code OpenLB and compared against more standard Finite-Volume Computational Fluid Dynamics (FV-CFD) results, which uses the enthalpy-porosity Eulerian Multiphase method, in commercial code Star-CCM+. The test cases comprise the solidification of two fluids with significantly different Prandtl numbers, molten salt and lead, in a 3D pipe geometry in laminar flow conditions with an initial Reynolds number of 100. The solid interface advances from the cooled wall towards the bulk of the pipe reducing the area of flow passage as the transient evolves. Results of LBM velocity fields, solid fraction profiles, temperature fields, and pressure fields are in good agreement with FV-CFD results for both coolants. Furthermore, LBM stability and accuracy limits are independently explored for molten salts and lead using Two-Relaxation Time (TRT) and Bhatnagar-Gross-Krook (BGK) collision operators. Simulation time between FV-CFD and LBM is also contrasted for both coolants. In this work, LBM solvers use explicit time schemes with a regular grid, whereas FV-CFD solutions are able to include implicit time schemes and near-wall mesh refinement. For molten salt flows (high Prandtl number) FV-CFD allows for considerably larger timesteps than LBM, which is limited by the maximum lattice Mach number limitation and stability considerations. For lead flows (low Prandtl number), the implicit scheme in FV-CFD is limited in terms of accuracy for large timesteps due to the limiting Fourier-number. On the other hand, LBM does not present notable stability issues for lead and the lattice Mach number limit is not as relevant as for the molten salt case for this Reynolds number, allowing for faster simulation times than FV-CFD.

Investigation of thermal hydraulic behavior of the High Temperature Test Facility's lower plenum via large eddy simulation

A high-fidelity computational fluid dynamics (CFD) analysis was performed using the Large Eddy Simulation (LES) model for the lower plenum of the High–Temperature Test Facility (HTTF), a ¼ scale test facility of the modular high temperature gas-cooled reactor (MHTGR) managed by Oregon State University. In most next–generation nuclear reactors, thermal stress due to thermal striping is one of the risks to be curiously considered. This is also true for HTGRs, especially since the exhaust helium gas temperature is high. In order to evaluate these risks and performance, organizations in the United States led by the OECD NEA are conducting a thermal hydraulic code benchmark for HTGR, and the test facility used for this benchmark is HTTF. HTTF can perform experiments in both normal and accident situations and provide high-quality experimental data. However, it is difficult to provide sufficient data for benchmarking through experiments, and there is a problem with the reliability of CFD analysis results based on Reynolds–averaged Navier–Stokes to analyze thermal hydraulic behavior without verification. To solve this problem, high-fidelity 3-D CFD analysis was performed using the LES model for HTTF. It was also verified that the LES model can properly simulate this jet mixing phenomenon via a unit cell test that provides experimental information. As a result of CFD analysis, the lower the dependency of the sub-grid scale model, the closer to the actual analysis result. In the case of unit cell test CFD analysis and HTTF CFD analysis, the volume-averaged sub-grid scale model dependency was calculated to be 13.0% and 9.16%, respectively. As a result of HTTF analysis, quantitative data of the fluid inside the HTTF lower plenum was provided in this paper. As a result of qualitative analysis, the temperature was highest at the center of the lower plenum, while the temperature fluctuation was highest near the edge of the lower plenum wall. The power spectral density of temperature was analyzed via fast Fourier transform (FFT) for specific points on the center and side of the lower plenum. FFT results did not reveal specific frequency-dominant temperature fluctuations in the center part. It was confirmed that the temperature power spectral density (PSD) at the top increased from the center to the wake. The vortex was visualized using the well-known scalar Q-criterion, and as a result, the closer to the outlet duct, the greater the influence of the mainstream, so that the inflow jet vortex was dissipated and mixed at the top of the lower plenum. Additionally, FFT analysis was performed on the support structure near the corner of the lower plenum with large temperature fluctuations, and as a result, it was confirmed that the temperature fluctuation of the flow did not have a significant effect near the corner wall. In addition, the vortices generated from the lower plenum to the outlet duct were identified in this paper. It is considered that the quantitative and qualitative results presented in this paper will serve as reference data for the benchmark.

