Axial Power Shape Research Articles

Experimental study on coupled quenching behavior in a dual-cooled annular rod channel

In this paper, a bottom reflooding experiment was performed in a single annular rod channel to investigate the coupled quench characteristics of the dual-cooled annular fuel. An indirect-heated annular rod with the axial uniform power shape was used. The asymmetric quench front propagation between the external and internal sides of the annular rod was observed. The internal cladding was quenched earlier than the external cladding, owing to heat accumulation in cylindrical shell and the blockage of the vapor flow by the grid spacers in external channel. A unique “secondary quenching” phenomenon was observed at the internal side of annular rod due to the radial heat conduction. The effects of reflooding velocity, inlet subcooling, initial cladding temperature and heating power on quenching behavior were also studied. The present study can contribute to the understanding of the coupled quenching characteristics of dual-cooled annular fuel and obtains valid data for the development of reflooding models and the validation of thermal hydraulics codes of annular fuel.

On the Role of Axial Neutron Flux Mode Excitation in BWR Stability and Nonlinear Oscillations

Neutron flux modal decomposition is a key tool for analytical and reduced order modeling of boiling water reactor (BWR) stability and oscillations. As a minimum, the fundamental flux mode is used for representing global oscillations while the addition of at least one azimuthal harmonic is needed for simulating the regional out-of-phase mode. Unlike the fundamental and first azimuthal modes, the excitation of an axial flux mode alters the axial power shape but not the total power in the channel and therefore cannot be self-sustained when coupled to density wave–generated reactivity, presumably explaining why it has not been explicitly included in previously published models. Although not self-sustained, the axial mode excitation driven by density wave propagation and interactions with other spatial modes play important roles in interpreting observed BWR stability and oscillations particularly in the nonlinear regime when the oscillation magnitude is large. In this paper, the characteristics of the steady-state axial modes are presented, and their impact on oscillation dynamics for small and large amplitudes of both the global and the regional oscillations is studied using reduced order analytical tools. Aside from the oscillating component, our research results identify an average nonzero axial mode component to develop during limit cycle oscillations that causes the average axial power profile to shift toward the bottom of the core and thus contributes a negative reactivity component. The emergence of this nonzero average axial mode component and the associated negative reactivity were found to diminish the power increase due to global mode power oscillations and contribute to nonlinear stabilization of regional oscillations. The physical interpretation of nonlinear power oscillations with the inclusion of the axial mode component resolves previously unexplained results obtained from high-fidelity numerical models.

Recurrent neural network-based prediction of critical heat flux in rod bundles with non-uniform axial power shape

The Critical Heat Flux (CHF) prediction is one of the most important phenomena for the safety analyses of nuclear reactors. Current state-of-the-art CHF correlations for PWR fuel assemblies are based on classical regression approaches. In this paper a new method based on Recurrent Neural Networks (RNNs) for the calculation of the CHF in fuel assemblies is proposed. The sequential nature of the RNN allows it to learn the influence of the geometrical features of the fuel assembly and of the history effect by calculating the CHF not only based on the local flow conditions at the CHF location, but also upstream from it. The model was tested on two rod bundle datasets. The results show that the neural network model can reasonably well predict both the value of the CHF and its location without the need of any additional correction factors.

DEVELOPMENT OF MPACT FOR FULL-CORE SIMULATIONS OF MAGNOX GAS-COOLED NUCLEAR REACTORS

The MPACT code, jointly developed by Oak Ridge National Laboratory and University of Michigan, is designed to perform high-fidelity light water reactor (LWR) analysis using wholecore pin-resolved neutron transport calculations on modern parallel-computing hardware. MPACT uses the subgroup method for resonance self-shielding, while the primary neutron transport solver uses a 2D/1D method that is based on the method of characteristics (MoC) for the x-y planes coupled with a 1D diffusion or transport solver in the axial dimension. Additional geometry capabilities are currently being developed in MPACT to support hexagonal-pitched lattices, as well as interstitial geometry (i.e., control rods at the corner of four adjacent pin cells). In this research, the MPACT method is tested on gas-cooled reactors by applying MPACT to full-core MAGNOX reactor test problems. MAGNOX test problems were chosen due to the availability of high-quality reactor design and validation data (available through an ongoing collaboration with the National Nuclear Laboratory in the United Kingdom) and the existence of a relatively complex axial power shape that is expected to challenge the MPACT method. MPACT’s convergence for partial- and full-core problems will be tested and verified. MPACT will be compared with high-fidelity continuous-energy Monte Carlo simulations to verify core reactivity, power distributions, and performance of the available cross section data libraries and energy group structures.

