Heated Rod Bundle Research Articles

Toward a Better Understanding of Reflood Thermal Hydraulics: A Summary of the OECD/NEA RBHT Project

Reflood thermal hydraulics remains a difficult and complex subject, and understanding the physical phenomena that occur during a reflood transient is important to nuclear safety. The Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) Rod Bundle Heat Transfer (RBHT) project was designed to provide unique experimental data for code assessment and model development. Participants, which came from 21 international organizations, used analysis codes including APROS, ATHLET, CATHARE, CTF, MARS, RELAP5, TRACE, and SPACE to simulate the tests performed in the RBHT facility. The experimental campaign carried out within the OECD/NEA RBHT project produced data for a total of 16 reflood tests conducted in two test series. An “open” test series consisted of 11 experiments, and a “blind” test series consisted of 5 experiments. In the blind tests, only the initial and boundary conditions were provided to participants prior to simulation of those experiments. Reflood rates ranged from 0.5 to 15 cm/s, thus producing data applicable to dispersed flow film boiling and inverted annular flow film boiling. Inlet subcooling ranged from 2.8 to 80 K. Tests with variable reflood rates and oscillatory reflood rates were included in the test matrix. This paper describes the project and presents a summary of major experimental and analytical findings.

Development of a new criterion for predicting critical heat flux in rod bundles with supercritical water flowing upward

Investigations of heat transfer characteristics in channels with supercritical fluids are crucial for power cycle applications, such as supercritical water-cooled reactors. To ensure the safety of the system, heat transfer deterioration (HTD) should be avoided. The criteria to identify the onset of HTD in heated rod bundles are still not determined. This study summarizes indicators for the onset of HTD in a rod bundle with supercritical water flowing upward based on observations from experimental data. Existing criteria for the tube flow are assessed by comparing predicted critical heat flux (CHF) values with experimentally identified values. Considering effects of different operating parameters, a new criterion for predicting HTD behavior in rod bundles is developed. The results indicate that the new criterion has a better performance than all existing criteria, with a mean average relative deviation of 0.26 %. Predicted CHF values are within ± 20 % of the respective experimentally identified CHF values.

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.

The influence of droplet breakup model on the prediction of reactor core parameters during reflood conditions

During reflood conditions, spacer grids affect the complex flow dynamics inside the reactor core. They can breakup the dispersed droplets, thereby significantly reducing their diameters. This process enhances local heat and mass interfacial transfer, consequently reducing the peak cladding temperature (PCT). Therefore, accurately predicting droplet transport and size within a reactor core is very important for reactor safety analysis. Recently, the US. NRC TRACE code has been modified by extending its governing equations to incorporate a three-field approach that captures the dynamic behavior of droplet entrainment, transport, and their interactions with structures within the reactor core. The new TRACE (v5.0 patch 8) three-field framework incorporates droplet models accounting for entrainment from liquid pools, liquid films on vertical surfaces, and droplet breakup. This paper focuses on the influence of the spacer grid droplet breakup model on the behavior of reactor core during reflood transient. The selected droplet breakup model effectively captures the enhancement in interfacial heat transfer observed in experimental data. We evaluate the influence of the droplet breakup model on the three-field version of TRACE by comparing it to recent Rod Bundle Heat Transfer (RBHT) experiments. These experiments had previously indicated that TRACE tended to overpredict the PCT for various initial and boundary conditions. Subsequently, we conducted a detailed comparison between TRACE v5.0 patch 7 and the three-field TRACE (v5.0 patch 8) with and without the spacer grid droplet breakup model. Our findings demonstrate a significant improvement in predictive accuracy with the new breakup model compared to the base capability of TRACE v5.0 patch 7. This improvement results in more accurate predictions of PCTs and quenching times for the assessed conditions.

