Pressurized Water Reactor Design Research Articles

Fluidelastic instability of a fuel rod bundle subjected to combined axial flow and jet cross-flow

Nuclear fuel bundles are subjected to axial coolant flow in addition to jet cross-flow in a specific design of pressurized water reactors (PWRs). The combined axial flow and jet cross-flow was found to induce fretting wear of the fuel rods at the spacer grid position. Investigating a reduced fuel rod bundle under jet in transverse flow (JITF) condition is necessary to determine safe vibration free operation conditions for these reactors.This paper presents an experimental and theoretical framework to understand the fluidelastic behavior of a rod bundle subjected to axial flow and localized jet cross-flow. A 6x5 axisymmetric flexible single-span rod bundle is used to simulate a part of a fuel assembly under combined flow. Two working flow conditions, pure axial flow, and JITF are studied on a mock-up PWR assembly. The experiments show that the arrays are stable under pure axial flow as expected. Then, the responses of the rod bundles under combined flows (i.e. JITF) are measured at three test axial flow velocities, VAxial= 1.0 m/s, 1.5 m/s, and 2.0 m/s. The results show that the critical jet cross-flow velocity increases linearly with the axial flow velocity with a velocity ratio (VR = VJet/VAxial) of 1.6.The second part of the work is the development of a fluidelastic instability model to predict the stability boundary for rod bundles under combined axial flow and jet cross-flow. The effect of axial flow on the jet cross-flow is introduced by implementing an incidence angle (θ) of jet flow with respect to the rods. The fluid-added damping terms in the developed model are shown to be functions of θ. The fluidelastic forces are expressed as functions of the projected rod area derivative to account for rod vibration. The stability analysis shows the capability of the model to predict the trend of critical jet cross-flow velocity with the axial flow velocity. The model validation shows the predicted critical velocities to be within an absolute error range from 7% to 12.5% when compared to experiments. The theoretical analysis highlights the importance of the cross-coupling fluidelastic forces in predicting the instability of the mock-up PWR assembly.

Nondestructive Characterization and Monitoring of Cracking in Pressurized Piping Using Surface Strain Measurements

Primary loop piping integrity is a regulatory and safety concern for the long-term operation (LTO) commercial light water reactor (LWR) nuclear power plants (NPPs). Typically, primary loop piping in existing LWRs was manufactured from cast or wrought austenitic stainless steel alloy material and operates at a pressure of ~2250 psi (16MPa) for a pressurized water reactor (PWR) design type and ~1000 psi (7 MPa) for a boiling water reactor (BWR). The objective of this research was to determine if measurement of surface strains can be used to characterize inner diameter flaws. For this feasibility study, wrought carbon steel Alloy A106 Grade B material was selected in lieu of stainless steel. This research was performed on pre-notched, 10 in. (25 cm) nominal outer diameter (OD) Schedule 120 (0.844 in. [21.4 mm) wall thickness) pipe samples of 24 in. (61 cm) length. The notches were made using electrical discharge machining (EDM) on the inner diameter (ID) of the pipe samples, with notch features two-dimensional (2D) and planar in nature to simulate cracking. The surface strains in the samples were evaluated under increasing hydrostatic pressure, up to a maximum of 2300 psi (16MPa) to determine their response to flaws with variable depth. Reusable end caps were utilized in lieu of welded caps for the pressure testing. This article describes the test procedure for the pressure testing and results of the feasibility study, which is the first phase of research with the end goal of the development of online crack monitoring capability for use at operating NPPs.

Neutronics analysis of UN-PuN fuel for 300MW pressurized water reactor using SRAC-COREBN code