Insight into the thermocapillary radiative flow of hybrid nanoliquid film over an infinite rotating disk with entropy generation and heat source

Research into hybrid nanofluid flow over a disk surface is growing due to numerous processes in marine or gas turbines, rotating disk reactors for biofuels generation, cooling of spinning equipment components and other industrial applications. The purpose of this research study is to examine the entropy generation for the transient thermocapillary flow of hybrid nanoliquid thin films across a disk surface with different shapes of nanoparticles. To analyze the nanomaterial, the Tiwari–Das hybrid nanofluid flow model is established, and in doing so, Prandtl’s boundary layer theory is incorporated into this model. The energy equation considers the impacts of thermal radiation, two different kinds of heat sources — namely an exponentially space-dependent heat source and a linear thermal heat source — and viscous dissipation. The nonlinear partial differential equations, which explain the flow processes, are transformed into the nondimensional ordinary differential equations (ODEs). Once the ODEs have been constructed, they are next solved using the bvp4c technique. In addition, the surface drag force and the rate of heat transfer are both evaluated as functions of the shape factor of the nanoparticles that are disseminated in the base fluid. The influence of significant parameters on the flow fields is depicted graphically, and adequate physical explanations are provided for each representation. One of the important findings of this study indicates that the largest amount of heat transfer can be accomplished at the disk surface by employing nanoparticles with blade shape, while this physical quantity is the least when nanoparticles with spherical shapes are used. The temperature profile grows as the thermal and exponential heat sources increase. Entropy generation improves with the increase in either the magnetic parameter or the Brinkman number.

Nuclear Power Reactor- An Overview

Nuclear energy is the second-highest producer of carbon-free electricity after hydropower. Currently, nuclear energy is generated from fission- where the nuclei of an atom split into several parts; however, research on the energy generation from fusion is ongoing. Nuclear reactors are termed as heart of this process. This paper reviews the major nuclear reactor involved in these reactors' power generation and working and provides a broader aspect of the GEN-IV reactor.

Analisis Kebocoran Boiler Pipe Akibat Korosi pada Recovery Boiler 5 PT. ABC.

Boiler is one component that is widely used in the energy generation industry and reactors. One of the main components of the boiler is the boiler pipe. In field activities, there was a leak in the boiler pipe which experienced corrosion and erosion of the walls of the pipe. This condition results in a decrease in power and capacity in the boiler. The purpose of this practical work is to observe the causes of damage to the boiler pipe, and provide recommendations for prevention due to damage to the boiler pipe. The methods used are: literature study, observation, interviews, and measurement. The results obtained: (1) The calculation results show that the value of the Reynolds number obtained is 16,032,025.86 (Re>4000, turbulent flow). (2) The cause of corrosion on the boiler pipe is due to liquid droplet impingment which is accompanied by an increase in flow velocity and the occurrence of two phases (steam and water) thereby accelerating depletion in the boiler.

Corium pool behaviour in the core catcher of a sodium-cooled fast reactor

Within the framework of the severe accident mitigation strategy for GEN-IV reactors, the behaviour of molten corium in the SFR core catcher is a subject of significant interest. This article aims to define the main considerations within the core catcher during a prospective severe accident scenario, in particular with regard to the rate of heat transfer to the core catcher structures. On the basis of the core catcher design adopted at the end of the ASTRID project, a simplified model is proposed to conservatively evaluate the quasi-steady state behaviour in a stratified oxide-metal molten corium pool with mechanically stable crusts at the interfaces between the stratified layers and at the core catcher walls. The corium oxide layer is subject to significant internal heat generation due to the radioactive decay heat. The proposed model implies that the core catcher should withstand reasonable worst case estimates for the thermal load from the corium pool, avoiding sodium boiling and thermal ablation of the core catcher structures. Thermal ablation is only predicted in the cases of an order of magnitude under-prediction of the heat transfer coefficients at the lateral boundary of the corium oxide layer, or for local heat flux values in excess of 3.7 times the surface average. Further investigations will be dedicated to the revision of the more conservative model assumptions, in particular regarding the magnitude of the heat flux from high aspect ratio encloses with large internal Rayleigh numbers. A refined transient version of the model is intended for implementation within the PROCOR platform developed by the CEA.

