Subchannel Analysis Code Research Articles

Coupled simulation for pin-wise reactor core analysis using subchannel analysis code CUPID and fuel performance code GIFT

Recent efforts to develop high-fidelity, multi-physics safety analysis methodologies aim to assess realistic safety margins. However, they remain conservative using initial conditions based on the hot rod with maximum power for the whole reactor core. This study presents the development and verification of a coupled code for pin-wise reactor core analysis using the subchannel analysis code CUPID and the fuel performance code GIFT. The goal was to generate accurate pin-wise fuel rod conditions during normal operation for safety analysis. CUPID was applied for thermal–hydraulic analysis considering fuel behavior, and GIFT was extended to multiple fuel rods. The coupled code was verified for a single assembly to examine subchannel geometry deformation, followed by a practical simulation for a quarter reactor core during the first cycle. Key parameters for safety analysis were examined, revealing that maximum stored energy occurred in a specific fuel rod rather than the hot rod with maximum power.

Two-group drift-flux model in tight lattice subchannel

Tight lattice rod bundles can increase equipment compactness and improve heat exchange efficiency, and they are widely adopted in heat exchange engineering. Understanding the flow characteristics of gas-liquid two-phase flow in tight lattice subchannels is essential for developing heat exchangers and fuel assemblies with these tight bundles. The drift-flux model (DFM) is a crucial two-phase flow model extensively used in subchannel analysis codes. Investigating the DFM in tight lattice subchannels benefits the advancement of these codes. For two-phase flow interfacial structures, significant two-group characteristics exist, with bubbles divided into small (group-one) and large (group-two) bubbles. The substantial differences in flow characteristics between these two groups provide a solid foundation for developing a two-group DFM. This study examined the two-group characteristics of two-phase flow in a tight lattice interior subchannel and developed a corresponding two-group DFM. The two-group correlations for the distribution parameters and drift velocities at the tight lattice interior subchannel level were proposed and verified using experimental data. The accuracy of the developed two-group DFM in predicting group-wise void fraction and gas velocity was also verified. The standard relative deviation between the model predictions and experimental data was 8.11 % and 13.1 % for group-one and group-two void fractions and 8.33 % and 11.7 % for group-one and group-two gas velocities, respectively. This indicates satisfactory accuracy for the newly developed two-group DFM in a tight lattice interior subchannel.

Estimation of Turbulent Mixing Factor and Study of Turbulent Flow Structures in Pressurized Water Reactor Sub Channel by Direct Numerical Simulation

Abstract Subchannel analysis codes are presently a requirement for design and safety analysis of nuclear reactors. Among the crucial inputs for these codes, the turbulent mixing factor holds particular significance. However, acquiring this factor through experimental means proves to be a challenging endeavor, primarily due to the necessity for precise pressure equilibrium between subchannels. Consequently, this requirement leads to the undertaking of expensive and intricate experiments for each new reactor or in cases where there are modifications in fuel bundle design. The need for direct numerical simulation (DNS) stems from the challenges and costs involved in experimental techniques, and the uncertainties due to empiricism in computational fluid dynamics (CFD) models. In this study, DNS has been conducted across six Reynolds numbers, ranging from 17,640 to 1.176 × 105, in the geometry of a pressurized water reactor (PWR) subchannel. The resulting turbulent flow structures have been computed and their dynamics are examined. Furthermore, this paper presents a methodology for directly calculating the turbulent mixing factor from the fluctuating velocity field obtained from DNS data. The turbulent mixing process has been scrutinized in-depth, and a correlation for the turbulent mixing factor is developed. It is noted that most of the mixing occurs in the near-wall region. The study suggests different mixing factors for mass and momentum mixing. This paper aims to provide a comprehensive insight into the turbulent mixing phenomenon.

