Liquid Metal Fast Reactor Research Articles

Assessment of RANS and LES in predicting the flow in a 61-pin wire-wrapped bundle with a porous blockage

One of the most promising nuclear reactor designs is the Liquid-Metal Fast Reactor (LMFR). Its fuel bundle rods are separated with wire spacers that are helically wrapped around each fuel pin. Since this bundle presents a tight rod pitch to diameter ratio, accumulation of debris, cladding deformation or detachment of the wires may generate the formation of porous blockages. Understanding the coolant behavior in LMFR fuel bundles under the influence of blockages is mandatory for a complete nuclear safety evaluation. Therefore, this paper presents Computational Fluid Dynamics (CFD) results using the RANS k-ε Realizable and the LES models to simulate the flow of a 61-pin wire-wrapped bundle with the presence of a porous blockage that obstructed six interior subchannels. The simulated Reynolds numbers were 5,000 (transition regime) and 14,000 (turbulent regime). Results of the first and second order velocity statistics were compared with the Particle Image Velocimetry (PIV) experimental data acquired in the Thermal-Hydraulics Research Laboratory facility of Texas A&M University. The simulations results were found in reasonable agreement with the PIV data. Additional analysis of the flow structures, pressure distribution and power spectral densities were also performed to give more insight of the flow characteristics in LMFR reactors under the presence of porous blockages.

Numerical study of the impact of blockages on the Thermal-Hydraulic performance of 19 and 61-pin Wire-Wrapped bundles with conjugate heat transfer using RANS

Liquid metal fast reactors assemblies with wire spacers present a compact rod pitch to diameter ratio. Therefore, a typical off-normal condition that may occur is the blockage accident, caused by the accumulation of corrosion products or due to the detachment of broken wires. Blockage accidents may threaten the safety performance, leading to fuel rod damage. This work presents a thermal–hydraulic analysis using the RANS k-ε Realizable model to assess the impact of blockages. The model was validated on a blocked 19-pin wire-wrapped assembly comparing cladding and coolant temperatures and pressure drop with experimental data. Then, a numerical comparison between nominal and blocked conditions on a 61-pin wire-wrapped bundle was performed, analyzing solid and porous blockages. The greatest peak temperatures were obtained for the solid blockage case. Both types of blockages resulted in a similar velocity prediction upstream of the obstruction, while major differences were found in the recirculation region.

A study of thermophysical and physicochemical characteristic of the KALLA experiment facility

Activities are currently underway across the world to build heavy liquid metal-cooled reactors. For example, the ALFRED and BREST-OD-300 reactor designs use lead coolant, while the MYRRHA and SVBR-100 reactors use lead–bismuth coolant. The high corrosive and erosive activity of the coolant requires control of oxygen concentration and flow velocity. Due to the complex geometry of the reactors cooling circuit flow, numerical simulation methods are used extensively to justify the operation modes for liquid metal coolants. Correct justification of the complex oxygen transport processes in liquid metals requires the detailed physicochemical computational model, which takes into account the major reactions of oxygen with the coolant and the structural materials.This paper presents a physicochemical model, which includes the following processes: erosion, growth, and dissolution of the two-layer oxide film, precipitation and dissolution of metal oxides in the circuit with subsequent deposition in a filter, and inflow of oxygen by the mass exchanger. The physicochemical model was implemented using the model of passive scalar, which allows simulating the impurity transport in the coolant.Capabilities of the presented model are demonstrated based on the results of investigating thermohydraulic and physicochemical processes at the KALLA laboratory experimental facility. The duration of the simulation is 1000 hr.The spatial distributions of metal oxides concentrations, areas of enhanced erosive activity, as well as mass of oxides deposited in the filter and the amount of oxygen entering the circuit have been obtained. The surface distribution of the oxide film thickness on the test facility surfaces contacting the coolant has also been calculated.Accounting for thermocouples and the manufacturing technology made it possible to improve the accuracy of calculating the thermohydraulic characteristics of the coolant flow compared to earlier studies.