Neutronic analysis of gadolinia bearing fuel assemblies

Abstract Gadolinium is mixed with nuclear fuel for extended burnup, increased reactor cycle Lengths, and in-out core refueling schemes. The purpose of this work is to investigate the effect of Gadolinium on neutronic parameters. A comprehensive analysis of different fuel assemblies bearing gadolinia with different concentration and different number of gadolinia rods was carried out. The EPR fuel assemblies were chosen in this study. The fuel assemblies are of seven different types, each assembly has unique design characteristics in terms of enrichment, number of gadolinia rods and gadolinium concentration. The calculations were achieved using MCNP6 code and with ENDF/B-VII library. A fine mesh tally was superimposed on the geometry to illustrate flux distribution in the whole assembly and its change with burnup. The variation of reactivity and isotopic composition for each region of the assembly with burnup and its influence on flux distribution was studied. The results showed that the presence of gadolinium greatly affects the power distribution. It causes some skewness in axial power shape towards the top of the assembly, the assembly which has largest number of gadolinia rods and the lowest enrichment has the largest skewness of power distribution at BOC. At EOC the flux and axial power have a double hump shape due to the buildup of gadolinium isotopes and the large decrease in reactivity with burnup.

CFD Modeling Development for DNB Prediction of Rod Bundle with Mixing Vanes Under PWR Conditions

Critical heat flux (CHF) is a primary parameter for nuclear fuel design and plant operation safety. CHF values are normally obtained from fuel bundle integral departure from nucleate boiling (DNB) or dryout experiments. These experiments are expensive, and detailed measurements (bubble dynamics, void fraction distribution, etc.) are difficult to obtain, particularly under typical pressurized water reactor (PWR) conditions of high pressure and temperature. Therefore, it is highly desirable that computational tools such as computational fluid dynamics (CFD) provide detailed flow and heat transfer information that will efficiently facilitate design improvements of PWR fuel designs.For the CFD studies described in this paper, an Eulerian-Eulerian two-phase modeling approach was adopted to predict DNB in a fuel assembly with mixing vane grids. Subcooled flow boiling was simulated using heat flux partition modeling and phase interactions. Direct addition of heat to the vapor was activated when the local vapor volume fraction reached a specified critical value. Emphases were placed on bubble departure diameter, phase interactions, and pressure drop for two-phase modeling development. Simulations were conducted in steady state. Solution convergence was closely monitored for physical variables in terms of local and global scales. A multi-indicator approach was used to judge DNB occurrence, and a new integrated DNB indicator is proposed.For validation, this CFD-based DNB modeling methodology was applied to two 5 × 5 rod bundle tests equipped with mixing vane grids and uniform axial power shape. The tests were performed under PWR conditions (16.5 MPa) and produced an exit quality of −4% and 11%. The CFD results show the validity of the new DNB indicator and the improved reliability with the multi-indicator approach. Correctly predicted DNB occurrence locations show the promise of the current modeling approach. Utilization of measured pressure drop and fluid temperature data may permit some bottom-up validations, and this effort may prompt further improvements for experimental measurements, particularly under high-pressure conditions.

Numerical investigation of heat transfer in parallel channels with water at supercritical pressure

Thermal phenomena such as heat transfer enhancement, heat transfer deterioration, and flow instability observed at supercritical pressures as a result of fluid property variations have the potential to affect the safety of design and operation of Supercritical Water-cooled Reactor SCWR, and also challenge the capabilities of both heat transfer correlations and Computational Fluid Dynamics CFD physical models. These phenomena observed at supercritical pressures need to be thoroughly investigated.An experimental study was carried out by Xi to investigate flow instability in parallel channels at supercritical pressures under different mass flow rates, pressures, and axial power shapes. Experimental data on flow instability at inlet of the heated channels were obtained but no heat transfer data along the axial length was obtained. This numerical study used 3D numerical tool STAR-CCM+ to investigate heat transfer at supercritical pressures along the axial lengths of the parallel channels with water ahead of experimental data. Homogeneous axial power shape HAPS was adopted and the heating powers adopted in this work were below the experimental threshold heating powers obtained for HAPS by Xi. The results show that the Fluid Centre-line Temperature FCLT increased linearly below and above the PCT region, but flattened at the PCT region for all the system parameters considered. The inlet temperature, heating power, pressure, gravity and mass flow rate have effects on WT (wall temperature) values in the NHT (normal heat transfer), EHT (enhanced heat transfer), DHT (deteriorated heat transfer) and recovery from DHT regions. While variation of all other system parameters in the EHT and PCT regions showed no significant difference in the WT and FCLT values respectively, the WT and FCLT values respectively increased with pressure in these regions. For most of the system parameters considered, the FCLT and WT values obtained in the two channels were nearly the same. The numerical study was not quantitatively compared with experimental data along the axial lengths of the parallel channels, but it was observed that the numerical tool STAR-CCM+ adopted was able to capture the trends for NHT, EHT, DHT and recovery from DHT regions. The heating powers used for the various simulations were below the experimentally observed threshold heating powers, but heat transfer deterioration HTD was observed, confirming the previous finding that HTD could occur before the occurrence of unstable behavior at supercritical pressures. For purposes of comparing the results of numerical simulations with experimental data, the heat transfer data on temperature oscillations obtained at the outlet of the heated channels and instability boundary results obtained at the inlet of the heated channels were compared. The numerical results obtained quite well agree with the experimental data. This work calls for provision of experimental data on heat transfer in parallel channels at supercritical pressures for validation of similar numerical studies.