A subchannel analysis code for advanced liquid metal fast reactor cores and study on heat transfer characteristics of core geometry parameters

The liquid metal fast reactor represents an important reactor type for the future development of nuclear energy. Accurately modeling and predicting the subchannel flow and heat transfer phenomenon in the rod bundles plays a key role in ensuring core safety. Consequently, a subchannel code suitable for the liquid metal fast reactor is analyzed based on the subchannel analysis code of the Pressurized Water Reactor (PWR). The modified code incorporates various types of liquid metals, such as lead, lead-bismuth eutectic (LBE), and sodium, along with their constitutive models, enhancing the applicability of the liquid metal reactor systems. This code has been verified and validated at several levels, including comparing the code simulation results with other similar subchannel codes results, high-dimensional Computational Fluid Dynamics (CFD) simulations, and experiment data. The sensitivity analysis of core geometric parameters and turbulent mixing coefficients is performed based on the modified version code. The influence of the core geometry, including the rod Pitch to Diameter ratio (P/D), Rod to Wall gap (RTW) and the wire wrap pitch (H) on the sensitivity of outlet temperature has been investigated. The results show that the new subchannel analysis code suitable for liquid metal cooled reactor shows reasonable agreement with the verified and validated results. The geometric parameters, such as P/D, RTW and H have noticeable effects on the outlet temperature difference of the subchannels. Additionally, the outlet temperature distribution in the subchannel is significantly affected by different turbulent mixing coefficients and the distribution of the turbulent mixing coefficient also influenced by the reactor core sizes. Overall, this study could support the rod bundles design and safety analysis in the liquid metal reactors.

Mechanistic prediction of Westinghouse TRITON11® BWR fuel critical power with MEFISTO-T subchannel analysis code

The TRITON11® fuel design is the latest Boiling Water Reactor (BWR) fuel product developed by Westinghouse, based on an 11 × 11 optimized fuel rod lattice, including mixing vane spacer grids, three large water rods and 18 part-length rods. The design offers larger fuel cycle cost saving, improved fuel reliability and increased thermal margin over previous Westinghouse fuel products. The critical power performances of the TRITON11 fuel design were assessed at the Westinghouse thermal–hydraulic FRIGG loop using a full-scale test bundle covering a wide range of BWR core conditions (covering normal operation and Anticipated Operational Occurrences) with various radial and two axial power distributions. The resulting FRIGG steady state database was simulated with Westinghouse subchannel analysis code MEFISTO-T based on a two-phase three-field approach of annular two-phase flow, accounting for the drop deposition enhancement provided by the spacer grids. A new model of local film entrainment was introduced due to the liquid “scrapping off” effects provided by the frame and side vanes of the spacer grids along the fuel channel and water rods. After spacer grid effect calibration, the steady state critical power is simulated mechanistically by power iterations up to complete local film dryout. The MEFISTO-T code can successfully predict the critical power performance of the Westinghouse TRITON11 BWR fuel design, including on part-length rods, with almost no bias and trend, within a standard deviation of about 5 %.

Development of a liquid metal subchannel code applied to ocean conditions and study on the effect of core geometry parameters and ocean motions on flow and heat transfer characteristics

The Generation IV International Forum (GIF) proposes a type of liquid metal reactor, due to its thermal properties and efficient heat transfer in a relatively small system. This special heat transfer characteristic allows the liquid metal reactor to be modular in design and flexible in installation on a nuclear-powered ship. Accurate prediction of the coolant temperature and flow distribution between subchannels is important to ensure nuclear safety in the complex ocean environment. In this study, an ocean version of subchannel code is developed by adding the motion terms in the momentum equations for the liquid metal reactor core based on the previously land-based subchannel code. Since the research on the characteristics of flow resistance and heat transfer mechanism under ocean conditions is not mature, the original empirical model is still used in the ocean version code. The constitutive models of the liquid metal are analyzed and incorporated into the ocean code. The code has been verified and validated by comparing with Computational Fluid Dynamics (CFD) simulation results, and experimental data. The effect of geometric parameters such as rod Pitch to Diameter ratio (P/D), Rod to Wall gap (RTW) is studied in the liquid metal lead 7 rod bundles. In addition, the effect of heaving start-up and shutdown motions between liquid metal lead and water is discussed in the 5 × 5 bare rod bundles. In general, the modified subchannel analysis code can well complete the calculation of liquid metal reactor under ocean conditions, as well as provide support for the design and safety analysis of a liquid metal cooled fast reactor.