Adaptation of the turbulence model to the heat transfer in a rod bundle

Fast neutron reactors using liquid metal or gas coolants are one of the most actively studied fields of research. Increased safety requirements and insufficient knowledge of the thermohydraulic characteristics of these coolants require validation of RANS simulations via information from experiments or DNS. This information can also be applied to improve the RANS turbulence model ifnecessary.In this paper, a series of thermohydraulic DNS simulations of the uniformly heated rod bundle with a relative pitch of 1.4 have been performed to improve RANS. A total of 12 test cases with four different Prandtl numbers of Pr = 0.0043, 0.0154, 0.0324 and 0.71, which correspond to liquid metals and gas, and three Reynolds numbers of Re = 2500, 5000 and 10,000 are considered. The present simulation results show good agreement with the experimental and DNS data. It is revealed that the intermittent flow regime forms at Re = 2500. Furthermore, the turbulent Prandtl number is structured with a maximum of up to y+≈25 and, from y+≈50 onwards, in most cases takes on a constant value. By increasing the Prandtl or Reynolds number, the average value of the turbulent Prandtl number tends to be 1.0. It was also revealed that the turbulent contribution to heat transfer is linear to the Reynolds number for the selected coolant. A new formula for the temperature profile was also obtained, which shows better agreement with DNS data compared to its analogues.As a result, the selected RANS model has been improved (the maximum integral error dropped from 49% to 4%) by the new model of the turbulent Prandtl number and by introducing model coefficients from DNS.

Experimental investigation and correlation development for liquid carryover during reflood

A correlation is developed for the liquid carryover fraction during both constant and oscillatory reflood with application to nuclear power accidents. The correlation, semi-empirical in nature, is based on extensive experiments in a 7 × 7 rod bundle array at the Nuclear Regulatory Commission/Pennsylvania State University Rod Bundle Heat Transfer (NRC/PSU RBHT) facility. Liquid carryover and entrainment can significantly impact quench behavior and maximum cladding temperatures during a postulated accident; therefore, accurate predictions of these phenomena are critical. The parameters associated with the stability at the liquid–vapor interface, and hence the liquid entrainment and subsequent carryover are established. These parameters provide the basic form of the correlation with experimental data from the NRC/PSU RBHT facility to determine the model coefficients. Using a weighted least-squares method that minimizes the error between modeled and experimental carryover fraction, we implemented a genetic algorithm to identify the parameters of this correlation. The proposed model is compared with experimental data and predictions from the NRC’s TRACE code and lies within 20% error margin.

Measurements of Droplet Size and Velocity Distributions During Rod Bundle Core Reflood

As part of the Organisation for Economic Co-operation and Development/Nuclear Energy Agency Rod Bundle Heat Transfer (RBHT) project, an experimental study was performed to investigate the entrained droplet sizes and velocities in a rod bundle under reflood conditions. Experimental results were obtained from the U.S. Nuclear Regulatory Commission/The Pennsylvania State University RBHT test facility using advanced dual laser measurement systems that allow for the simultaneous measurement of droplet behaviors at two axial locations during reflood transients. The RBHT facility is highly instrumented and contains a 7×7 electrically heated bundle with dimensions matching those in commercial pressurized water reactors. The combination of the measurement capabilities of the RBHT facility and the choice of appropriate experimental conditions allows for the measurement of unique droplet size and velocity distributions under different transient reflood conditions.

Spatio-temporal characteristics of void fraction in heated rod bundle under saturated pool boiling due to thermal power oscillation

Disruption in core coolant injection leads to pool boiling in a reactor. The void behavior during this pool boiling significantly affects long-term core cooling. Thus, a pool boiling experiment was performed using a 5 × 5 heated rod bundle with steady-state and sinusoidal oscillating thermal power at 0.05 and 0.10 Hz for a pressure range of 0.5–7.2 MPa. Subchannel void sensors were installed in the 5 × 5 rod bundle. They partially simulated the fuel assembly of an actual reactor core to analyze the spatio-temporal variations and frequency characteristics of the temporal evolution of void fraction distribution. Analysis of experimental results with wavelet transforms revealed the two-phase flow structure due to thermal power oscillation and its sensitivity to pressure and frequency of thermal power oscillation.

Experimental study of bottom reflooding for 5×5 non-uniformly heated rod bundle with spacer grid

Water heat transfer effectiveness during reflooding process in rod bundle geometry is very important for reactor safety under loss of coolant accident. A series of reflooding experiment is carried out at the REFB_NUSOL test facility for a 5 × 5 rod bundle to study the effect of several key parameters on overall reflooding process, including subcooling of coolant (20 °C to 80 °C), coolant flow velocity (2.36 cm/s to 11.19 cm/s), initial peak surface temperature (500 °C–900 °C) and steam temperature on quench front progress. The temperature, heat flux and heat transfer coefficient on the heated rod surfaces have been achieved via a self-developed Reflooding Experimental Data Analysis (REDA) code. Studies on surface temperature variation, quench front progress, heat transfer enhancement caused by spacer grid and effect rules of minimum film boiling temperature are discussed to shed a light on the thermal-hydraulic behaviors during reflooding process in different conditions for rod bundles. Additionally, the prediction accuracy of three published empirical correlations on minimum film boiling temperature are evaluated using our experiment data. The meaning of this work is to explain the reflooding phenomenon of 5 × 5 rod bundle meanwhile provide validation data and suggestions for empirical equation development for nuclear thermal hydraulic analysis code to better predict reflooding process.