Effect of axial power distribution on flow instability in parallel channels with water at supercritical pressures

Stability of the operation of SCWR at the supercritical conditions has become a major concern to the nuclear engineers worldwide, especially around the pseudo-critical point where dramatic changes of the fluid properties are experienced. This concern of addressing supercritical flow instability is due to various studies showing that the efficiency of light water reactors can be improved considerably from 33% at subcritical pressures to 45% at supercritical pressures.This work investigates effect of axial power distribution on flow instability in parallel channels with water at supercritical pressures. The axial power distributions considered include axially decreased and homogeneous axial power shapes. The study also seeks to examine the performance of the 3D numerical tool STAR-CCM+ in predicting flow instability at supercritical pressures. The effects of parameters such as mass flow rate, pressure and gravity on flow instability were also discussed. Sensitivity analysis of some selected turbulence models and time steps were initially carried out with the aim of selecting suitable turbulence model and time step for the numerical simulations. For a system operated exclusively at two different pressures 23 MPa and 25 MPa with 125 kg/h mass flow rate, inlet temperatures from 180 °C to 360 °C and with gravity influence, stability of the system with axially decreased or homogeneous power shape decreases and increases respectively below and above a particular threshold power with inlet temperature. The system was found more stable with homogeneous Axial Power Shape (HAPS) than that of axially decreased power shape (ADPS). Similar observation was made for a system operated at mass flow rate of 125 kg/h, system pressure of 23 MPa, and with or without gravity influence. When the system is operated at 23 MPa pressure, 145 kg/h mass flow rate and with gravity influence, there is a threshold power below which stability decreases and above which stability increases with inlet temperature for HAPS. In this case, there is no inflection point for ADPS and the stability deceases with inlet temperature. At low inlet temperatures, the system is more stable with ADPS but at high inlet temperatures the system is more stable with HAPS for the system operated at 23 MPa pressure and 145 kg/h mass flow rate. The system with ADPS or HAPS becomes more stable with the change of the system operated with gravity influence to the system operated without gravity influence, and also with the change of system pressure from 23 MPa to 25 MPa. With the change of the system mass flow rate from 125 kg/h to 145vkg/h, the system with ADPS becomes more stable at low inlet temperatures and less stable at high inlet temperatures, whereas the system with HAPS becomes less stable for the various inlet temperatures. For a system operated under experimental conditions of 125 kg/h mass flow rate, 23 MPa pressure, inlet temperatures from 180 °C to 260 °C and with gravity influence, the trends of the numerical instability boundary results agree quite well with the trends of the experimental instability boundary results for most of the inlet temperatures, but there is no inflection point for experimental HAPS as it is the case for numerical HAPS. The system with HAPS is more stable than that with ADPS, a finding also obtained for the experimental results. The dynamics characteristics such as amplitude and period were also compared with experimental results for ADPS at mass flow rate of 125 kg/h, 23 MPa, and with gravity influence. The experimental amplitude was largely under-predicted and the experimental period was quite well predicted by the numerical tool adopted. The results of this study show that the type of axial power shape adopted in supplying heat to the fluid flowing through heat transfer system has significant effect on the stability of the system. From the comparison of the numerical and experimental results, the 3D numerical tool, STAR-CCM+ code could predict flow instability in the parallel channels irrespective of the type of axial power shape adopted. However, there is an evidence that the presence of heating structures in the geometrical model adopted may change the predicted behavior, as shown in previous works. More relevant experiments at supercritical pressures should be carried out for validation of numerical tools adopted for similar studies.