Nuclear Power Plants (NPP) is one of alternative energy that can be used to replace fossil energy. NPP is a clean energy that doesn't emit CO2 as a residue from the power plant. Pressurized Water Reactor (PWR) is the most widely used commercial reactor to date. This reactor usually uses UO2 as fuel with enrich U-235 and produce plutonium as a waste nuclear. In this study, the fuel uses natural uranium and the addition of plutonium from the spent fuel of PWR, calling it Uranium-Plutonium Nitride (UN-PuN) fuel. Plutonium from spent fuel PWR (U-fueled) consists of Pu-238, Pu-239, Pu-240, Pu-241 and Pu-242, which Pu-239 and Pu-241 are fissile material. The aim of using plutonium is to reduce the amount of nuclear waste and prevent the use of plutonium as a nuclear weapon. The addition of neptunium and protactinium is also carried out in the fuel. Neptunium is one of the minor actinide materials which have high toxicity, so add Np-237 it is very useful to reduce the amount of neptunium in the world. And also, the addition of Np-237 and Pa-231 aims to decrease the keff value both in the Beginning of Life (BOL) and End of Life (EOL) in the reactor. This research calculates the neutronic calculations for PWR use SRAC2006 with COREBN Code and Japanese Evaluated Nuclear Data Library 4.0 (JENDL-4.0) for the nuclear data library. The COREBN code is a calculation method based on interpolation of macroscopic cross-section and finite-difference diffusion theory. The optimum design reached the maximum k-eff value 1.017077 and the maximum power density was 58.47 W/cc. The results obtained are the excess reactivity value of 1.68 % smaller than previous study which produces an excess reactivity value of 11.94 %. So this design has the potential to be applied in the PWR design

Aerosol retention within submerged Containment-Type iPWRs: Experiments and modeling

Integral pressurized water reactor (iPWR) designs employing submerged, high-pressure steel containment vessels rely on water in the outer operating pool as the primary heat sink during an accident. This drives very strong steam condensation, and along with it, strong diffusiophoretic deposition rates for fission product aerosols. In this study, a loss-of-containment isolation accident is proposed and simulated with MELCOR, where the strong condensation/diffusiophoresis is the last barrier preventing direct radionuclide releases to the environment. This is supported by experiments that measured the thermal stratification, steam condensation rates, and aerosol deposition rates that would be experienced in a prototypic, iPWR-like geometry. The development of the steam layer, whose thickness was proportional to steam injection rates, impacted the aerosol deposition by limiting the area over which condensation-driven diffusiophoresis could occur. Diffusiophoresis was still confirmed to be the primary mechanism for aerosol deposition, with observed rates that were in line with established theory.

Investigating various factors affecting neutron spectrum in the pressure tube in pressurized heavy water reactor design

The fuel bundles in the pressurized heavy water reactor design (PHWR), such as the CANDU (Canada Deuterium Uranium), are placed inside horizontal fuel channels. Each fuel channel consists of concentric tubes called the pressure tube (PT) and the calandria tube (CT). Due to continuous exposure to extreme neutronics and thermal hydraulics conditions, the PTs undergo physical deformations in both the axial and radial directions, as the reactor ages. The deformations include diametral creep, wall thinning, sag, and axial elongation. One of the contributing factors to these deformations is the neutron spectrum in the PT, especially neutrons with energy greater than 1 MeV. In this paper, the sensitivity of the neutron spectrum in the PT to various operating conditions of a PHWR is evaluated. Sensitivity assessments have been completed on several parameters namely the coolant purity, coolant temperature, moderator purity, moderator temperature, and pressure tube diametral creep (PTDC). Since the deformations are inevitable, identifying parameters that have the greatest impact on the population of neutrons with energy greater than 1 MeV in the PTs would be important for fine-tuning the operational conditions which are conducive to reducing the rate of deformation. From this study, it is concluded that the coolant temperature and PTDC are the two variables having the most impact on the fast energy neutron population in the pressure tubes.

A numerical study of heat transfer in bottom-heated and side/top-cooled liquid metal layers with different aspect ratios

A liquid metal layer heated from bottom and cooled from both side and top can be encountered in industrial applications. A special interest is from safety design of advanced pressurized water reactors that adopt the so-called in-vessel melt retention (IVR) to mitigate severe accident risk. Quantification of heat transfer in a stratified melt pool in the lower head of a reactor pressure vessel (RPV) is of great importance to the qualification of the IVR strategy. The upper liquid metal layer of the stratified melt pool is heated by the lower molten oxide layer (with decay heat) underneath, and cooled by water outside the reactor vessel and by radiation or flooded water at the top. This is essentially a problem of natural convection and heat transfer in a liquid metal layer heated from bottom and cooled from both side and top. The present study is conducted to numerically investigate the heat transfer characteristics of such layer with an emphasis on the influence of the aspect ratio (ratio of radius to height; R/H) of the liquid metal layer. Based on the numerical outcomes, three correlations of heat transfer coefficients (for downward, upward and sideward flows) are also developed to account for the impact the aspect ratio. The numerical simulation results show that, under the same Rayleigh number, the bulk temperature and the upward and sideward heat fluxes all increase with R/H, but the downward heat flux decreases with R/H. The Nusselt numbers in all directions decrease with increasing R/H, as a reduced cooling-heating area ratio due to increasing R/H shall suppress the cooling efficiency and the convection. When R/H is larger than a threshold (∼8), the heat transfer characteristics are no longer sensitive to R/H. Each correlation of heat transfer coefficient is developed as the product of two terms: a base correlation of heat transfer coefficient that is Ra dependent only, and an aspect ratio factor that considers the effect of aspect ratio R/H. The developed correlations are compared with the numerical simulation results of cases with different aspect ratios and Rayleigh numbers, and good agreements achieved.