Coupled neutronics/thermal-hydraulic analysis of ANTS-100e using MCS/RAST-F two-step code system

The feasibility of using the Monte Carlo code MCS to generate multigroup cross sections for nodal diffusion simulations RAST-F of liquid metal fast reactors is investigated in this paper. The performance of the MCS/RAST-F code system is assessed using steady-state simulations of the ANTS-100e core. The results show good agreement between MCS/RAST-F and MCS reference solutions, with a keff difference of less than 77 pcm and root-mean-square differences in radial and axial power of less than 0.5% and 0.25%, respectively. Furthermore, the MCS/RAST-F reactivity feedback coefficients are within three standard deviations of the MCS coefficients. To validate the internal thermal-hydraulic (TH) feedback capability in RAST-F code, the coupled neutronic/TH1D simulation of ANTS-100e is performed using the case matrix obtained from MCS branch calculations. The results are compared to those obtained using the MARS-LBE system code and show good agreement with relative temperature differences in fuel and coolant of less than 0.8%. This study demonstrates that the MCS/RAST-F code system can produce accurate results for core steady-state neutronic calculations and for coupled neutronic/TH simulations.

Development of an efficient and improved core thermal-hydraulics predictive capability for fast reactors: Summary of research and development activities at the North Carolina state University

The improved understanding of the safety, technical gaps, and major uncertainties of advanced fast reactors will result in designing their safe and economical operation. This paper focuses on the development of efficient and improved core thermal-hydraulics predictive capabilities for fast reactor modeling and simulation at the North Carolina State University. The described research and development activities include applying results of high-fidelity thermal-hydraulic simulations to inform the improved use of lower-order models within fast-running design and safety analysis tools to predict improved estimates of local safety parameters for efficient evaluation of realistic safety margins for fast reactors. The above-described high-to-low model information improvements are being verified and validated using benchmarks such as the OECD/NRC Liquid Metal Fast Reactor Core Thermal-Hydraulic Benchmark and code-to-code comparisons.

Direct Numerical Simulation of Low and Unitary Prandtl Number Fluids in Reactor Downcomer Geometry

Mixed convection of low and unitary Prandtl fluids in a vertical passage is fundamental to passive heat removal in liquid metal and gas-cooled advanced reactor designs. Capturing the influence of buoyancy in flow and heat transfer in engineering analysis is hence a cornerstone to the safety of the next-generation reactor. However, accurate prediction of the mixed convection phenomenon has eluded current turbulence and heat transfer modeling approaches, yet further development and validation of modeling methods is limited by a scarcity of high-fidelity data pertaining to reactor heat transfer. In this work, a series of direct numerical simulations was conducted to investigate the influence of buoyancy on descending flow of liquid sodium, lead, and unitary Prandtl fluid in a differentially heated channel that represents the reactor downcomer region. From time-averaged statistics, flow-opposing/aiding buoyant plumes near the heated/cooled wall distort the mean velocity distribution, which gives rise to promotion/suppression of turbulence intensity and modification of turbulent shear stress and heat flux distribution. Frequency analysis of time series also suggests the existence of large-scale convective and thermal structures rising from the heated wall. As a general trend, fluids of lower Prandtl number were found to be more susceptible to the buoyancy effect due to stronger differential buoyancy across the channel. On the other hand, the effectiveness of convective heat transfer of the three studied fluids showed a distinct trend against the influence of buoyancy. Physical reasoning on observation of the Nusselt number trend is also discussed.

Study of Stable Stratification in HiRJET Facility With Direct Numerical Simulation