Neutronic analysis of the European sodium cooled fast reactor with Monte Carlo code OpenMC

Abstract The sodium-cooled fast reactor is a Generation-IV International Forum recommended technology, with an aim to improve sustainability, safety, and proliferation resistance. To ensure accurate reactor physics calculation and safety analyses, nuclear data libraries require continuous improvement through modifications based on additional measurements, evaluations, and validation studies with criticality experiments. In this work the Sodium-cooled Fast Reactor Uncertainty Analysis in Modeling (SFR-UAM) benchmark served as a basis to assess differences in nuclear data libraries and estimate variability in criticality and power distribution results. The research has been carried out using the OpenMC code and the study presented here covers two SFR models: MOX-3600 and ABR-1000. The neutronic calculation of numerous parameters in fast spectrum systems including effective multiplication factor (k eff), effective delayed neutron fraction (β eff), sodium void reactivity (Δρ Na), Doppler constant (Δρ Doppler), and control rod (ρ CR) worth were calculated and compared mainly to five libraries: ENDF/B-VII.1, ENDF/B-VIII, JEFF-3.3, JENDL-4.0 and TENDL-2019. In addition, sensitivity calculations using GPT-free method were conducted to understand relevant sensitivities for a given quantity of interest in major isotope/reaction pairs. The major driver of observed uncertainty in k eff are found for the high actinide isotopes mainly capture cross section of 239, 240Pu as well as fission reaction of 239Pu.

The Primary Irradiation Damage of Hydrogen-Accumulated Nickel: An Atomistic Study.

Nickel-based alloys have demonstrated significant promise as structural materials for Gen-IV nuclear reactors. However, the understanding of the interaction mechanism between the defects resulting from displacement cascades and solute hydrogen during irradiation remains limited. This study aims to investigate the interaction between irradiation-induced point defects and solute hydrogen on nickel under diverse conditions using molecular dynamics simulations. In particular, the effects of solute hydrogen concentrations, cascade energies, and temperatures are explored. The results show a pronounced correlation between these defects and hydrogen atoms, which form clusters with varying hydrogen concentrations. With increasing the energy of a primary knock-on atom (PKA), the number of surviving self-interstitial atoms (SIAs) also increases. Notably, at low PKA energies, solute hydrogen atoms impede the clustering and formation of SIAs, while at high energies, they promote such clustering. The impact of low simulation temperatures on defects and hydrogen clustering is relatively minor. High temperature has a more obvious effect on the formation of clusters. This atomistic investigation offers valuable insights into the interaction between hydrogen and defects in irradiated environments, thereby informing material design considerations for next-generation nuclear reactors.

Evolution of the microstructure of superfine grain graphites under thermal oxidation

Nuclear graphite is used as a moderator and structural material in a number of Gen-IV reactor designs, such as the (Very) High Temperature Reactors. Chronic or acute thermal oxidation affects the mechanical properties of the graphite components and the impact on the structural integrity of the core depends on the evolution of the microstructure during oxidation.This study qualitatively examined the microstructure of four superfine grain graphites using polarised and fluorescent optical microscopy. These graphites had similar physical properties but notable differences in the oxidation rate at 700 °C. Using fluorescent microscopy, it was possible to distinguish between open and closed porosity in graphite, which is useful in understanding the progression of thermal oxidation and penetration depth. Based on the samples in this work, fine open and closed porosity may be linked to increased surface oxidation and relatively high oxidation rate, whereas high penetration depth may be linked to relatively low open porosity consisting mainly of large pores. Fairly large domains of coherent crystallite orientation, perhaps due to the presence of agglomerates, may also indicate lower oxidation rates at 700 °C.This paper demonstrated that fluorescent microscopy combined with image analysis, is a useful tool in understanding the progression of thermal oxidation and the corresponding effect on the microstructure.