Demonstration of the E-BEPU methodology for SL-LOCA in a Gen-III PWR reactor

The paper presents the first practical application of the alternative Extended Best Estimate Plus Uncertainty (E-BEPU) methodology. The E-BEPU is the systematic risk-informed combined deterministic and probabilistic methodology dedicated to design verification of Nuclear Power Plants. The approach covers applying the best-estimate computer code, realistic input data for initial and boundary conditions with uncertainties, and plant systems availability based on Probabilistic Safety Analysis methods. Furthermore, it extends the scope of the uncertainty analysis to include the availability of safety systems as an additional uncertain item. The approach is characterized by an enhanced Defence-in-Depth; it can detect cliff-edge effects and include risk insights in the deterministic analysis framework. It is an advanced approach to Deterministic Safety Analysis beyond the BEPU approaches and an alternative realization of the Extended-BEPU. In this study, the generic Generation-III Pressurized Water Reactor design was investigated to demonstrate the methodology with a practical example. The Postulated Initiating Event was a surge line double-ended break (SL-LOCA), a variant of the Large Break Loss of Coolant Accident (LB-LOCA), and considered an example of the Design Basis Event. The considered reactor design passed successfully through the procedure with a large margin due to the robustness of its safety systems.

Development of a Critical Heat Flux Correlation based on a Mechanistic Model under Subcooled Flow Boiling Conditions

In this study, a mechanistic model-based correlation for the critical heat flux is proposed to meet the practical requirements for the design and safety analysis of pressurized water reactors. The correlation is based on a liquid sublayer dryout model for a single tube, which includes physical parameters, such as sublayer thickness and vapor blanket velocity, expressed as an explicit nondimensional equation and modified using a look-up table for critical heat flux to ensure a wide applicable range in coolant conditions. The correlation predicts critical heat flux for the subcooled region under the pressure range of 0.1–18 MPa, mass velocity range of 500–8000 kg/m2s, and thermal equilibrium quality ranging from −0.5 to −0.05 with high accuracy of the departure from nucleate boiling ratio limit based on a 95% probability and a 95% confidence of 1.20.

Evaluating the variations of point kinetics parameters in pressurized heavy water reactor analyses

Since the pressurized heavy water reactor design utilizes natural uranium bundles as its fuel, there is a need to keep replenishing the reactor core with unirradiated fuel bundles in order to keep the reactor critical. This practice causes the fuel inventory in the core to be very dynamic since the changes are not solely due to the depletion of the fuels due to irradiation but also due to the presence of new fuel bundles in the core. An average behaviour of the core after operating in an equilibrium stage for a long time can be captured through the time-average approach. This approach has been extensively used in the pressurized heavy water reactor fuel management, especially during the design phase where the actual data from the operating reactor are not readily available. The time-average core is utilized for many purposes such as to determine the average fueling rate, to determine the average worth of various reactivity control devices, as well as to determine the point kinetics parameters for transient analyses. The objective of this paper is to quantify potential variations in the point kinetics parameters (namely the decay constants and the delayed neutron fractions) as a function of core inventories with respect to the time-average-based values. In particular, 100 core configurations representing various snapshots from an equilibrium operation of the core have been utilized to quantify these potential variations. It should also be noted that the point kinetics equation (PKE) is utilized in this study, instead of full-core space–time model, since the objective of this paper is to bring about the awareness of potential variations and the PKE should be sufficient. The results of this study indicate that, for a point kinetics equations system involving six precursor groups, the variations in precursor decay constants and delayed neutron fractions could be as high as ±0.45% and ±4.85%, respectively. Using the point kinetics parameters in a simple transient involving an insertion of positive reactivity, it is demonstrated that the variation in the trip time for the transient could vary up to approximately ±3.55%, depending on the amount of positive reactivity insertion. These potential variations should be acknowledged since it is an important part of safety analysis. However, it is also important to note that, based on the amount of positive reactivity insertion evaluated in this study, the higher variation corresponds to a lower level of positive reactivity insertion which means that the transient will be slowly developing, could be observed more readily, and, therefore, could be easily terminated.

