Next Generation Nuclear Research Articles

Modelling and assessment of hydrogen combined cycle power plant using aluminum-water reaction as renewable fuel

Owing to contamination of air and environmental hazards, the demand for sustainable fuels in power plants and other heat engines has climbed undeniably. Aluminum metal is classified among high energy density and low carbon content, which can be reacted with water and repeatedly released a large amount of heat and hydrogen. Since hydrogen is one of the zero-carbon and clean fuels, aluminum can be exploited as a promising energy-carrying and renewable source of energy. This paper is dedicated to the customization of two novel schemes of the combined cycle of power generation systems. The main components of the proposed integrated installations include a hydrogen turbine, gas turbine, multi pressure steam turbines, condenser, heat recovery steam generator and aluminum-water reactor, in which pressurized water absorbs produced heat of the reactor then superheated steam expands in steam turbines. Direct supplied hydrogen, which is generated from the Al-water reaction, expands in the hydrogen turbine and subsequently burns in the combustion chamber of the gas turbine. Hydrogen properties, consumed heat of reactor, generated powers, effects of streams pressure, energetic and exergetic productivity variation of both schemes were evaluated comparatively to ensure the best system performance along with the high net efficiency. According to the numerical analysis of the thermodynamic characteristics in steady state regime using EES software, the energy and energy efficiencies of scheme A were calculated to be 40.72% and 38.36%, besides in scheme B were 39.6% and 37.3%, respectively.

Study of reducing losses, short-circuit currents and harmonics by allocation of distributed generation, capacitor banks and fault current limiters in distribution grids

The growing interest for researches that focus on renewable energies, as well as the increase in funding for this sector, is thanks to the advantages of renewable energy namely the reduction of environmental impacts and their great reliability. However, to maintain the correct operation of the electrical grid and its quality, it is essential to study the impacts that these generating sources have on the electrical system. The present research paper presents a novel methodology that considers the optimal allocation of photovoltaic distributed generation, capacitor bank, and fault current limiters reactors in distribution feeders, whose purpose is minimizing simultaneous the short-circuit current, active losses and harmonic distortion rate of distribution feeders. For this, a Multi-objective Gray Wolf Optimize has been applied and tested in three different distribution feeders. The result demonstrated that during the entire optimization process, the greatest reduction of the analyzed parameters, which are losses, harmonic distortion rate and short circuit current are associated with the location and number of devices that are allocated in the system.

Precipitates change of P92 steel under 3.5 MeV Fe13+ ion irradiation at 400 °C

9–12% Cr ferritic/martensitic steels are promising candidate materials for applications in fourth generation nuclear reactors. 9%Cr ferritic/martensitic steel P92 was normalized at 1050 °C for 0.5 h and subsequently tempered at 765 °C for 1 h. The precipitates of the normalized-and-tempered steel after 3.5 MeV Fe13+ ion irradiation at 400 °C to 0.66 dpa were studied using transmission electron microscopes. After irradiation, the number of precipitates increased, especially for the number of large-sized precipitates, and there was an increase in the overall size of precipitates. Pre-existing Cr-rich M23C6, Nb-rich MX and V-rich MX precipitates in the normalized-and-tempered steel still existed in the irradiated steel, and the M23C6 precipitates remained almost unchanged in their composition. Irradiation-induced Nb-rich M6C5 carbide with a base-centered monoclinic structure, V-rich M2X (V2C type) carbonitride with a simple orthorhombic crystal structure, and Fe71Cr26(Mo,W)3 (σ-FeCr type) intermetallic compound with a simple teragonal crystal structure were identified in the irradiated steel. The formation of these irradiation-induced phases is also discussed.

Comprehensive study on catalytic coating tubular reactor with electromagnetic induction heating for hydrogen production through methanol steam reforming

This study presents an induction heating-based reactor for on-board methanol steam reforming for hydrogen production reactors and distributed hydrogen refueling stations, with the goal of developing a hydrogen production process with zero carbon emissions. The numerical results showed that the enhanced hydrogen production performance and temperature cold spots could be eliminated, and the methanol conversion was close to 100%. at 150 A and 20 kHz. Induction heating may significantly shorten the reactor's cold start-up time. Under 50 kHz, the temperature of the reactor wall can reach more than 500 K in 50 s, in other words, induction heating overcomes the problems of slowly heating/cooling rate for on-board methanol steam reforming for hydrogen generation reactors. Overall, this study supports the feasibility of producing green hydrogen with zero carbon emissions for advancing the development of an environmental hydrogen economy and the cold start-up of distributed hydrogen energy and vehicle-used hydrogen production equipment.

Construction Materials of Active Zones of New Generation Nuclear Reactors

The materials of the article consider the analysis of construction materials of active zones of new generation nuclear reactors. The analysis reflects general ideas about the development of reactor technologies: in the 1950s and 1960s, the first generation of reactors was created; in the early 1970s, the operation of industrial reactors began - reactors of the second generation: pressurized water reactors (WWER, PWR), boiling water reactors (RBMK, BWR), heavy water reactors (CANDU), as well as gas-cooled reactors (AGR). Further development of some types of reactors made it possible to create reactors of the third generation in the 1980s. Priority when choosing directions of development in the category of revolutionary projects should have proposals capable of bringing a new quality to solving the problems of the nuclear energy industry of the future. Promising reactors have advantages in economy, safety, reliability and non-proliferation of nuclear materials. The effectiveness and reliability of structural materials are determined by the totality of changes in the characteristics of the materials as a result of the entire complex of phenomena occurring in them in the field of irradiation, in connection with the changing parameters and operating conditions. The use of high-purity metals as initial components of new structural materials and the development or optimization of their smelting technologies should ensure the required level of characteristics and properties of products made from them. The implementation of these concepts should be ensured by the development of new structural materials: ferritic-martensitic and austenitic steels, nickel and other new alloys.

Laser Cladding of Alloy 82 Powder for Corrosion Protection in Small Modular Reactors

There has been an increased demand for surface modification technologies to enhance corrosionresistance and wear-resistance. Among the representative surface coating technologies, laser cladding has attracted great attention because of its multiple advantages including low heat input, low dilution ratio, controllable clad height, etc. Recently, laser cladding has been considered as a surface coating technology for the next-generation small modular reactor, although submerged arc welding was utilized for the 3rd generation nuclear reactor. Cobalt-free materials are required as cladding materials in the nuclear reactor because cobalt has a long half-life. In this work, nickel based cobalt-free Alloy 82 powder was utilized. The main experimental parameters for laser cladding were intensively investigated by varying laser power, scan speed, powder supply, carrier and shield gas, overlap ratio, etc. Additionally, cross-sectional area calculation and EPMA analysis were carried out to examine the dilution ratio. Mechanical properties were also evaluated using microhardness tests and wear tests at high temperatures. Finally, corrosion tests were performed to compare a laser clad surface with an uncoated carbon steel surface. Our study indicates that the laser cladding using Alloy 82 powder is feasible for corrosion protection in next-generation small modular reactors.

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.