The effects of high-temperature ion-irradiation on early-stage grain boundaries serrations formation in Ni-based alloys

Nickel-based superalloys display outstanding properties such as excellent creep strength, remarkable fracture toughness parameters, and corrosion resistance. For this reason, Ni-based materials are considered as materials dedicated to the IV generation of nuclear reactors. Although these materials seem promising candidates, their radiation resistance and impact of radiation damage on the deformation mechanism are still not fully understood. In this work, two commercially available nickel-based alloys (Hastelloy X and Haynes 230) were investigated. Structural and mechanical properties have been described by means of SEM/EBSD, TEM, and nanoindentation tests. Radiation damage has been performed by Ar-ion with energy 320 keV with two doses up to 12dpa. Obtained results have revealed a hardening effect for both levels of damage. However, more intensive effects were observed for Hastelloy X. Moreover, a significant change in precipitates' morphology in Hastelloy X has been observed. It has been proposed that structural differences between both alloys determine the type of occurring radiation-induced processes. Excess energy deposited into materials' structure during ion-irradiation can lower the temperature of nucleation of high-temperature phases, which initiates the formation of grain boundary serrations.

Flow field characteristics of a 127-pin rod bundle with hexagonal spacer grids

A determination of nominal flow phenomena in liquid metal fast reactor (LMFR) fuel assemblies is critical toward generation-IV reactor development. Axially positioned spacer grids are used to maintain the geometry of hexagonal rod bundles and simultaneously introduce perturbations in the flow. Three-dimensional (3D) printed asymmetric honeycomb spacer grids were installed in a prototypical 127-pin LMFR fuel assembly model to study complex fluid dynamics interactions induced by the spacer grid and rods. To characterize flow dynamics in this intricate geometry, time-resolved particle image velocimetry (TR-PIV) using the matched-index-of-refraction method was employed to obtain non-intrusive velocity measurements for three axial planes (one near-wall and two interior planes) at a Reynolds number of 6000. The statistical TR-PIV results compared sub-channel-dependent normalized time-averaged velocity, velocity fluctuations, Reynolds stress, vorticity, and turbulence kinetic energy distributions. TR-PIV line profiles characterized downstream spacer grid flow dynamics. Two-point spatial and spatial–temporal cross-correlation fields revealed local coherent structures and quantified convection velocities of traveling vortices. Spatial–temporal decomposition using dynamic mode decomposition (DMD) applied to the near-wall vorticity fields extracted turbulent structures and flow instabilities in the wake region of the spacer grid, along with their decay and frequency rates. Reduced-order velocity fields from DMD reconstructions identified the most energy-containing coherent structures persistent in the near-wall region. This research provides experimental data sets and analyses of flow behavior in rod bundles with hexagonal spacer grids. The results are critical toward LMFR design and geometry optimization, crucial for the validation of computational fluid dynamics and reduced-order flow models.

Miniature DC electromagnetic pumps of molten lead and sodium to support development of Gen-IV nuclear reactors

This work developed designs of submerged, dual-regions, miniature DC-EM pumps for circulating molten lead and liquid sodium at ≤500 °C without active cooling for use in ex-pile and in-pile test loops to support of materials and fuel developments for Gen-IV sodium and molten lead fast nuclear reactor. These pumps with two pumping regions are 57 mm, 66.8 mm, 95.4 mm, and 133.5 mm in diameter and have two Alnico 5 permanent magnets with Hiperco-50 pole pieces for focusing the magnetic field lines in the flow duct. The Equivalent Circuit Model (ECM) linked to the Finite Element Method Magnetics (FEMM) software in MATLAB platform calculated the pumps characteristics for a wide range of parameters. Also determined are the pumps dimensions for the highest cumulative pumping power and peak efficiency. These dimensions are the height, width, and length of the flow duct, the thickness of the Alnico magnets, the length of current electrodes, and the separation distance between the two pumping regions. For molten lead, the pumping power increases from 368 W to 728 W, and the peak efficiency from 14.7% to 31.6% with increased pump diameter from 57 mm to 133.5 mm. For liquid sodium, both the pumping power and peak efficiency are higher, increasing from 392 W to 767 W and from 44.3 to 51.2%, respectively, with increased pump diameter from 57 mm to 133.5 mm.