Stable density stratification in a large enclosure could significantly hamper the effectiveness of natural convection cooling in pool-type liquid metal or gas-cooled advanced reactors. In addition, accurate prediction of stratified front behavior remains to be a challenging task for turbulence modeling. With the rapid growth of high-performance-computing capabilities in recent years, conducting high-fidelity simulations for a large-timescale transient has become more affordable and hence a valuable data source to support turbulence modeling as well as to gain further physical insights. In this work, direct numerical simulation is performed at experiment-consistent conditions to simulate the density stratification transient High-Resolution Jet (HiRJET) facility. Specifically, we focus on the case where an injected aqueous sugar solution has 1.5% density higher than that in the enclosure. In the early stage of the transient, the impingement of the denser jet to the bottom surface of the enclosure promoted turbulent mixing locally. This rendered the establishment of the mixture layer, formation and swift upward propagation of the stratified front, and elevation with (locally) the highest vertical concentration gradient. As the front rose, the diminishing turbulent mass flux slowed down the propagation, and a larger vertical concentration gradient was established. In this stage, both the velocity and the concentration scalar showed large-timescale fluctuation behavior around the stratified front. For the concentration time signal, the characteristic frequency in the power spectral density was found to agree well with the Brunt-Väisällä frequency. The preliminary validation endeavor showed that the stratified front location and the corresponding concentration gradient magnitude in the simulation agreed well with the experiment data. Further validation will mainly revolve around benchmarking against high-resolution density measurement and high-order flow statistics.

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.

Experimental and numerical analysis of the liquid metal mixing phenomenon in complex jets of Gen-IV nuclear system

Nuclear energy is essential in addressing the problem of climate change and energy crisis. Liquid metal mixing phenomenon at the core outlet of liquid metal fast reactors is an important issue in Gen-IV nuclear systems. In this paper, the experimental and numerical analysis of the liquid metal mixing process in complex jets were carried out based on the structure of fuel assembly heads in sodium-cooled fast reactors. The liquid metal mixing characteristics were obtained in the experiment. And a high fidelity LES model was developed and validated experimentally. The impact of different head structures on the mixing process was analyzed and an index of liquid metal temperature mixing degree was newly proposed and compared with previous formulas. Result shows that, the temperature mixing region can be divided into five zones along the flow direction. Compared to other three structures, the three-sided outlet head with offset hedge has the lowest fluctuations of velocity and temperature in the upstream of the mixing flow. Also, the three-sided outlet head with offset hedge has the best temperature mixing ability, and the index is two times larger than that of the prototype. The three-sided outlet head with offset hedge proved to be the optimal structure.

Technological Pathways to Produce Compressed and Highly Pure Hydrogen from Solar Power.

Hydrogen (H2 ) produced from renewables will have a growing impact on the global energy dynamics towards sustainable and carbon-neutral standards. The share of green H2 is still too low to meet the net-zero target, while the demand for high-quality hydrogen continues to rise. These factors amplify the need for economically viable H2 generation technologies. The present article aims at evaluating the existing technologies for high-quality H2 production based on solar energy. Technologies such as water electrolysis, photoelectrochemical and solar thermochemical water splitting, liquid metal reactors and plasma conversion utilize solar power directly or indirectly (as carbon-neutral electrons) and are reviewed from the perspective of their current development level, technical limitations and future potential.

Technologische Pfade für die Herstellung von komprimiertem und hochreinem Wasserstoff mit Hilfe von Sonnenenergie

AbstractWasserstoff (H2) aus erneuerbaren Energien wird einen wachsenden Einfluss auf die globale Energiedynamik hin zu nachhaltigen und kohlenstoffneutralen Standards haben. Der Anteil von grünem H2 ist noch zu gering, um das Netto‐Null‐Ziel zu erreichen, während die Nachfrage nach hochwertigem Wasserstoff weiter steigt. Diese Faktoren verstärken den Bedarf an wirtschaftlich tragfähigen H2‐Erzeugungstechnologien. Der vorliegende Artikel zielt darauf ab, die bestehenden Technologien zur Erzeugung von hochwertigem H2 auf der Grundlage von Solarenergie zu bewerten. Technologien wie die Wasserelektrolyse, die photoelektrochemische und die solare thermochemische Wasserspaltung, Flüssigmetallreaktoren und die Plasmakonversion nutzen die Sonnenenergie direkt oder indirekt (als kohlenstoffneutrale Elektronen) und werden im Hinblick auf ihren derzeitigen Entwicklungsstand, ihre technischen Grenzen und ihr Zukunftspotenzial untersucht.