Comparing the dynamic response of U-tube and helical coil steam generators in small modular reactor flexible operation

A steam generator (SG) is one of the most important subsystems in a pressurized water reactor (PWR). As the interface between the reactor and the turbine for power/energy transform, its behavior dictates the response of the entire plant, especially when maneuvering the power output under dynamic operation. Generally, two types of SGs are most commonly employed in PWR designs: a recirculation-type U-tube steam generator (UTSG) and a once-through steam generator (OTSG). Among the PWR-type SMR concepts, some designs use a UTSG while several others rely on the helical coil steam generator (HCSG), which is a type of OTSG. These two designs differ in a variety of ways, namely construction, physical size for a given capability, thermodynamic characteristics, and control techniques – leading to a very different dynamic behavior under flexible operation. This paper examines the dynamic characteristics of a UTSG and a HCSG in the context of SMR flexible operation and highlights the key implications of choosing one against the other. Detailed moving boundary models are developed for both SGs. Simulation studies have been carried out in MATLAB/Simulink environment to compare their characteristics quantitatively under different loading conditions. The results have indicated that there are significant differences in dynamic behaviors between these two SGs during transient operations. Relatively larger pressure excursions in the UTSG during power maneuvers suggest more stringent considerations in equipment design and selection for safe operation.

Burnable Poison Selection and Neutronics Analysis of Plate Fuel Assemblies

Burnable poisons play a critical role in long-life pressurized water reactors. Plate fuel elements have good application prospects in long-life pressurized water reactors. In long-life pressurized water reactors with large initial residual reactivity in the core, a reasonable selection of burnable poisons can suppress the large residual reactivity at beginning of lifetime and can achieve a long burnup depth at end of lifetime. Therefore, the selection of burnable poisons is a crucial factor to be considered in the design of long-life pressurized water reactors. In this study, the selection of burnable poisons and neutronics characteristics of long-life PWR plate fuel assembly were studied. The transport-burnup calculations of different burnable poison fuel assemblies were carried out. Some candidate BPs are selected to realize the effective control of reactivity. The results show that when the enriched isotopes 157Gd, 167Er and B4C are used as burnable poisons, there is almost no reactivity penalty; when PACS-J and 231Pa are used as burnable poisons, due to their own characteristics, not only does not cause reactivity penalty at end of lifetime, but also the fuel assembly life is extended, the fuel utilization rate is improved. The combination of PACS-J and the slow-burnup burnable poisons can obtain a better reactivity curve. The results of this article show that the plate fuel assemblies can be selected with enriched isotope 157Gd, enriched isotope 167Er, B4C, 231Pa and PACS-J as burnable poisons, and the combinations of burnable poisons can be selected with two combination schemes, PACS-Er and PACS-Pa.

Analysis of the Technological Evolution of Materials Requirements Included in Reactor Pressure Vessel Manufacturing Codes

The growth of green energy technologies within the frame of the 7th Sustainable Development Goal (SDG) along with the concern about climatic changes make nuclear energy an attractive choice for many countries to ensure energy security and sustainable development as well as to actively address environmental issues. Unlike nuclear equipment (immovable goods), which are often well-catalogued and analyzed, the design and manufacturing codes and their standardized materials specifications can be considered movable and intangible goods that have not been thoroughly studied based on a detailed evaluation of the scientific and technical literature on the reactor pressure vessel (RPV) materials behavior. The aim of this work is the analysis of historical advances in materials properties research and associated standardized design codes requirements. The analysis, based on the consolidated U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.99 Rev.2 model, enables determination of the best materials options, corresponding to some of the most widely used material specifications such as WWER 15Kh2MFAA (used from the 1970s and 1980s; already in operation), ASME SA-533 Grade B Cl.1 (used in pressurized water reactor-PWR 2nd–4th; already in operation), DIN 20MnMoNi55 and DIN 22NiMoCr37 (used in PWR 2nd–4th) as well as ASTM A-336 Grade F22V (current designs). Consequently, in view of the results obtained, it can be concluded that the best options correspond to recently developed or well-established specifications used in the design of pressurized water reactors. These assessments endorse the fact that nuclear technology is continually improving, with safety being its fundamental pillar. In the future, further research related to the technical heritage from the evolution of materials requirements for other clean and sustainable power generation technologies will be performed.

