Advanced Nuclear Systems Research Articles

The development of nuclear reactor three-dimensional neutronic thermal–hydraulic coupling code: CorTAF-2.0

Numerical reactor technology is the development direction of advanced nuclear energy system simulation, which has the ability of high-fidelity multi-physical field coupling calculation. The thermal–hydraulic and neutron physics fields in reactor core exhibit strong feedback relationships. Establishing the neutronic thermal–hydraulic coupling calculation model of the whole core “pin-by-pin” is the key technology for the development of numerical reactors. In this study, the neutron diffusion analysis model and cross-section database have been developed based on the open source CFD platform OpenFOAM. The three-dimensional neutron diffusion analysis model based on the finite volume method has been validated by benchmark problems. By integrating the neutron diffusion analysis model and cross-section database into nuclear reactor core three-dimensional thermal–hydraulic characteristics analysis code CorTAF, the strongly coupled neutronic thermal–hydraulic coupling analysis code CorTAF-2.0 has been developed. The neutronic thermal–hydraulic coupling phenomena in the full core of AP1000 have been simulated using CorTAF-2.0, yielding the sub-channel level parameters distribution of the whole core. This work offers valuable references for reactor full-core neutronic thermal–hydraulic coupling analysis.

Advancement Toward the Experimental Measurement of the Lead–Bismuth Eutectic Thermal Properties Using Differential Scanning Calorimetry Technique

Abstract The differential scanning calorimetry (DSC) method has recently emerged as a sophisticated and precise technique for promising contributions to the thermal analysis of various materials, including heavy liquid metal (HLM) coolants. However, there is a lack of experimental studies on the thermal properties of lead-based fluids, such as lead–bismuth eutectic (LBE) and lead–lithium eutectic, which are potential candidates for use as coolants, breeders, and neutron multipliers in advanced nuclear systems like the fourth-generation lead-cooled fast reactor. The available experimental data on the thermal properties of LBE and other lead-based fluids is limited, and the measurements have significant uncertainty. In addition, the composition of components used in the previous studies is inconsistent, and the environmental conditions were often unknown. Therefore, to fill these gaps and advance the thermal properties measurement technique for heavy liquid metal coolants, ENEA Brasimone, in collaboration with DICI-UNIPI, has installed a DSC instrument setup. The experiments performed at the installed DSC setup are focused on measuring some essential thermal properties of LBE using DSC. The experience gained from this work will facilitate the measurement of other fluids based on lead alloy, especially lead–lithium eutectic, a potential candidate for breeder, coolant, and neutron multiplier in demonstration power plant fusion reactors. This study represents the first effort to advance the DSC approach for accurately measuring the thermal characteristics of heavy liquid metals that are highly reactive, such as lead–lithium, which has significant potential in advanced nuclear systems.

Transport vibration analysis of mobile liquid Pb-Bi micro-reactors with non-uniform mass distribution

The mobile micro-reactors, as advanced nuclear systems, had the capability to provide energy to remote areas and islands. However, the non-uniform mass distribution characteristic of Liquid Lead-Bismuth Eutectic (LBE), which possessed a high density, substantial mass, and compact volume, mobile micro-reactors required special attention during transportation when mounted on trucks and passing through rough roads. This paper aimed to investigate the vibration responses of the mobile micro-reactor with non-uniform mass distribution during transportation. To achieve this, a separated model was developed using the Lagrange multi-body dynamic equations, the recorded responses data were evaluated using MATLAB, a numerical simulation software. Furthermore, a parametric study was conducted to examine the influence of the mass of LBE mass and the positions of the micro-reactor on the vibration responses during transportation. The results indicated that the mass of the micro-reactorwas directlyproportionaltothe vibration responses, and the responses could be different with the moveable situation of the reactor. This study demonstrated that the separated model could evaluate the transportation vibration responses of both the mobile micro-reactors and the truck more accurately, thereby facilitating the safe transportation of mobile micro-reactors with non-uniform mass distribution.

Effects of Alloying Elements and Mechanical Alloying on Characteristics of WVTaTiCr Refractory High-Entropy Alloys.