A surrogate model for the helium production rate in fast reactor MOX fuels

Helium production in the nuclear fuel matrix during irradiation plays a critical role in the design and performance of Gen-IV reactor fuel, as it represents a life-limiting factor for the operation of fuel pins. In this work, a surrogate model for the helium production rate in fast reactor MOX fuels is developed, targeting its inclusion in engineering tools such as fuel performance codes. This surrogate model is based on synthetic datasets obtained via the SCIANTIX burnup module. Such datasets are generated using Latin hypercube sampling to cover the range of input parameters (e.g., fuel initial composition, fission rate density, and irradiation time) and exploiting the low computation requirement of the burnup module itself. The surrogate model is verified against the SCIANTIX burnup module results for helium production with satisfactory performance.

Recent Progress on Creep Properties of ODS FeCrAl Alloys for Advanced Reactors.

In order to meet the growing energy demand, more environmentally friendly and efficient GEN-IV reactors have emerged. Additionally, nuclear structural materials need larger more safety margins for accident scenarios as a result of the Fukushima accident. In order to extend the failure time and lessen the effect of accidents, this design method for accident-tolerant fuel materials calls for cladding materials to maintain good corrosion resistance and mechanical properties during a beyond design basis accident (BDBA). Accidents involving nuclear reactors would undoubtedly result in higher temperatures, which would make it much harder for materials to withstand corrosion. Oxide dispersion strengthened (ODS) FeCrAl alloys have shown promise as candidate materials because of their extraordinarily slow reaction rates under high-temperature steam. However, the addition of the Al element renders the alloy's high-temperature mechanical properties insufficient. In particular, studies on the alloy's creep properties are extremely rare, despite the fact that the creep properties are crucial in the real service environment. Therefore, this paper focuses on the creep properties of ODS FeCrAl alloy, summarizes and analyzes the research results of this material, and provides a reference for future research and applications.

Synthesis and characterization of 9Cr ODS F/M steel with optimized Y2O3 and Ti content and its comparison with P91 steel

Oxide dispersion strengthened (ODS) steels are considered as potential candidates for future Gen-IV reactors core structural applications. Ti is added to this alloy to refine the dispersoid and facilitates formation of Y-Ti-O complex oxides. Optimisation of Y2O3 and Ti content in these ODS steels governs size and stoichiometry of the dispersoids. Equal amount of both Ti and Y2O3 content resulted in formation of finer and stable Y-Ti-O dispersoids. In this study two varieties of ODS alloys are synthesized: one with 0.35 wt% Ti and Y2O3 while another with 0.2 wt% Ti and Y2O3 by mechanical milling followed by SPS. The microstructure and microtexture of these two varieties of ODS steels were compared with modified 9Cr1Mo (P91) steel. The creep-rupture studies reveal that the ODS steels have better creep-rupture life as compared to P91 steel, and the steel with 0.35 wt% Ti and Y2O3, possess superior mechanical properties.