Thermal and hydraulic characteristics analysis of horizontal lead-bismuth reactor assembly based on sub-channel analysis code

In this paper, the self-developed subchannel analysis program SACOS-PB was used to analyze the thermal hydraulic characteristics of 127 rods in horizontal lead-bismuth reactor under steady state and typical ocean transient conditions. The results of the steady-state analysis showed that the coolant flow and temperature of the vertical assembly were symmetrically distributed, the coolant flow of the horizontal assembly gradually increased along the gravity direction, the temperature rose first and then fell, and the peak coolant temperature of the cross-sectional channel was about 365 °C. Three typical ocean transient operating conditions, inclined, undulating and swing, were also carried out. The results showed that the maximum temperature of the coolant at the exit of the assembly increased with the increase of the transverse heeling angle and the decrease of the trim angle after the incline motion changes the gravity component. Under the undulating and swing conditions, the coolant hydrothermal parameters fluctuated periodically, and the parameter fluctuation period is 10s in line with the oceanic conditions. In the above transient conditions, the peak coolant temperature at the assembly outlet increases and exceeds 365 °C. Compared with the pitching motion on the Y-axis, the rolling and tumbling motion on the X-axis had a significant effect on the coolant flow rate and temperature distribution. The calculation results can provide support for the analysis study of the safety characteristics of lead-bismuth power units under marine conditions.

Predicting the core thermal hydraulic parameters with a gated recurrent unit model based on the soft attention mechanism

Accurately predicting the thermal hydraulic parameters of a transient reactor core under different working conditions is the first step toward reactor safety. Mass flow rate and temperature are important parameters of core thermal hydraulics, which have often been modeled as time series prediction problems. This study aims to achieve accurate and continuous prediction of core thermal hydraulic parameters under instantaneous conditions, as well as test the feasibility of a newly constructed gated recurrent unit (GRU) model based on the soft attention mechanism for core parameter predictions. Herein, the China Experimental Fast Reactor (CEFR) is used as the research object, and CEFR 1/2 core was taken as subject to carry out continuous predictive analysis of thermal parameters under transient conditions., while the subchannel analysis code named SUBCHANFLOW is used to generate the time series of core thermal-hydraulic parameters. The GRU model is used to predict the mass flow and temperature time series of the core. The results show that compared to the adaptive radial basis function neural network, the GRU network model produces better prediction results. The average relative error for temperature is less than 0.5 % when the step size is 3, and the prediction effect is better within 15 s. The average relative error of mass flow rate is less than 5 % when the step size is 10, and the prediction effect is better in the subsequent 12 s. The GRU model not only shows a higher prediction accuracy, but also captures the trends of the dynamic time series, which is useful for maintaining reactor safety and preventing nuclear power plant accidents. Furthermore, it can provide long-term continuous predictions under transient reactor conditions, which is useful for engineering applications and improving reactor safety.