Refractory high-entropy alloys (RHEAs) are among the promising candidates for the design of structural materials in advanced nuclear energy systems. The effects of Cr, V, Ta, and Ti elements and ball milling on the microstructural evolution and mechanical properties of model RHEAs were investigated. The results show that W-rich BCC1 and Ta-rich BCC2 solid solution phases were generated after a long milling duration. After high-temperature sintering, the (Cr, Ta)-rich phase associated with the Laves phase was observed in the Cr-containing model RHEAs. In addition, a high level of Ti, Ta, and V contents promoted the in situ formation of oxide particles in the alloys. Complex TiTa2O7 and Ta2VO6 oxide phases were identified by TEM, which suggests a solid-state reaction of Ti-O, Ta-O, and V-O subjected to high-energy ball milling. The oxide particles are uniformly dispersed in the BCC matrix, which can result in dispersion strengthening and the enhancement of mechanical properties.

Atomistic simulation of brittle-to-ductile transition in silicon carbide embedded with nano-sized helium bubbles

The tensile response of cubic silicon carbide (SiC) bulk containing cavities (voids and He bubbles) has been investigated using molecular dynamic simulations. The formation of cavities in SiC leads to a significant degradation in the mechanical properties of SiC with more influence on material fracture than initial elastic deformation. The brittle-to-ductile transition occurs in cavity-embedded SiC as the pressure in He bubbles increases. This is associated with the deformation mechanism that bond breaking at a low He bubble pressure transfers to extensive dislocation activities at a higher He bubble pressure. The cavities can effectively concentrate stress around them in the direction perpendicular to the tension, which leads to preferred cracking in the region with a higher tensile stress. The failure mechanism as revealed by this study improves understanding of property degradation in SiC that may be useful for applications of SiC in advanced nuclear energy systems.

In-reactor behaviour and economic assessment of enriched gadolinia burnable absorbers

The increasing demand for energy and rising energy prices have increased the importance of the efficiency and economic viability of operating nuclear power plants. One approach to improving the reactivity control in such cases is the use of burnable absorbers such as gadolinium oxide (Gd2O3). While Gd2O3 has proven effective in regulating reactivity in light water reactors, it has certain drawbacks, including the displacement of fissile isotopes from the fuel composition and residual reactivity penalties caused by isotopes other than the main neutron absorbers. In this study, the potential impacts of 157Gd-enriched Gd2O3 on the operation, safety, and economy of nuclear power plants were assessed. Fuel cycle analyses were conducted using the advanced nuclear design code system Studsvik CASMO4/SIMULATE3. Several aspects were analysed, including peaking factors, reactivity feedback parameters, power distribution, and shutdown margin. The results of the analysis revealed that by removing the isotopes responsible for residual reactivity from Gd2O3, it was possible to achieve the same fuel cycle length with lower uranium enrichment levels, enabling a higher concentration of fissile isotopes and yielding economic benefits without compromising safety.

Cladding Failure Modelling for Lead-Based Fast Reactors: A Review and Prospects

Lead-cooled fast reactors (LFRs) are considered one of the most promising technologies to meet the requirements introduced for advanced nuclear systems. LFRs have higher neutron doses, higher temperatures, higher burnup and an extremely corrosive environment. The failure studies of claddings play a vital role in improving the safety criteria of nuclear reactors and promoting research on advanced nuclear materials. This paper presented a comprehensive review of the extreme environment in LFRs based on the fuel performance analyses and transient analyses of reference LFRs. It provided a clear image of cladding failure, focusing on the underlying mechanisms, such as creep, rupture, fatigue, swelling, corrosion, etc., which are resulted from the motions of defects, the development of microcracks and accumulation of fission products to some extent. Some fundamental parameters and behavior models of Ferritic/Martensitic (F/M) steels and Austenitic stainless (AuS) steels were summarized in this paper. A guideline for cladding failure modelling was also provided to bridge the gap between fundamental material research and realistic demands for the application of LFRs.

Effects of cold rolling and heat treatment on microstructure and mechanical properties of 15Cr-15Ni ODS austenitic steel

Oxide dispersion strengthened (ODS) austenitic steels are considered as the promising candidate cladding materials application for advanced nuclear energy systems. However, the microstructure of ODS austenitic steels fabricated by mechanical alloying is usually not very homogeneous. Thermal mechanical treatment is an effective method for controlling microstructure of alloys. In this work, the effect of cold deformation and subsequent heat treatments on the improvement of microstructure homogeneous and mechanical property of a 15Cr-15Ni ODS austenitic steel was systemically investigated by means of electron backscatter diffraction, transmission electron microscopy and tensile test. Results show that the average grain size decreased and the distribution of grains was more homogeneous after optimized thermal mechanical treatment. The deformation of cold rolling and heat treatment temperatures shows obvious influence on the average size of dispersion particle, the proportion of high angle grain boundary, Schmidt factor and tensile property of 15Cr-15Ni ODS austenitic steel.