Transient analysis of SIRIO using RELAP5/MOD3.3 system code

The main outcome of the present paper is the feasibility analysis of SIRIO (Sistema di rimozione della Potenza di decadimento per Reattori InnOvativi) facility with conditions based on those of its reference facility. The aim of SIRIO project is to study an innovative Decay Heat Removal System (DHRS) for liquid metal reactor and advanced Light Water Reactor (LWR). Such system must ensure passive control of the power removed from the primary system in abnormal condition, and must ensure reactor cooling in both short and long term. This study present numerical simulations developed with RELAP5/MOD3.3, of two operational procedures: the first one is a steady-state and the second one is a transient phase with decay heat generation. The thermal-hydraulic model, developed with RELAP5/MOD3.3, simulates the whole facility including lines, valves, water and gas tanks, and the Molten Salts (MS) gap. Since there is not experimental data, the present paper is a pre-test study based on SIRO facility design.

Analysis of lead–bismuth eutectic-cooled solid reactor under flow blockage accident by 3D neutronics thermal-hydraulics coupling method

As the most common and serious accident in liquid metal-cooled reactors, the occurrence of a flow blockage accident may threaten the integrity of the cladding and the safety of the core. Therefore, the accurate prediction of fuel and cladding under blockage accidents is particularly important in reactor design and thermal–hydraulic analysis. This paper uses a three-dimensional neutronics thermal-hydraulics coupling method to analyze the blockage accident of a lead–bismuth-cooled solid reactor proposed by the Innovative Nuclear System Laboratory (INSL) of Shanghai Jiao Tong University. Based on the accurate and material models, the solid reactor was modeled, and 15 blockage accidents were carried out. The effects of the blockage's radial position, thickness and axial position on the fuel and cladding temperature were compared. Unlike the conventional open-channel blockage, the results of the closed-channel blockage accident analysis show that the thickness and axial position of the blockage has less influence on the maximum fuel and cladding temperatures but slightly affect the fuel and cladding axial temperature distribution. Meanwhile, the blockage accident in the closed channel is more serious, and the maximum temperature rise of fuel and cladding is higher, but both are lower than the limit value required by the design criteria.

Experimental Analysis of Steam Generator Tube Rupture in CIRCE Facility for MYRRHA Configuration

One of the main safety issues of Generation IV (Gen IV) heavy liquid-metal fast reactors is the postulated steam generator tube rupture (SGTR) accident. This event is characterized by primary and secondary coolant interaction, referred to in the literature as a coolant-coolant interaction event having a nonzero probability to occur. This accident scenario could affect the safety of a pool-type reactor, as a consequence of water secondary coolant flashing into the primary coolant liquid metal. The SGTR event needs to be experimentally characterized to evaluate the pressure waves effect, tube rupture propagation (domino effect), oxide precipitation and slug and plug formation, cover gas pressurization of the reactor, and steam flow paths through the pool and eventually the core, entailing the risk of positive reactivity insertion (due to positive local void coefficient). The design phase of the Gen IV MYRRHA plant has dealt with postulated SGTR safety issues in the framework of the MAXSIMA project, which is supported by the European Commission. A relevant contribution to this research activity was provided by the Italian Agency for New Technologies, Energy and Sustainable Economic Development Research Center Brasimone, where a new test section has been designed, assembled, instrumented, and implemented in the large-scale integral-effects pool facility CIRCE for investigating the SGTR event in a relevant configuration for the heat removal system of MYRRHA. This research reactor is not oriented to steam production for running a turbogenerator (no electric production), thus the heat removal system is referred to as the primary heat exchanger (PHX) and not as a steam generator. Four full-scale portions (four bundles of 31 tubes) of the MYRRHA PHX were adopted to carry out four independent SGTR experiments. Water flowed upward in the central tube of the bundle and two rupture positions were investigated at the lower and upper levels, named the bottom and middle scenarios, respectively. After the rupture, water was injected at 16 bar and 200°C into lead bismuth eutectic alloy at 350°C. The experimental results showed a remarkable repeatability and were presented in terms of (1) CIRCE vessel pressurization up to 2.7 and 4 bar absolute for the middle and bottom scenarios, respectively; (2) vapor flow paths through the bundle and its cooling effect up to 120°C and 140°C for the middle and bottom tests, respectively; and (3) strain measurements on tubes and bundle shells up to 2800 μm/m. The integrity of the tubes surrounding the ruptured one and the effectiveness of implemented safety device (rupture disks) pressure relief were significant engineering feedback for MYRRHA designers. The acquired high-quality data also constitute a database increase for future code verification and validation and possible new model development. The performed experimental analysis provided the awareness that a suitable design of a depressurization system (e.g., rupture disks) could allow for addressing postulated SGTR events in the MYRRHA configuration with confidence and safety.