Flow characteristics of dispersed two-phase flows in an 8 × 8 rod bundle

Two-phase flows in rod bundles or tube bundles appear in heat transfer systems. The bundle-average (or one-dimensional) void fraction is a critical parameter in overall system design, performance evaluation, and safety assessment. Accurate bundle-average void fraction prediction requires a precise interfacial momentum transfer term formulated using the distribution parameter and drift velocity, two essential drift-flux parameters in the drift-flux model. The currently utilized modeling approach is based on the distribution parameter and drift velocity determined together through drift-flux plots. This results in possible compensation errors between the distribution parameter and drift velocity in the model validation process. The independent validation of the constitutive equations for the distribution parameter and drift velocity has been challenging due to experimental difficulty in measuring detailed two-phase flow structures in rod bundles. The present study proposes an approximation methodology to obtain subchannel average two-phase flow parameters using local two-phase flow data measured at two local points: the subchannel center and minimum gap center in a rod bundle. The distributions of subchannel average two-phase flow parameters are invaluable in benchmarking subchannel analysis codes. Comprehensive subchannel average void fraction and gas velocity mappings are given to discuss the effects of gas and liquid velocities, pressure, spacer grid, and developing length on the two-phase flow characteristics in the rod bundle. The bundle-average distribution parameter values calculated by subchannel average two-phase flow parameters are used for independent validation of the constitutive equations of the distribution parameter. The bundle-average drift velocity values back-calculated using the distribution parameter values are also used for independent validation of the constitutive equation of the drift velocity.

Prediction of critical heat flux in non-uniform heated rod bundle based on modified CHF empirical correlations

The accurate estimation of critical heat flux (CHF) continues to be a significant challenge in the study of radial and axial non-uniform heated rod bundles under NHR-200Ⅱ operating conditions. The present work develops the subchannel analysis code CTF algorithm that incorporates modified CHF empirical correlations to forecast CHF of non-uniform heated rod bundles preliminarily. The article selected experimental data from HTRF, which includes both uniform heated and non-uniform heated rod bundle CHF experiments, for preliminary validation and analysis of this method. The results demonstrated that the predicted results of modified Barnett correlation have the good agreement with the experimental values. Additionally, the modified Barnett correlation is used to predict the CHF of rod bundles under various power distribution. In different radial power distribution, the critical power of rod bundles decreases with the increase of radial power factor ratio. In different axial power distribution, the critical power in inlet and outlet peak distribution is lower than uniform heated rod bundle. Regarding the case of radial uniform heating and axial inlet peak distribution, CHF locations often occur near the central section of the heated rod height. In the axial intermediate peak distribution, CHF positions often emerge in the intermediate and downstream sections of the heated rod to the outlet. In a typical operating scenario of the NHR-200II, the CHF will occur towards the exit of the hot central subchannel, particularly near the radial heated rod exhibiting the maximum thermal power. The occurrence of CHF can be attributed to factors such as lower mass flux, higher local equilibrium quality, and heat flux.

An inter-channel mixing model based on the void distribution in a rod bundle

Turbulent mixing effect is crucial for coolant subchannels in a nuclear reactor, as it directly influences flow distribution and cooling capacity. The enthalpy in the hottest channel is reduced in this way, thus improving the location of critical heat flux (CHF) to some extent. In order to predict CHF accurately, the inter-channel mixing model is incorporated into the subchannel analysis code to simulate thermal-hydraulic phenomena during a reactor operation. In this paper, a novel inter-channel mixing model has been developed through extensive investigation of experimental research. In addition, the power-law profile is considered for predicting the void distribution resulting from the void drift effect. Finally, the developed mixing model is embedded in a subchannel analysis code, and pressurized water reactor sub-channel and bundle tests (PSBT) data are selected as a benchmark for the model validation. The simulation results show that the current mixing model can accurately predict void fraction and thermal equilibrium quality at specific axial locations with absolute errors of 0.1 and 0.04, respectively.

A subchannel analysis code SACOS-Na for sodium-cooled fast reactor

The strong coupling heat transfer between subassemblies and inter-wrapper gaps is closely related to the safety of the sodium-cooled fast reactor (SFR). Therefore, it is very important to carry out a detailed thermal hydraulic analysis of the core. In this paper, the inter-wrapper channel is modeled using the subchannel analysis method. A core thermal-hydraulic analysis code SACOS-Na (Subchannel Analysis Code Of Safety for SFR) is developed by using SIMPLE algorithm.For the subassembly module, a total of three experiments, including ORNL 19-pin experiment, TOSHIBA 37-pin experiment and KNS 37-pin experiment, are adopted to demonstrate the feasibility of the mathematical and physical models and numerical algorithms of the SACOS-Na as well as the rationality of the selected auxiliary models. For the inter-wrapper gap module, WEISS 2-subassembly experiment and CCTL-CFR 3-subassembly experiment are used to verify not only the simulation capability for backflow and strong inter-subassembly heat transfer, but also the heat structure model and the subchannel modeling approach on inter-wrapper gaps. In addition, the prediction accuracy of SACOS-Na has reached the advanced international level by comparing with the worldwide well-known core thermal hydraulic analysis codes.