Validation and Uncertainty Quantification of Transient Reflood Models Using COBRA-TF and Machine Learning Techniques Based on the NRC/PSU RBHT Benchmark

The accurate prediction of the fluid flow mass and the heat transfer process as well as the system response during reflood transients has long been a critical and challenging issue for reactor system safety analyses. Accurate characterization of the flow and energy transport can also significantly facilitate the various system/component design and optimization tasks. In the current study based on the U.S. Nuclear Regulatory Commission/Pennsylvania State University Rod Bundle Heat Transfer (RBHT) reflood experimental data, a comprehensive uncertainty analysis framework is developed using DAKOTA. The developed framework is used to perform an in-depth reflood model validation and verification for the subchannel analysis code COBRA-TF. In the meantime, the artificial intelligence (AI)–based machine learning (ML) model for rod cladding temperature prediction during reflood is also developed and evaluated using the current framework. Key input parametric effects for reflood thermal-hydraulic prediction include the system pressure, inlet liquid temperature/enthalpy, inlet mass flow rate, and average bundle power input. The figure of merit under consideration is the peak cladding temperature variations. It is found in the current study that, while further model improvement is needed, COBRA-TF can predict the correct parametric trends when compared with the RBHT data. On the other hand, it is challenging for the pure AI-based ML models to correctly reflect the parametric trends. Suggestions for future ML model development are provided in the end.

Computational Fluid Dynamics Modeling of a Pressurized Water Reactor Fuel Assembly to Estimate Loss Coefficients in Support of Subchannel Thermalhydraulics Modeling of Pressurized Water Reactor Small Modular Reactors With Advanced Fuels

Abstract Pressure loss coefficients are generally required by subchannel and system thermalhydraulics codes. These coefficients are not readily available for small modular reactors (SMRs) featuring nonconventional designs and novel coolants. In this study, the pressure loss coefficients were obtained using three-dimensional (3D) computational fluid dynamics (CFD) modeling for an advanced water-cooled reactor. A representative light water fuel assembly used in the Organization for Economic Co-operation and Development (OECD)/National Research Council Canada (NRC) pressurized water reactor subchannel and bundle tests (PSBT) benchmark was selected for CFD modeling and simulation under various working conditions. The fuel assembly includes three types of pressurized water reactor (PWR) spacer grids: simple grid (SG), nonmixing vane grid (NMVG), and mixing vane grid (MVG). Turbulent flow through subchannels of both nonheated and heated rod bundles was simulated to predict recoverable and nonrecoverable pressure distribution along the length of the bundle. It was observed that vortices were generated at the tips of spacer grids, affecting the cross-flow in subchannels significantly. The estimated pressure loss coefficients were found to be influenced by the flow conditions (Reynolds number or the upstream flow history) and spacer grid configuration. Pressure loss coefficient values ranged from 1.14 to 1.80, depending on the spacer grid type, design, and flow conditions. The CFD method used in this study was demonstrated to have the potential to generate input parameters required for the subchannel analysis and optimization of fuel assembly designs and serve as a surrogate for empirical correlations.

Experimental and Numerical Studies of Pressure and Oscillatory Flow Effects on the Two-Phase Flow and Heat Transfer in a Rod Bundle

An experimental and numerical study was performed to investigate the effects of pressure and constant vs oscillatory flooding rates on the two-phase flow and heat transfer behavior of a rod bundle under reflood transient conditions. Experimental results were obtained from the NRC/PSU Rod Bundle Heat Transfer (RBHT) test facility from various test cases covering a range of system pressures with light water as the working coolant. For each pressure case, two experiments were performed, one for a constant flow rate, and one for an oscillating flow rate about the constant flow rate. The RBHT test facility, which contains 49 vertical, 3.66 m (12 ft) long test rods (four unheated corner rods and 45 heated rods) with Inconel 600 cladding in a 7 × 7 geometry, having the rod diameters, rod pitches and spacer grids comparable to those in commercial PWRs, was specifically designed to obtain fundamental flow and heat transfer data during reflood transients. The thermal–hydraulic code TRAC/RELAP Advanced Computational Engine (TRACE) was used in this study by performing simulations with the same geometry and operating conditions as the RBHT facility for each experiment. Results of the TRACE simulations were compared to the experimental data obtained in the RBHT tests. It was found that the trends on the pressure effects for constant and oscilatory flows on the thermal–hydraulic behavior of the rod bundle (i.e., the two-phase flow and heat transfer behavior of the rod bundle during reflood transients) predicted by the TRACE model agree well with the RBHT data. This comparison of results has also assisted in other studies to investigate numerical discrepancies currently underway.