Effects of solutes on thermodynamic properties of (TMZrU)C (TM = Ta, Y) medium-entropy carbides: a first-principles study

High entropy carbide ceramics have garnered significant interest as a novel class of ultra-high temperature and superhard metallic materials. In the present work, a comparative investigation was conducted for the first time on the stability, mechanical, and thermodynamic properties of two medium entropy carbides (MECs), (TaZrU)C and (YZrU)C, using high-throughput first-principles calculations. Additionally, data from groups IV and V transition metal monocarbides were employed for comparison. The temperature-dependent thermodynamic properties, including bulk modulus (B), constant volume/constant pressure heat capacity (Cv/Cp), Gibbs free energy, volume, entropy, and thermal conductivity, were evaluated using the Debye-Gruneisen model. The results demonstrate that (TaZrU)C and (YZrU)C exhibit similar trends in their thermodynamic properties, with (YZrU)C displaying slightly superior performance as the temperature rises. This work provides valuable insights into the design of innovative high entropy fuels, holding significant implications for the advancement of MEC ceramic fuels in advanced nuclear power systems and nuclear thermal propulsion systems.

Assessment of flow and heat transfer of triply periodic minimal surface based heat exchangers

An efficient intermediate heat exchanger is crucial for ensuring the safe operation, lifespan and energy efficiency of the advanced nuclear systems. Benefiting from the development of additive manufacturing technology, it is possible to design and optimize heat exchangers by introducing triply periodic minimal surface (TPMS) structures. With the excellent mechanical and thermal performance of TPMS, further improvements in the energy efficiency of advanced nuclear systems are achievable. In this study, I-WP surface, Primitive surface based heat exchangers and a printed circuit heat exchanger for accelerator driven subcritical systems are constructed in three dimensions. In order to obtain the potential enhancement of thermal performance of TPMS-based heat exchangers, the fluid flow and conjugate heat transfer characteristics in TPMS heat exchangers, especially for the specific working medium of lead-bismuth eutectic (LBE) are investigated. A parametric analysis is conducted on the key design variables including the solid volume fraction and hydraulic diameter. The results indicate that TPMS-based heat exchangers could achieve about 2–3 times the total heat transfer rate with half the volume of a printed circuit heat exchanger. The TPMS topologies contribute greater heat transfer enhancement on the Helium side than the LBE side, which helps to optimize the thermal resistance allocation in Helium-LBE heat exchangers. An increase of solid volume fraction enhances the convective heat transfer of Helium in I-WP heat exchanger, while homogenizing the local thermal performance. This study provides a database for the design and optimization of TPMS-based heat exchangers, which contributes to the sustainable future target.

He-ion Irradiation Effects on the Microstructures and Mechanical Properties of the Ti-Zr-Hf-V-Ta Low-Activation High-Entropy Alloys.

High-entropy alloys (HEAs) have shown promising potential applications in advanced reactors due to the outstanding mechanical properties and irradiation tolerance at elevated temperatures. In this work, the novel low-activation Ti2ZrHfxV0.5Ta0.2 HEAs were designed and prepared to explore high-performance HEAs under irradiation. The microstructures and mechanical properties of the Ti2ZrHfxV0.5Ta0.2 HEAs before and after irradiation were investigated. The results showed that the unirradiated Ti2ZrHfxV0.5Ta0.2 HEAs displayed a single-phase BCC structure. The yield strength of the Ti2ZrHfxV0.5Ta0.2 HEAs increased gradually with the increase of Hf content without decreasing the plasticity at room and elevated temperatures. After irradiation, no obvious radiation-induced segregations or precipitations were found in the transmission electron microscope results of the representative Ti2ZrHfV0.5Ta0.2 HEA. The size and number density of the He bubbles in the Ti2ZrHfV0.5Ta0.2 HEA increased with the improvement of fluence at 1023 K. At the fluences of 1 × 1016 and 3 × 1016 ions/cm2, the irradiation hardening fractions of the Ti2ZrHfV0.5Ta0.2 HEA were 17.7% and 34.1%, respectively, which were lower than those of most reported conventional low-activation materials at similar He ion irradiation fluences. The Ti2ZrHfV0.5Ta0.2 HEA showed good comprehensive mechanical properties, structural stability, and irradiation hardening resistance at elevated temperatures, making it a promising structural material candidate for advanced nuclear energy systems.