On the validation of BEPU methodologies through the simulation of integral experiments: Application to the PKL test facility

Validation is a key step in the field of thermal hydraulics and has played a crucial role in the implementation of system codes. These codes are then included in licensing methodologies being those either best estimate plus uncertainties (BEPU) or conservative approaches. However, only a small amount of efforts has been done on the validation of the methodologies as a whole. Such validation would imply the comparison of the bounding limits and the experimental data. Due to the lack of actual data of accidental situations, such validation should be carried out with data from integral test facility (ITF) experiments.The OECD/NEA PKL-4 test programme is investigating safety issues relevant for current pressurized water reactor (PWR) plants as well as for new PWR design concepts focusing on complex heat transfer mechanisms under two-phase flow. Within the project, a blind benchmark activity has been launched where an IBLOCA experiment has been simulated by 12 different organizations. The Universitat Politècnica de Catalunya (UPC) participated in the activity with a blind BEPU approach with the main objective of assessing the validity of the methodology as a whole.At first, the scalars that characterize the phenomenology of the particular scenario have been considered. Secondly, in order to broaden the extent of the validation, the results of the different participants have been interpreted as if they were blind experiments as well. Finally, different output parameters have been considered providing a matrix of results that is used to evaluate the validity of the methodology.

Dynamic modeling and two-degree-of-freedom controller design for pressurized water reactor core

The task of this investigation is to model pressurized water reactor core and design a reactor power controller that has a strong robustness in various operating conditions. The comparisons of different nodal models of the reactor core, which includes point kinetics, heat transfer between fuel and coolant, and reactivity feedbacks, indicate that the 2C/1F finite-difference nodal model with four fuel nodes has sufficient transient response accuracy. Based on the transfer function model that is obtained by linearizing the nonlinear model, a two-degree-of-freedom controller is proposed to reduce the effect of power level variation on the system dynamics. The simulations demonstrate that the proposed two-degree-of-freedom controller can give satisfying performance for both reactor power and coolant temperature over a wide range of reactor operation.

Analysis on CHF deviation caused by geometry of the heater blocks in a validation test of IVR-ERVC technology

The innovative technology of In-Vessel Retention – External Reactor Vessel Cooling (IVR-ERVC) is a mitigation countermeasure for severe accidents, which has been applied in the design of advanced pressurized water reactors. The validity of IVR-ERVC is restricted to boiling crisis and critical heat flux (CHF) of the reactor vessel lower head. A cylindrical slice heater block was usually employed to simulate the reactor vessel lower head in the majority of existing experiments, which may cause some uncertainties in CHF under natural circulation conditions. In the present paper, an experiment was constructed to study the characteristics of natural circulation and critical heat flux. Apart from the conventional cylindrical slice, a spherical slice was also employed as the heater block. The experimental data indicated that the natural circulation flow rate increases with increasing inlet water temperature, and shows large oscillations with the lift of circulation height. Besides, CHF decreases with the increase of inlet temperature. It was found that the CHF in the spherical slice is lower than that in the cylindrical slice from 10.3% to 28.4%, depending on the inlet water temperature. For the first time, the CHF uncertainty caused by the geometrical difference of the heater block was clarified.

Analytical investigation on rod bundle CHF-regime criterion based on dimensionless groups

The accurate prediction of rod bundle critical heat flux (CHF) is an important task for thermal-hydraulic design of pressurized water reactor (PWR) core. Under normal and transient conditions of PWR, based on the two-phase flow pattern, there are different CHF regimes. In order to correctly distinguish the CHF regimes in rod bundle, it is necessary to analytical investigate rod bundle CHF-regime criterion. In this study, a general dimensionless rod bundle CHF correlation is presented based on dimensional analysis method. The real vapor quality, Katto number and boiling length are introduced to establish the basic form of the dimensionless CHF correlation. According to the deep analysis of the distribution trend of typical CHF experimental data and the CHF characteristics of different CHF regimes, the CHF in rod bundle are separated into DO, DNB and HP-DNB regimes. Concise dimensionless CHF correlations corresponding to each regime are developed, and the dimensionless equations between the boundaries of each regime are established. Also, a clearly defined map of characteristic regimes of CHF in rod bundle is determined. This study proposes a classification criterion for rod bundle for rod bundle CHF regimes, and it is very important for the upcoming development of rod bundle CHF mechanistic model.