Development of Subchannel Void Sensor for Wide Pressure and Temperature Ranges and Its Application to Boiling Flow Dynamics in a Heated Rod Bundle

A subchannel void sensor (SCVS) acquires the two-phase flow in a rod bundle as the time-series data of cross-sectional distributions. Herein, the temperature and pressure ranges of an SCVS were extended to include the rated conditions of boiling water reactors. The improved SCVSs were installed in a 5 × 5 heated rod bundle at eight height levels. In a boiling experiment using the rod bundle, the three-dimensional distributions of the boiling two-phase flow were measured over a wide pressure range (up to 7.2 MPa). The new experimental data were compared with existing experimental data and the results of a subchannel analysis. Experimental results were consistent with those of a high-energy X-ray computed tomography study of a heated rod bundle with the same geometry and under the same heat and flow conditions as those used in our study. The subchannel analysis code reproduced the experimental results fairly well, and the obtained database is applicable for validating and improving thermal-hydraulic analysis codes.

Development of a new spacer grid pressure drop model in rod bundle for the post-dryout two-phase flow regime during reflood transients

Flow blockages such as the spacer grids play an important role in a two-phase flow system as they modify the flow and heat transfer behavior. Their existence, on the one hand, increase the turbulent mixing and heat transfer capability of fluid directly downstream; while on the other hand, create additional pressure loss within the system thus requiring a detailed and comprehensive understanding. In the present work, a new two-phase flow pressure drop correlation induced by spacer grids in the dispersed flow film boiling (DFFB) regime is developed based on the unique reflood experimental data obtained from the Nuclear Regulatory Commission/Pennsylvania State University rod bundle heat transfer test facility. The test conditions cover wide ranges of system parameters for two-phase flow system which can be used for model development. It is found that the most dominating system parameter is the mixture mass flow rate across the spacer grid location. As a result, the spacer grid loss coefficient is correlated with the mixture Reynolds number and a blockage geometry correction factor. The current correlation is found to significantly improve the prediction of the spacer grid pressure drop in the DFFB regime generally within a 25% error span, while the other exiting correlations failed to capture the two-phase flow hydraulic behavior involved during the reflood transients. The current correlation developed can be readily implemented into existing thermal-hydraulic analysis codes to further improve their prediction capabilities in the DFFB regime with the presence of flow blockages.

Numerical investigation of rod bundle thermal–hydraulic behavior during reflood transients using COBRA-TF

Accurate prediction of the reflood transients during LOCA has long been a challenging task. One reason for this is the substantial cost incurred to perform large-scale reflood tests, which are formidable in experimental expenditure and intensive in technology as well as program management. Another reason preventing a comprehensive understanding of reflood transients is the high-level complexity involved in the two-phase flow mass and heat transfer processes, which makes the measurement extremely difficult and inefficient. The situation has been significantly improved with the design and operation of the Nuclear Regulatory Commission (NRC)/Pennsylvania State University (PSU) Rod Bundle Heat Transfer (RBHT) test facility. A variety of advanced instrumentations were developed and used at this facility and very high-resolution data has been obtained, especially for the liquid droplet field. In the current study taking advantage of the unique NRC/PSU RBHT data, an extensive and comprehensive code evaluation and validation is carried out using the thermal–hydraulic sub-channel analysis code COBRA-TF. The system parametric effects investigated include: the system pressure, inlet liquid subcooling temperature, inlet flooding rate and rod bundle power. A variety of thermal–hydraulic quantities predicted by the numerical code are evaluated including: the quench front propagation, cladding, spacer grid and vapor temperature variations, two-phase pressure drop, liquid droplet velocity and coolant void fraction. In addition, the prediction errors are presented for each of the quantities in great detail in order to have a comprehensive understanding of the code performance. In general, COBRA-TF agrees relatively well with the experimental data in terms of the overall quench front propagation, only predicting slightly earlier quench. However, under low inlet subcooling and high flooding rate conditions, significant discrepancies are observed. The comparison with rod bundle thermal–hydraulic parameters indicates that COBRA-TF is able to predict the cladding temperature before quench well within a 15% range. The spacer grid and vapor temperature predictions are within 20% error. While the overall droplet velocity prediction is found to be within 30% error, COBRA-TF is able to capture the decreasing trend for droplet velocity during reflood. In addition, the prediction of the dispersed flow film boiling void fraction shows that the code always under-predicts the void fraction. While the most prediction discrepancy is found to be the two-phase flow pressure drop after quench, which involves more than 50% under-prediction for the bulk liquid boiling regime. The present study clarifies the effect of different system parameters on the various two-phase flow quantities during reflood. The results obtained provide answers for many existing code modeling issues and thus can be instructive and useful for future model development and code upgrading.