Numerical investigation of flow instability in parallel channels with supercritical water

SCWR is one of the selected Gen IV reactors purposely for electricity generation in the near future. It is a promising technology with higher efficiency compared to current LWRs but without the challenges of heat transfer and its associated flow instability. Supercritical flow instability is mainly caused by sharp change in the coolant properties around the pseudo-critical point of the working fluid and research into this phenomenon is needed to address concerns of flow instability at supercritical pressures.Flow instability in parallel channels at supercritical pressures is investigated in this paper using a three dimensional (3D) numerical tool (STAR-CCM+). The dynamics characteristics such as amplitude and period of out-of-phase inlet mass flow oscillation at the heated channel inlet, and heat transfer characteristic such as maximum outlet temperature of the heated channel outlet temperature oscillation are discussed. Influences of system parameters such as axial power shape, pressure, mass flow rate, and gravity are discussed based on the obtained mass flow and temperature oscillations. The results show that the system parameters have significant effect on the amplitude of the mass flow oscillation and maximum temperature of the heated outlet temperature oscillation but have little effect on the period of the mass flow oscillation. The amplitude of mass flow oscillation and maximum temperature of the heated channel outlet temperature oscillation increase with heating power. The numerical results when compared to experiment data show that the 3D numerical tool (STAR-CCM+) could capture dynamics and heat transfer characteristics of the flow quite well and also predict flow instability in the parallel channels.

Effects of axial power shapes on CHF locations in a single tube and in rod bundle assemblies

Abstract Currently, the prediction of rod bundle CHF is dependent on CHF correlations derived from CHF data. A simple correction factor, such as F-factor, is often used to account for the axial power shape differences based on a simple accumulated energy concept, which has totally no consideration on the impact of true local condition on CHF mechanism. Subsequently, as expected, large uncertainty is often associated with the CHF value and CHF location predictions. For the purpose of obtaining different power shapes effects on CHF, CFD calculated parameter values were used to predict the possible CHF occurrence location. The possible CHF location prediction method proposed in this paper is calculated void fraction, heat transfer coefficient (HTC), liquid temperature distribution and detailed local parameters. And the uniform and non-uniform CHF were analyzed. The prediction of possible CHF locations in a 5 × 5 rod bundle may provide useful information for the design of a full-length CHF test, enhance the accuracy of CHF and CHF location prediction, and reduce the costs of the experimentation.

A Steady-State Thermal-Hydraulic Analysis Method for Prismatic High-Temperature Gas-Cooled Reactors

A new methodology for the accurate and efficient determination of steady-state thermal-hydraulic parameters for prismatic high-temperature gas reactors is developed. Whole-core steady-state temperature, pressure, and mass flow distributions are determined for the conceptual MHTGR-350 [Modular High Temperature Gas Reactor] reactor design and also for a range of values of the important parameters. Full-core three-dimensional heat conduction calculations are performed at the individual fuel pin and lattice assembly block levels. A simplified one-dimensional fluid model is developed to predict convective heat removal rates from solid core nodes. Downstream fluid properties are determined by performing a channel energy balance along the axial node length. To establish flow distribution, channel exit pressures are compared, and inlet mass flows are adjusted until a uniform outlet pressure is reached. Bypass gaps between assembly blocks as well as coolant channels are modeled. Finite volume discretization of energy and momentum conservation equations are formulated and explicitly integrated in time. Iterations are performed until all local core temperatures stabilize and global convective heat removal matches heat generation.Whole-core steady-state, thermal-hydraulic results are presented for various axial power and uniform radial power configurations. For all cases, peak temperatures were below expected normal operational limits for TRISO fuels. Bottom-peaked axial power shapes had the highest peak temperatures but the lowest average temperatures. Different reactor designs with increased core inlet temperatures, reduced flow rates, or higher-power-density fuels could however challenge temperature limits. Partial assembly hydrodynamic and temperature results compared favorably with those available in the literature for similar analyses.