A validation study of the SLTHEN code for hexagonal assemblies of wire-wrapped pins using liquid metal heating experiments

This paper presents a validation study of the subchannel analysis code SLTHEN used for the core thermal-hydraulic design of the Prototype Gen-IV sodium-cooled fast reactor (PGSFR). To assess the performance of the ENERGY model of SLTHEN, four liquid metal heating experiments conducted by ORNL, WARD, and KIT with hexagonal assemblies of wire-wrapped rod bundles were analyzed. These experiments were performed with 19- and 61-pin bundles and varying power distributions of axial and radial peaking factors up to 1.4 and 3.0, respectively. The coolant subchannel temperatures measured at different axial locations were compared with the SLTHEN predictions with the Novendstern, Chiu-Rohsenow-Todreas (CRT), and Cheng-Todreas (CT) correlations for flow split and mixing in wire-wrapped pin bundles. The results showed that the SLTHEN predicts the measured subchannel temperatures reasonably well with root-mean-square errors of ∼10 % and maximum errors of ∼20 %. It was also observed that the CRT and CT correlations consistently outperform the Novendstern correlation.

Parallelization and optimization for a pin-by-pin whole core thermal–hydraulic analysis code

With increasingly developing high-performance computing (HPC) technology, pin-by-pin thermal–hydraulic simulations are now feasible. The in-house subchannel analysis code KMC-FBc has been extended to perform pin-by-pin calculations. To expedite the whole-core thermal–hydraulic analysis process, various parallelization and optimization strategies are proposed and implemented at the single-assembly and multi-assembly levels, including acceleration methods for solving large sparse matrix equation systems, memory and file input/output (I/O) optimization, hybrid MPI/OpenMP-based whole-core acceleration method and so on. Parameters for evaluating the parallel performance, such as the running time, speedup, and parallel efficiency, are tested on an HPC cluster. Results show that the maximum acceleration ratio for the whole-core problem with a mesh size in the tens of millions can reach approximately 120, with a parallel efficiency of around 32%, which demonstrates the rationality and effectiveness of the parallel algorithm and optimization methods and lays the foundation for future high-resolution and high-fidelity thermal–hydraulic analysis.

Sub-channel analysis for LBE cooled fuel assembly with grid type spacers of China initiative accelerator driven system

The present study focuses on the thermal hydraulic characteristics of fuel assemblies in the China Initiative Accelerator Driven subcritical System (CiADS), which utilizes grid-type spacers to support lead bismuth eutectic (LBE) cooled fast reactors. While the structural design and mechanical analysis of CiADS triangular fuel assemblies with grid-type spacers have been preliminarily investigated, there is a pressing need to study the thermal hydraulic behavior of liquid LBE within these fuel assemblies. This analysis is crucial for assessing the safety and economy of the CiADS subcritical reactor. The sub-channel analysis code, based on the lumped parameter method, plays a vital role in rapidly evaluating the safety features of LBE-cooled fast reactor fuel assemblies. To comprehensively evaluate the thermal hydraulic performance of fuel assemblies with grid-type spacers in the CiADS LBE cooling system, the flow pressure drop model, turbulent mixing model, and convective heat transfer model for the fuel assembly structure with grid-type spacers were incorporated into the existing CiADS sub-channel analysis code. This enhanced code has been successfully employed in the thermal hydraulic analysis of fuel assemblies utilizing wire-type spacers in the CiADS system. To verify the validity and accuracy of the modified CiADS sub-channel analysis code in calculating triangular fuel assemblies with grid-type spacers, the code was first utilized to replicate a 19-rods flow heat transfer experiment with grid-type spacers cooling liquid LBE. A comparison between the simulation and experimental results confirms that the CiADS sub-channel analysis code can predict the coolant temperature and fuel rod surface temperature in fuel assemblies with grid-type spacers. Subsequently, the structure and thermal hydraulic design of the latest CiADS fuel assembly with grid-type spacers were reviewed. Finally, taking into account the aforementioned enhancements to the CiADS sub-channel analysis code, a comprehensive thermo-hydraulic calculation of the CiADS fuel assembly with grid-type spacers was conducted. The results demonstrate that the coolant outlet temperature, maximum cladding temperature, and maximum fuel pellet temperature all fall within the design parameters of CiADS. Furthermore, the impact of the number of rods in different rod bundles (91, 61, 37, and 7 rods) on thermal hydraulic performance was analyzed. The findings indicate that, while ensuring calculation accuracy and efficiency, the results obtained from the 37-rod bundle can effectively reflect the thermal hydraulic behavior of the reactor core.