Fully developed turbulent convection of Lead Bismuth Eutectic in the elementary cell of the NACIE-UP Fuel Pin Bundle

One of the requirements for achieving high levels of safety in fourth generation nuclear power plants, is that core thermal-hydraulics can be simulated numerically at a good level of accuracy. To this aim, detailed validation of turbulent heat transfer models needs to be carried out and discussed, in relevant flow and heat transfer configurations, and for low Prandtl-number fluids.The focus of this research is the turbulent, fully developed convection in a heated rod bundle with a triangular arrangement, and a pitch-to-diameter ratio P/D=1.4. Statistics derived from a set of Direct Numerical Simulations (DNS) at the moderate Reynolds number of Re=8290 are compared to Reynolds Averaged Navier-Stokes (RANS) solutions where the closure is provided by the two-equation model SST k-ω and by the Reynolds-Stress Model. Comparisons are drawn for forced flow and in mixed convection conditions (Ri≈0.25). Global quantities extracted from experiments performed in very similar conditions are also compared against the numerical results. Profiles of the turbulent Prandtl number about the unit flow cell are also displayed and discussed.This work clarifies through detailed comparison that if, on the one hand, only the Reynolds-Stress model is able to reasonably capture important flow features like secondary flow components, on the other hand also the SST k-ω two-equation model considered is acceptably accurate in predicting the integral quantities of interest.

Void-fraction measurement with high spatial resolution in a 5×5 rod bundle by linear-accelerator-driven X-ray computed tomography over a wide pressure range

Void fraction (i.e., the volume fraction occupied by gas) is a key parameter for determining the coolability and neutron-moderating performance of a water-cooled nuclear reactor. To develop computational multi-fluid dynamics models for determining the void-fraction distribution, experimental data of comparable quality are required. We have developed a high-energy X-ray computed tomography (CT) system to acquire three-dimensional void-fraction distributions. The CT system comprises a linear-accelerator-driven high-energy X-ray source and a linear detector array. We quantified a boiling two-phase flow in a 5 × 5 heated rod bundle at high pressure, simulating a fuel-rod bundle in a boiling water reactor (BWR). Because the axial travel of the CT system is 4 m and includes the entire BWR fuel-rod bundle, we optimized the CT imaging conditions and reconstruction method for rod-bundle visualization to reduce uncertainties due to density fluctuations in the boiling flow and imaging artifacts. We conducted a boiling experiment at a low flow rate and low thermal power and acquired three-dimensional distributions of the void fraction over a wide pressure range of 0.1–7.2 MPa. The experiment provided three-dimensional void-fraction distributions with high spatial resolution, especially in subchannel regions surrounded by rods, and the results are suitable for validating three-dimensional thermal-hydraulic analysis codes.

Development of a droplet breakup model for dry spacer grid in the dispersed flow film boiling regime during reflood transients