Solving the issue of reliability data for FOAK equipment in an innovative nuclear energy system

There is a demand for the improved safety and economics of advanced nuclear energy systems. The world's first high-temperature gas-cooled reactor pebble-bed module nuclear power plant (HTR-PM Demo Project) is about to be completed in China. This power plant achieved initial full power at the end of 2022. The success of the HTR-PM Demo Project shows the importance of risk-informed initiatives throughout the project with solid support from comprehensive risk assessment. This paper focuses on a technical issue that needs to be solved when conducting risk-informed initiatives for HTR-PM, namely the issue of reliability data for first-of-a-kind (FOAK) equipment. In short, no data for the FOAK equipment are available in existing reliability databases. On the basis of the HTR-PM pilot experience, this paper proposes a process of estimating the required data for the reliability parameters of FOAK equipment. The process has five steps: (1) identification of key parts and assemblies, (2) preliminary determination of the reliability parameters and the associated reference values for the key parts and assemblies, (3) analysis of maintenance management tactics, (4) estimation of reliability parameters for the FOAK equipment using a Bayesian network, and (5) Bayesian updating with operational data. The helium circulator of HTR-PM is taken as an example to demonstrate the feasibility of the proposed process. A lack of reliability data is a common issue for innovative nuclear energy systems. The proposed process is supposed to be universal to all types of innovative design.

In-situ observation of damage evolution in Mo-SiCf/SiC heterogeneous composite

Innovative metal-SiCf/SiC heterogeneous structure composites are promising accident tolerance fuel cladding materials for the Gen IV advanced reactors. The mechanical damage of these composites is a critical issue for their safe application. In this work, the damage process and mechanism of the Mo-SiCf/SiC cladding are revealed by combining In-situ C-ring test and Finite Element Analysis. The results show that the mirco-cracks originate from the pores in the outermost side of the SiCf/SiC layer, and propagate along the fiber bundles boundaries. At the same time, the “bamboo joint”-like damage occurs in the SiCf/SiC layer during C-ring test due to the periodic pore distribution. The fact that the formation of the main crack in SiCf/SiC layer prior to the fracture of Mo layer indicate the Mo-SiCf/SiC cladding can maintain the hermeticity and structure integrity until the sample fails completely. Moreover, the C-ring strength of the Mo-SiCf/SiC cladding is 360 MPa, which is much higher than 201 MPa of SiCf/SiC cladding. Overall, the Mo-SiCf/SiC cladding structure exhibits better crack growth resistance and higher strength compared to SiCf/SiC cladding. Meanwhile this work will provide technical support and theoretical basis for the structural design of cladding materials for advanced nuclear energy systems.

Effects of grain boundaries and temperature on spinodal decomposition in a binary Fe-Cr alloy: A phase-field simulation

Due to its excellent oxidation resistance, corrosion resistance and mechanical properties, Fe-Cr based stainless steel is a promising candidate for advanced nuclear power systems. At the temperature of 300–600 °C, spinodal decomposition usually occurs in Fe-Cr alloys. However, irradiation can greatly accelerate this process. Fe-Cr based stainless steel decompose into a Cr-enriched α′ phase and a Fe-enriched α phase during spinodal decomposition, and the Cr-enriched α′ phase plays an important role in the hardening/embrittlement of Fe-Cr based stainless steels. In the present work, the evolution of vacancy and the Cr-enriched α′ phase in Fe-Cr alloy is investigated using phase-field method. The simulation results show that the vacancy concentration has a great influence during the spinodal decomposition. The Cr-enriched α′ phase occurs initially at the positions with high vacancy concentration. During spinodal decomposition, the vacancy forms a defect concentration loop around the Cr-enriched α′ phase. In addition, with the increase of temperature, the concentration of Cr atoms in the center of the Cr-enriched α′ phase decreases, and the coarsening rate increases, while the number of the Cr-enriched α′ phase formed in the alloy decreases. At low temperature, α′ phase is mainly coarsened by interlinking; while at high temperature, α′ phase is mainly coarsened by Oswald maturation. Moreover, the Cr-enriched α′ phase occurs at the grain boundary (GB) and the Fe-enriched α phase occurs at both sides of the GB during the spinodal decomposition. At the same time, in polycrystalline system, GB movement during grain growth is inhibited during the spinodal decomposition.