Demonstration of Pronghorn’s Subchannel Code Modeling of Liquid-Metal Reactors and Validation in Normal Operation Conditions and Blockage Scenarios

Pronghorn-SC is a subchannel code within the Multiphysics Object-Oriented Simulation Environment (MOOSE). Initially designed to simulate flows in water-cooled, square lattice, subchannel assemblies, Pronghorn-SC has been expanded to simulate liquid-metal-cooled flows in triangular lattices, hexagonal subchannel assemblies. For this purpose, the algorithm of Pronghorn-SC was adapted to solve the subchannel equations as they are applicable to a hexagonal wire-wrapped sodium-cooled fast reactor. Cheng–Todreas models for pressure drop and cross-flow models were adopted and a coolant heat conduction term was added. To solve these equations, an improved implicit algorithm was developed robust enough to deal with the numerical issues, associated with low flow and recirculation phenomena. To confirm the prediction capability of Pronghorn-SC, calculations and comparisons with available experimental data of 19- and 37-pin assemblies were performed, as well as other subchannel codes. Finally, a flow blockage modeling feature was added. This capability was validated for both water-cooled square sub-assemblies and sodium-cooled hexagonal sub-assemblies, using experimental data of partially and fully blocked cases.

A Design and Optimization Methodology for Liquid Metal Fast Reactors

A liquid metal fast reactor (LMFR) design and optimization methodology (DOM) has been developed. The methodology effectively explores a search space by initially sampling the search space, excluding invalid design samples prior to performing expensive multiphysics analysis, and then performing local searches of the design space. The design samples are evaluated using the multiphysics capabilities of the LUPINE LMFR simulation suite. Two studies have been performed to demonstrate DOM. First, the Westinghouse long-life core lead fast reactor (WLFR) is optimized. This reactor is 950 MW th and fueled with uranium nitride (UN) fuel which has a natural nitrogen isotopic abundance. The objective of the optimization is the reduction of the levelized fuel cycle cost (LFCC) while complying with the design constraints. Considering the challenges associated with using natural nitrogen in nitride fuel, a second study was performed to design a competitive 15N-enriched UN-fueled long-life core LFR. Based on this design, the cost of the 15N enrichment process necessary to achieve a competitive LFCC was calculated.

Steady-State and Transient Compliance of CIRCE-THETIS CFD Model

The pool-type experimental facility CIRColazione Eutettico (CIRCE) has been built by ENEA in support to the development of liquid-metal fast reactors. A computational fluid dynamics (CFD) model of the new test section, called the Thermal-hydraulic HElical Tubes Innovative System or THETIS, is being built, and is described along with the pretest numerical results. The pretest study analysis supports the dimensioning of the experimental setup and provides a sound but still not complete basis for the experimental transients. Appropriate thermal boundary conditions for the steady state are implemented. These conditions are found to be not suitable for thermal transients, hence the necessary compliance is analyzed and defined, requiring additional preliminary experimental tests. The main novelty in the refurbishment of the CIRCE is the installation of an innovative helical coil steam generator (HCSG). The realization of its numerical counterpart is challenging; therefore, a specific CFD model of the HCSG has been developed separately from the global model. Its preliminary results are described and discussed.