Small Modular Reactors Licensing Process Based on BEPU Approach: Status and Perspective

The competitiveness of small modular reactors (SMRs) has been planned based on design simplification, short construction time, passive safety systems, and enabling self-financing by ramp-up construction. Due to the global energy challenges, SMRs have received pervasive attention from a wide range of researchers, designers, developers, stakeholders, and customers. Besides the many advantages related to the design of SMRs, there are challenges ahead of these reactors. SMR licensing is one of the most critical challenges in the front deployment of these reactors. This challenge stems from innovations in SMR designs and systems, such as modularity or deployment for desalination, energy storage, hydrogen production, process heat, and district heating. Due to the lack of experimental data and technical knowledge, the licensing challenges for non-water coolant SMRs are more complicated. Nearly all previous generation reactor licenses were based on conservative analysis while the decision-making methods based on best-estimate and realistic approaches have received more attention in recent years. Thus, the method known as the best estimate plus uncertainty (BEPU) approach is selected for licensing in some cases. At this time, using the BEPU approach in licensing for conventional NPPs is a mature technology and ready for industrial application. Nevertheless, because most previous reactors were licensed based on conservative methods, developers and even regulatory bodies resist re-assessments based on the BEPU approach, while using the choice of conventional conservative methods is a type of roll-back for next-generation SMRs. Thus, this work reviews the BEPU approach and clarifies the possibility of using this approach in the licensing process of SMRs. The lack of experimental data and tight coupling of phenomena along with uncertainty quantification are the main challenges ahead of using BEPU in the licensing process of SMRs.

Design and development of low density, high strength ZrNbAlVTi high entropy alloy for high temperature applications

Superior high temperature strength, better thermodynamic stability and good workability are some of the desirable attributes of structural materials for next generation nuclear reactors. In this regard, the present study involves the development of a low density, equiatomic ZrNbVTiAl High Entropy Alloy (HEA) for nuclear applications. The alloy was designed using a combination of thermodynamic simulations and first principle techniques and later synthesized using levitation melting to ensure better purity and compositional homogeneity. The melted alloy possessed lower density (5.76 g/cm3) and higher hardness (5.7 GPa) in comparison to conventional structural materials like austenitic steel. Detailed characterization of the alloy showed the presence of bcc type phase along with small fractions of Al- Zr type intermetallics in as-solidified condition. The mechanical properties of the equiatomic ZrNbVTiAl alloy were studied using compression testing at room temperature and up to 1000 °C. The alloy exhibited superior specific yield strength as well as dynamic recrystallization at high temperatures resulting in the breaking of cast structure, thereby reflecting its appreciable workability. The alloy also showed a satisfactory elastic modulus (128 GPa) when measured using ultrasonic technique. The experimentally measured values of elastic moduli of alloy were compared with the values obtained through Density Functional Theory (DFT) simulations and the results matched within acceptable range. Results from the present study indicates that the equiatomic ZrNbVTiAl alloy could be a suitable candidate for light weight, high temperature and high strength structural applications.

Scale-up single chamber of microbial fuel cell using agitator and sponge biocarriers

ABSTRACT Despite the high efficiency of microbial fuel cells (MFCs), MFCs cannot be a suitable alternative for treatment plants because of insufficient power generation and tiny reactors. Additionally, the increased reactor size and MFC stack result in a reduction in production power and reverse voltage. In this study, a larger MFC with a volume of 1.5 L has been designed called LMFC. A conventional MFC, called SMFC, with a volume of 0.157 L, was constructed and compared with LMFC. Moreover, the designed LMFC can be integrated with other treatment systems and generate significant electricity. In order to evaluate MFC's ability to integrate with other treatment systems, the LMFC reactor was converted into MFC-MBBR by adding sponge biocarriers. A 9.5 percent increase in reactor volume resulted in a 60 percent increase in power density from 290 (SMFC) to 530 (LMFC). An agitator effect was also investigated for better mixing and circulating substrate, which positively affected the power density by about 18%. Compared with LMFCs, the reactor with biocarriers generated a 28% higher power density. The COD removal efficiency of SMFC, LMFC, and MFC-MBBR reactors after 24 h was 85, 66, and 83%, respectively. After 80 h of operation, the Coulombic efficiency of the SMFC, LMFC, and MFC-MBBR reactors was 20.9, 45.43, and 47.28%, respectively. The doubling of coulombic efficiency from SMFC to LMFC reactor shows the design's success. The reduction of COD removal efficiency in LMFC is the reason for integrating this reactor with other systems, which was compensated by adding biocarriers.