Comparative Analysis of Standard and Advanced USL Methodologies for Nuclear Criticality Safety

The American National Standards Institute/American Nuclear Society national standards 8.1 and 8.24 provide guidance on the requirements and recommendations for establishing confidence in the results of the computerized models used to support operation with fissionable materials. By design, the guidance is not prescriptive, leaving freedom to the analysts to determine how the various sources of uncertainties are to be statistically aggregated. Due to the involved use of statistics entangled with heuristic recipes, the resulting safety margins are often difficult to interpret. Also, these technical margins are augmented by additional administrative margins, which are required to ensure compliance with safety standards or regulations, eliminating the incentive to understand their differences. With the new resurgent wave of advanced nuclear systems, e.g., advanced reactors, fuel cycles, and fuel concepts, focused on economizing operation, there is a strong need to develop a clear understanding of the uncertainties and their consolidation methods to reduce them in manners that can be scientifically defended. In response, the current studies compare the analyses behind four notable methodologies for upper subcriticality limit estimation that have been documented in the nuclear criticality safety literature: the parametric, nonparametric, Whisper, and TSURFER methodologies. Specifically, the work offers a deep dive into the various assumptions of the noted methodologies, their adequacies, and their limitations to provide guidance on developing confidence for the emergent nuclear systems that are expected to be challenged by the scarcity of experimental data. To limit the scope, the current work focuses on the application of these methodologies to criticality safety experiments, where the goal is to calculate a bias, a bias uncertainty, and a tolerance limit for keff in support of determining an upper subcriticality limit for nuclear criticality safety.

Simulation of steel corrosion and iron yield in LFR with lead coolant

Liquid lead is the main coolant of the fourth generation of advanced nuclear energy system — lead cooled fast reactor (LFR), due to its good neutron economy, high heat transfer performance, stable chemical properties, constant low melting point, high boiling point, etc. Although there are many advantages, the corrosion of metal materials in liquid lead is one of the decisive factors restricting the development of lead cooled fast reactor. In this study, an engineering model for simulating the oxide film growth in liquid lead coolant is established, and the time-dependence of steel flux is analyzed based on the experimental data in the literature. The effects of circuit temperature, hydraulic diameter of section simulated, coolant velocity and steel types on the steel corrosion were investigated. The results showed that the oxide film formed on the steels is of micron grade and the diffusion yield of iron produced in pure lead coolant is of kilogram grade after ten years’ corrosion. The temperature and the coolant velocity have significant effects on the steel corrosion while the effect of hydraulic diameter is mild. Also, the steel types affect the growth of oxide film. The findings provide basic data for the evaluation of radioactive corrosion products of LFR.

Enhanced helium ion irradiation tolerance in a Fe-Co-Ni-Cr-Al-Ti high-entropy alloy with L12 nanoparticles

• Helium bubbles and large stacking faulted loops are observed as the dominant structural damage in the He ion irradiated HEA reinforced by L1 2 nanoparticles. • The L1 2 nanoparticles provide numerous interfaces for He entrapment and damage elimination, which suppresses He bubble growth. • A correlative TEM/APT characterization reveals that the RIS around He bubbles is dominated by the inverse Kirkendall mechanism. • Irradiation-induced dissolution and re-precipitation of the L1 2 nanoparticles can retain the main microstructure of the L1 2 -strengthened HEA and provide a sustainable irradiation resistance. L1 2 -strengthened high entropy alloys (HEAs) with excellent room and high-temperature mechanical properties have been proposed as promising candidates as structural materials for advanced nuclear systems. However, knowledge about their radiation response is fairly limited. In the present work, a novel HEA with a high density of L1 2 nanoparticles was irradiated with He ion at 500°C. Transmission electron microscope (TEM) and atom probe tomography (APT) were employed to study the evolution of microstructural stability and radiation-induced segregation. Similar to the single-phase FeCoNiCr HEA, the main microstructural features were numerous large faulted dislocation loops and helium bubbles. While the irradiation resistance of the present L1 2 -strengthened HEA is much improved in terms of reduced bubble size, which could be attributed to the considerable He trapping efficiency of the coherent precipitate/matrix interface and the enhanced capability of the interface for damage elimination when the matrix channel width is narrow. APT analysis revealed that an inverse Kirkendall mechanism-dominated radiation-induced segregation (RIS) occurs around bubbles, where a significant Co enrichment and Ni depletion can be clearly observed. In addition, the competing dynamics of ballistic mixing and elemental clustering that raised from the irradiation-enhanced diffusion in a highly supersaturated matrix, along with the low precipitation nucleation barrier due to the small lattice misfit, lead to a dynamical precipitation dissolution and re-precipitation appears under irradiation. Such a promising phenomenon is expected to promote a potential self-healing effect and could in turn provide a sustainable irradiation tolerance over the operational lifetime of a reactor.