The liquid droplet breakup process at the leading edge of a dry spacer grid during reflood transients is studied theoretically based on the experimental observations in the present work. Various rod bundle quenching experiments were carried out at the Rod Bundle Heat Transfer (RBHT) test facility to simulate the reflood transients of a Loss of Coolant Accident (LOCA) for Light Water Reactors. In the experiment, variation of the droplet size distributions across a spacer grid are measured and analyzed. The size of droplet generated downstream of a spacer grid is found to be affected by the condition of the spacer grid as well as the quench front propagation. In addition, the substantially reduced droplet size at downstream of the spacer grid indicates that droplet breakup takes place. Based on available experimental data obtained in the Rod Bundle Heat Transfer tests and in literature, a theoretical model incorporating two droplet groups (large and small droplet groups) is developed for the dry spacer grid droplet breakup by considering the conservation of liquid mass, energy transport, and including various important flow and heat transfer parameters including the Nusselt number, Reynolds number, Jakob number, Weber number, dimensionless radiation number, distance from the initial breakup point, incoming droplet size, and spacer grid blockage ratio. Moreover, a liquid droplet mass loss coefficient was proposed for small droplet group to specially account for the liquid mass evaporation due to interfacial heat transfer and radiation. The model so developed is found to be able to predict the RBHT data within 17% error with RMSE of 0.1060 and standard deviation of 0.0990 for the validation data set, showing significantly increased accuracy. Comparison made with other experimental results indicates that the current model developed correctly captures the mass and energy transport process during breakup. In addition, sensitivity analyses are performed to investigate the overall performance and relative importance of various parameters that are related to the breakup process in the current study. The current model developed can be incorporated into nuclear reactor thermal-hydraulic and safety analysis codes to enhance their capabilities of predicting the droplet behaviors and the fuel rod peak cladding temperature in the dispersed flow film boiling regime during reflood transients of postulated accident scenarios such as LOCA for light water reactors.

On the use of LES-based turbulent thermal-stress models for rod bundle simulations

An alternative methodology is proposed here to overcome the excessive cost of large eddy simulations (LES) of full-length heated rod bundle calculations, while improving the inaccurate results typically obtained with Reynolds-averaged Navier–Stokes equations (RANS). While the cost of the full-length LES is generally too high, LES of a small section of a single rod is usually affordable. The idea presented here consists of using the information granted by the small LES calculation to determine the appropriate turbulence viscosity or turbulent thermal diffusivity that can be used to solve only for the temperature field in a pseudo-RANS approach. The study has been performed with single-rod simulations with a P/D of 1.12 and 1.24, considering rod lengths that are representative of reactor applications, for the cases of uniform heat flux and a more realistic cosine-like axial heat distribution. The spectral element code Nek5000 has been used for all LES, RANS, and pseudo-RANS simulations. The recently proposed Nek5000 steady-state solver has been used for solving the temperature field in the pseudo-RANS approach and has proved significantly faster than transient schemes. Prediction of thermal quantities is compared with classical linear and nonlinear RANS models. LES for the full-length rods has also been performed and is used as a reference. Results of the proposed method show significant improvements with respect to those obtained with RANS.

Measurement uncertainty and quenching phenomena in uniform heating rod bundle CHF tests

The uniform heated rod bundle Critical Heating Flux (CHF) tests have been often conducted as a part of mixing vane performance evaluation processes. Presently, several commercial CHF correlations were developed based on a large amount of rod bundle CHF data using axial uniformly heated rod bundles. With a uniformly heated test section, majority of CHF events are dominated by dry-out mechanism under high quality, high void fraction local conditions. Considering the inverse proportional relationship between the quality and CHF, for an axially uniformly heated test section, the CHF events should occur near the end of heating length (EOHL) where the void fraction/quality is the highest. However, the upstream shift of CHF location is often observed in the uniform heating tests, sometime as much as 250 mm or higher upstream from the EOHL.In the past, without any sufficient convincing proof, long dryout areas as well as “unknown” “mysteries” mixing effects from the mixing vane grid were often considered as the contributing factors for this type of upstream shift. Based on further and detailed examination, this paper offer analysis and evidence to demonstrate another most plausible explanation, exit quenching effect, which ultimately concludes that rod bundle CHF using uniform heated rods could have data with huge measurement uncertainty and as much as over 30% of non-conservatism due to heat lose and quenching effects at the exit of the heated length. The quenching effect at the exit of heater length is mainly due to reflood phenomena generated between the high void/low flow test section exit region and the relatively cold single phase coolant immediately above in the top plenum resulting from either excessive heat loss or operating transient condition.In this paper, system code and CFD modeling were utilized to simulate the quenching phenomena in both steady state and transient conditions. The significant void fraction drop near the exit of the test section due to quenching were clearly observed in these simulations and also evidenced by the various degrees of coloration of the heaters surface in the experiment. This exit reflood quenching phenomena may change the CHF location, suppress boiling, cool down the heated surface, and eventually lead to non-conservative high CHF values. Based on this analysis and experimental observation, it is concluded that the CHF correlations developed based on these type of data are potentially non-conservative in various degrees pending on local thermal-hydraulic conditions, which should be taken into consideration for their applications, especially for safety analysis in nuclear reactor.