Flow characteristics within an interior subchannel of a 61-pin wire-wrapped hexagonal rod bundle with a porous blockage

Potential accumulation of undesirable debris in a subchannel of a Liquid Metal Fast Reactor (LMFR) hexagonal fuel bundle presents accident conditions, which are crucial to investigate. Very limited experimental research persists in the literature to understand the fluid dynamics effects of partially blocked subchannels due to the presence of porous blockages. It is imperative to comprehend flow regime-dependent fluid response in the vicinity of porous blockages to predict and counter abnormal conditions in an LMFR rod assembly. The presented experimental research investigates flow-field characteristics in a 61-pin wire-wrapped rod assembly with a three-dimensional (3D) printed porous blockage medium in an interior subchannel at Reynolds numbers (Re) of 350, 5000, and 14 000. Time-resolved velocimetry measurements were acquired yielding first- and second-order Reynolds decomposition flow statistics—revealing important fluid response upstream and downstream of the porous blockage. Profiles of velocities, velocity fluctuations, Reynolds stresses, and vorticities uncovered the downstream blockage perturbation effects. Spatial cross-correlations of the velocity fluctuations displayed eddy structure elongations and quantified eddy integral scale lengths. A time-frequency analysis of the velocity fluctuations further detailed the mechanisms of flow instabilities via power spectral analysis. The application of a one-dimensional continuous wavelet transform revealed complex Re-dependent flow and characterized the temporal turbulence occurrences—caused by the trailing edge effects of the porous blockage. This research provides unique and novel experimental analyses on flow regime-dependent fluid physics due to a porous blockage medium and provides data sets vital for computational model benchmarking and development, toward the enhancement of LMFR rod bundle designs.

On physics of a hypothetical core disruptive accident in Multipurpose hYbrid Research Reactor for High-tech Applications – MYRRHA

The sensitivity of the reactivity of a fast reactor core to changes in its geometry and/or fuel relocation calls for particular attention with regard to criticality events. A category of these events, the so-called Core Disruptive Accidents (CDAs), are intensively studied in the safety assessment of Sodium-cooled Fast Reactors (SFRs), and more recently also in the case of other systems. Differences between SFRs and Heavy Liquid Metal Fast Reactors (HLMFRs) are significant and therefore warrant an understanding of phenomena and the development of models specific to HLMFRs. This paper provides a qualitative overview of the physics relevant to the investigation of a CDA in HLMFR, with a particular application to the Multipurpose hYbrid Research Reactor for High-tech Applications – MYRRHA. At first, a core compaction mechanism viable for an HLMFR has been postulated. In what follows, simulation by an already existing severe accidents code, as well as modelling based on fundamental physics and engineering, have been performed. It is demonstrated that, for a linear insertion of reactivity due to hypothetical core compaction, the reversal of reactivity evolution happens due to the Doppler effect and the thermal expansion of core materials. Subsequent expansion by fuel melting terminates the prompt-critical event and makes the system delayed-supercritical. Successive fuel and/or coolant boiling is responsible for the hydrodynamic disassembly of the core and it therefore effectively terminates the transient.

Numerical study on inter-wrapper flow and heat transfer characteristics in liquid metal-cooled fast reactors

Inter-wrapper flow (IWF) is a typical thermal hydraulic phenomenon that exists in narrow gaps among the adjacent assemblies of a liquid metal fast reactor (LMFR) that can dissipate heat from assemblies and transfer heat among adjacent assemblies. Moreover, when blocked or reduced flow accidents occur, the IWF has a significant effect on reactor safety, especially reducing the peak cladding temperature. In this study, a computational fluid dynamics (CFD) model of the Karlsruhe liquid metal laboratory IWF experimental setup (called MYRRHA design) is built. The model contains three fuel assemblies with seven rods using lead–bismuth eutectic as working fluid. The new mesh generation method, correlation of turbulence model, and fluid–solid coupling heat transfer were considered in the CFD model. The established model was initially validated, and the simulation results were found to be in good agreement with experimental data. The influence of IWF on the thermal hydraulic characteristics within the rod bundle was revealed. The results indicated that the IWF rate can affect the heat transfer capability of the bundle under. The amount of heat removed increased by up to 10% with the IWF rate in a single bundle. The occurrence of IWF reduces the wall surface temperature and appropriately balances the temperature of adjacent bundles; however, it cannot effectively reduce the peak temperature of the cladding. The complex heat transfer characteristics and flow mechanism under the influence of IWF and wire mixing were revealed. The work presented in this paper represents an important step toward the IWF research on blocked flow scenarios.