Modeling of distribution parameters for upward steam-water boiling flows in subchannels of a vertical rod bundle

ABSTRACT In simulations of two-phase flow behavior in nuclear reactors, subchannel analysis codes are often used to evaluate the void fraction within a BWR fuel bundle in detail. When solving the momentum conservation equation averaged over a subchannel cross-section for upward two-phase flows, the distribution parameter is required to consider the void fraction and velocity distributions in the subchannel cross-section. In this paper, constitutive equations were developed for the distribution parameters for dispersed two-phase flows applicable to the inner, edge, and corner subchannels, which are typical subchannels in a fuel bundle. The distribution parameters could be calculated by giving the void fraction and the velocity distribution. Therefore, the distribution parameters were evaluated and modeled as a function of the geometrical parameters by assuming the void fraction and velocity distributions with bulk and subcooled boiling for each subchannel type. The developed constitutive equations were evaluated by comparing them with the distribution parameters estimated based on the NUPEC rod bundle void fraction test data. The developed distribution parameter model was implemented into the subchannel analysis code NASCA and compared with the measured cross-sectional average void fraction of the NUPEC rod bundle void fraction test data. In comparison with the original NASCA code, which assumed the distribution parameter to be unity, the improved NASCA with the distribution parameter model decreased the mean error of the measured cross-sectional average void fraction to less than half of the result of the original NASCA code, both in absolute and relative differences.

Study on the mixing performance of mixing vane grids and mixing coefficient by CFD and subchannel analysis code in a 5×5 rod bundle

Mixing Vane Grid (MVG) is one of the most important structures in fuel assembly due to its high performance in mixing the coolant and ultimately increasing Critical Heat Flux (CHF), which avoids the temperature rising suddenly of fuel rods. To evaluate the mixing performance of the MVG, a Total Diffusion Coefficient (TDC) mixing coefficient is defined in the subchannel analysis code. Conventionally, the TDC of the spacer grid is obtained from the combination of experiments and subchannel analysis. However, the processing of obtaining and determine a reasonable TDC is much challenging, it is affected by boundary conditions and MVG geometries. In is difficult to perform all the large and costing rod bundle tests. In this paper, the CFD method was applied in TDC analysis. A typical 5 × 5 MVG was simulated and validated to estimate the mixing performance of the MVG. The subchannel code was used to calculate the TDC. Firstly, the CFD method was validated from the aspect of pressure drop and lateral temperature distribution in the subchannels. Then the effect of boundary conditions including the inlet temperature, inlet velocities, heat flux ratio between hot and cold rods and the arrangement of hot and cold rods on MVG mixing and TDC were studied. The geometric effects on mixing are also carried out in this paper. The effect of vane pattern on mixing was investigated to determine which one is the best to represent the grid's mixing performance.