Multiobjective genetic algorithm strategies for burnable poison design of pressurized water reactor

Multiobjective genetic algorithm strategies for burnable poison design of pressurized water reactor

A Global Review of PWR Nuclear Power Plants

Nuclear energy is presented as a real option in the face of the current problem of climate change and the need to reduce CO2 emissions. The nuclear reactor design with the greatest global impact throughout history and which has the most ambitious development plans is the Pressurized Water Reactor (PWR). Thus, a global review of such a reactor design is presented in this paper, utilizing the analysis of (i) technical aspects of the different variants of the PWR design implemented over the past eight years, (ii) the level of implementation of PWR nuclear power plants in the world, and (iii) a life extension scenario and future trends in PWR design based on current research and development (R&D) activity. To develop the second analysis, a statistical study of the implementation of the different PWR variants has been carried out. Such a statistical analysis is based on the operating factor, which represents the relative frequency of reactors operating around the world. The results reflect the hegemony of the western variants in the 300 reactors currently operating, highlighting the North American and French versions. Furthermore, a simulation of a possible scenario of increasing the useful life of operational PWRs up to 60 years has been proposed, seeing that in 2050 the generation capacity of nuclear PWRs power plants will decrease by 50%, and the number of operating reactors by 70%.

Three-dimensional (X-Y-Z) Core Design of Long-Life Pressurized Water Reactor Using (Th-U)O2 Fuels with The Addition of Gd2O3 and Pa-231 as Burnable Poisons

Pressurized water reactors (PWRs) are one of the most dominant types of nuclear power plants that have been operated commercially to produce electricity in the world. The purpose of this study was to perceive a three-dimensional (X-Y-Z) core design of long-life PWR using Thorium-Uranium dioxide ((Th-U)O2) fuels with the addition of Gadolinium (Gd2O3) and Protactinium-231 (Pa-231) as the burnable poisons. A combination of Thorium and enriched Uranium fuels have a higher conversion ratio than other fuels, therefore can guarantee the reactor to operate longer. The burnable poison isotopes could be used to reduce excess reactivity due to the very high thermal neutron absorption cross-section. For core geometry analysis, a three-dimensional (X-Y-Z) geometry and a fuel volume fraction of 40% were applied. The computer code of SRAC 2006 from the Japan Atomic Energy Agency (JAEA) and the JENDL 4.0 as a nuclear data library were used for calculation. In this study, different fractions of Uranium dioxide, Uranium-235, Gadolinium, and Protactinium-231 in fuel were carried out. The result of this study was a three-dimensional core design of 800 MWt PWR using 60% Uranium dioxide fuel with enriched Uranium-235 of 12%-11% and the addition of 0,025% Gd2O3 and 1,0% Pa-231 which could operate for ten years without refueling. This research is expected to be a reference for long-life PWR design using the Thorium and Uranium fuel cycles.

CFD simulation on the flow characteristics in the PWR lower plenum with different internal structures

The detailed thermal-hydraulic phenomena inside the lower plenum of Reactor Pressure Vessel (RPV) in Pressurized Water Reactor (PWR) affect the mass flow rate distributions at core inlet significantly, leading to an impact on the operation safety of nuclear reactors. From the engineering point of view, a homogeneous flow rate distribution at the core inlet is expected. Actually, series of design concepts of lower plenum internal structures were proposed for different PWRs in the world. In this paper, the three-dimensional CFD simulations with three types of internal structures in the lower plenum, TYPE 1 (Vortex suppression plate with the secondary supports), TYPE 2 (Flow distribution skirt board with square coolant holes), TYPE 3 (Ellipsoid perforated drums and supports), were performed. The influence of different internal structures on the flow characteristics in the lower plenum were achieved and analyzed. The three types of internal structures were adapted into the same experimental facility “BORA”, which is a 1/5th scale 4-loops mock-up model, to realize the single variable criterion and obtain a uniform standard during the comparison. Results indicate that the mass flow rates in central fuel assemblies are much higher than that in peripheral fuel assemblies at core inlets with TYPE 1 and TYPE 2 internal structures, while the core inlet mass flow rate distribution with TYPE 3 internal structure is more homogeneous. This research has high reference value for the future advanced PWR lower plenum internal structures optimization and design.