Studies on high-temperature stability and strengthening mechanisms of high/medium-entropy alloys for potential nuclear applications: The case of FeCrV-based alloys

High-temperature stability and strengthening mechanisms of high/medium-entropy alloys (HEAs/MEAs) have been studied for Ti-added FeCrV-based MEAs. Experimental results show that FeCrVTix MEAs can maintain good high-temperature stability in chemical element, phase structure and mechanical property. A mixed-strengthening-mechanism model dominated by precipitation and fine-grain strengthening has been proposed by comparing experimental results and model calculations of the yield strength increments for the FeCrV-based MEAs owing to Ti addition. The calculated strength increments with this mixed-strengthening-mechanism model closely match the series experimental results. The excellent compressive fracture strain in the FeCrVTix MEAs mainly comes from crack tip plasticity, and Ti addition not only enhances compressive strain by grain refinement, but also improves compressive strain by crack deflection and bridging. Based on this strengthening model, the content, size and distribution of precipitates can be designed and controlled to regulate and optimize the mechanical properties of HEAs/MEAs. Therefore, this study provides new approaches and ideas for the design and development of HEAs/MEAs with high strength and good ductility for advanced nuclear energy systems.

Point defect properties in high entropy MAX phases from first-principles calculations

MAX phase materials exhibit both metal and ceramic properties due to their unique laminated atomic structures, making them promising candidates in advanced nuclear energy systems. Recently, high entropy MAX (HE-MAX) phases have been developed and attracted much attention due to their unique properties. However, the role of chemical disorder in HE-MAX phases in their point defects properties is still not clear. In this work, we investigated the point defect properties in (TiVNb)2SnC, (TiZrHf)2SnC, (TiVNbZrHf)2SnC, and five corresponding single-component M2SnC phases (M=Ti, V, Nb, Zr, Hf) using first-principles calculations. The average vacancy (VM, VSn, VC) formation energies in the HE-MAX phases are (TiZrHf)2SnC > (TiVNbZrHf)2SnC > (TiVNb)2SnC. With the addition of Zr and Hf atoms, the charge transfer between atoms in the HE-MAX phase increases, hindering the formation of these vacancies. Meanwhile, the obtained migration energies show that the migration barrier of VM through Ti is lower than that of V, Nb, Zr, or Hf, while Zr and Hf atoms increase the VC migration barrier due to their large atomic sizes. Additionally, the formation energies of antisite defects in all three HE-MAX phases are lower than the single-component M2SnC phases, indicating that the HE-MAX phases are more resistant to radiation-induced amorphization. This work provides a fundamental understanding of the effect of chemical disorder on point defect properties in MAX phases and proposes a new strategy for designing novel HE-MAX phases with better performance in nuclear applications.

Effects of Pipe Inclination on Global Two-Phase Flow Parameters

Inclined two-phase flow geometries can be found in advanced nuclear reactor systems, such as the helical coil steam generators being considered for use in current integral steam generator designs. While this geometry includes inclination and centrifugal effects coupled together on two-phase flow, there have been limited studies to separate these effects to develop robust models. The majority of two-phase flow research is conducted on vertical channels, with recent work being conducted in a horizontal orientation and limited work in inclined pipes. In the current work, experiments are conducted in an adiabatic two-phase flow test facility to investigate the inclination effect on an air-water flow in straight pipes near atmospheric pressure. The pipe is made of clear acrylic with an inner diameter of 25.4 mm. The inclination of the flow loop can be adjusted in increments of 0.1 deg. Measurement capabilities are included to obtain local two-phase flow parameters such as void fraction, interfacial area concentration, bubble velocity, and Sauter-mean diameter using a local multisensor conductivity probe, local two-phase flow static pressure and pressure drop using a pressure transducer, and flow visualization using a high-speed video camera system. The experimental studies performed in the current work demonstrate how changes in inclination angle can affect the gas distribution flow regime transition and two-phase frictional pressure drop. Based on these experimental results, existing correlations for frictional pressure drop are evaluated, and the modified Lockhart-Martinelli correlation is found to predict the two-phase frictional pressure drop for inclined two-phase flows. This method agrees with experimental data within 7% on average.