Subchannel thermal hydraulic analysis of 5 × 5 rod bundle with CRUD layer

This paper studies the thermal–hydraulic behaviors difference with or without the Chalk River Unidentified Deposits (CRUD) layer by using subchannel analysis code COBRA-EN. Based on similar structure characteristics, the EPRI CHF prediction model is improved and applied in a 5 × 5 rod bundle. Three different species transient scenarios are calculated for comparing different hydraulic behaviors with or without the CRUD layer. The results show that the heat transfer enhancement is limited under high pressure because the improved CHF model is mainly affected by local pressure. But when the pressure came low, the enhancement became more obvious, such as heat flux and minimum departure from nucleate boiling ratio (MDNBR). Under serious accidents, the influence can be mitigated with the CRUD layer, and recovery from the accident with the CRUD layer was faster than bare cladding when considering the effect of the CRUD layer.

Modeling of void fraction covariance and relative velocity covariance for upward boiling flows in subchannels of a vertical rod bundle

Accurate simulation of boiling two-phase flows in a rod bundle is indispensable for the robust, economical, and safe design of various heat transfer systems using the rod bundle configuration. Subchannel analysis codes are used for this purpose. Interfacial drag force modeling significantly affects the prediction accuracy of the void fraction. The void fraction and relative velocity covariances constitute the interfacial drag force. However, the covariances are currently not considered in existing subchannel codes due to a lack of reliable constitutive equations to calculate the void fraction and relative velocity covariances. This study aims to model the subchannel-average void fraction and relative velocity covariances for subcooled and saturated boiling flows in three types of subchannels in a rod bundle. The considered subchannels are interior, edge, and corner subchannels. The subchannel-average void fraction and relative velocity covariances for saturated boiling flow are modeled by the data obtained from local void fraction data collected for saturated boiling flow in an 8 × 8 rod bundle under pressures from 1.0 to 8.6 MPa. The subchannel-average void fraction and relative velocity covariances for subcooled boiling flow are modeled based on the bubble-layer thickness model. The modeled subchannel-average void fraction and relative velocity covariances are well validated with the experimental data. The modeled subchannel-average void fraction and relative velocity covariances are expected to be implemented in subchannel analysis codes to improve the void fraction prediction accuracy in each subchannel type.

Investigation on Applicability of Subchannel Analysis Code ASFRE to Thermal Hydraulics Analysis in Fuel Assembly With Inner Duct Structure of Sodium Cooled Fast Reactor

Abstract In the design study of an advanced sodium-cooled fast reactor (advanced-SFR) in Japan Atomic Energy Agency (JAEA), the use of a specific fuel assembly (FA) with an inner duct structure called fuel assembly with an inner duct structure (FAIDUS) has been investigated to enhance safety of Advanced-SFR. Due to the asymmetric layout of fuel rods by the inner duct, it is necessary to estimate the coolant temperature distribution to confirm feasibility of FAIDUS. In JAEA, an in-house subchannel analysis code named thermal-hydraulic analysis of asymmetrical flow in reactor elements (ASFRE) has been developed as a FA design tool. For the typical FAs, the numerical results of ASFRE had been validated by comparisons with experimental data, in the previous study. As for the FAIDUS, confirmation of validity of the numerical results by ASFRE was not enough because the reference data on the thermal hydraulics in FAIDUS have not been obtained by the mockup experiment, yet. In this paper, therefore, the code-to-code comparisons with numerical results of ASFRE and those of an in-house computational fluid dynamics (CFD) code named single-phase thermal-hydraulic analysis in an irregular rod array layout (SPIRAL) were applied to make further discussion on applicability of ASFRE to the thermal hydraulics analysis in FAIDUS. Thermal hydraulic analyses of a typical FA and FAIDUS at high and low flowrate conditions were conducted. The applicability of ASFRE was indicated through the confirmation of the consistency of mechanism on appearance of the specific temperature distributions between the numerical results by ASFRE and those by SPIRAL. In addition, the necessity of modification on the empirical constants in numerical model of ASFRE to improve the predictive accuracy was indicated.