Investigation of two-phase flow instabilities under advanced PWR conditions

The characteristics of two-phase flow instability under advanced pressurized water reactor (PWR) conditions are explored in this study. Density wave oscillations in vertical parallel tubes are experimentally investigated by employing three different test sections to observe the parametric effects of the axial power shape, unheated riser, and channel inlet throttling on the onset of flow instability (OFI). A critical heat flux (CHF) accompanied by flow oscillation is indicated under certain test conditions with high pressure and/or low inlet subcoolings. A frequency-domain linear stability analysis code, called ALFS_F, was developed on the basis of one-dimensional two-phase mixture equations and a reactivity feedback model with point neutron kinetics and lumped parameter heat conduction equations. The mass flux at OFI conditions and the oscillation periods are calculated by the ALFS_F to compare with the experimental data. The thermal margin characteristics are illustrated with a comparative analysis of OFI and CHF margins under the core conditions of SMART, which is an integral-type advanced PWR being developed at KAERI.

Transient thermal–hydraulic analysis of complete single channel blockage accident of generic 10 MW research reactor

Thermal–hydraulic behavior for a complete blockage of a single fuel channel in a generic 10MW research reactor is studied by using the system analysis code RELAP5/MOD3.3 which is widely used in the nuclear industry. Fuel assembly geometry is lumped into a 4 channel model to model high and average power cases which are spatially discretized. Various axial power shapes coming from different control rods positions are considered in the analysis, where the minimum wall subcooled margin is found to exist for case with highest peaking for an average powered channel blockage transient. Vapor generation is observed from first and second highest peaking cases where cyclic variation of vapor inventory inside a blocked channel resulted in oscillatory behavior of the fuel temperature. Effect of a presence of an oxide layer is also tested which showed a slight increase in structure temperatures and vapor generation. Point kinetics model is utilized in the analysis code to observe the effect of reactivity feedback and consequences from different application ranges are compared. Analysis shows a consideration of assembly wise feedback results in increased feedback effect and decreased boiling which deviate from single channel wise feedback case. This calls for a detailed multi-dimensional simulation with neutronics and thermal–hydraulics simultaneously considered. Analyses results show that the consideration of feedback improves the outcome in terms of fuel temperature, and its integrity is conserved for all test cases.

An experimental investigation of flow instability between two heated parallel channels with supercritical water

Super critical water reactor (SCWR) is the generation IV nuclear reactor in the world. Under normal operation, water enters SCWR from cold leg with a temperature of 280°C and then leaves the core with a temperature of 500°C. Due to the sharp change of temperature, there is a huge density change in the core, which could result in potential flow instability and the safety of reactor would be threatened consequently. So it is necessary to carry out relevant investigation in this field.An experimental investigation which concerns with out of phase flow instability between two heated parallel channels with supercritical water has been carried out in this paper. Due to two INCONEL 625 pipes with a thickness of 6.5mm are adopted, more experimental results are attained. To find out the influence of axial power shape on the onset of flow instability, each heated channel is divided into two sections and the heating power of each section can be controlled separately. Finally the instability boundaries are obtained under different inlet temperatures, axial power shapes, total inlet mass flow rates and system pressures. The dynamics characteristics of out of phase oscillation are also analyzed.

Core axial power shape reconstruction based on radial basis function neural network

The core axial power shape reconstruction method based on radial basis function neural network (RBFNN) is proposed, and 18-node axial core power shape can be reconstructed from 6-segment in-core detector signals or 6-segment ex-core detector signals. Alternating conditional expectation algorithm is also used to validate the effectiveness of RBFNN algorithm. The results show that no matter what kind of detector is used, the RBFNN algorithm performs better than the ACE algorithm. The RBFNN algorithm has a good anti-noise ability when the in-core detector signals with noise are used, while the correct axial power shape can’t be reconstructed from the ex-core detector signals with noise. By analyzing the ex-core axial spatial weighting function’s condition number, the determination of the axial power shape from ex-core detector signals is found to be a typical ill-posed inverse problem. A regularized RBFNN algorithm is used to eliminate this ill-posedness and get the physically meaningful axial power shape.

Reactor power shape synthesis using Group Method of Data Handling

This paper describes the Group Method of Data Handling (GMDH) algorithm to reconstruct 20-node axial core power shapes from five-level in-core detector power measurements. Conventional methods in this field are the parameter identification framework by adopting fixed basis functions and the determination of parameters weighted to each terms of known functions like Fourier series or spline forms. This study proposes GMDH, an evolutionary algorithm, to find the optimal functional form and parameters simultaneously those describe the power distribution accurately. A total of 1060 cases of axially 1-dimensional core power distributions of 20-nodes are generated by 3-dimensional core design code covering beginning-of-life to end-of-life core burnup histories to validate the method. Axially five-point box powers at in-core detector positions are considered as measured snapshots. The reconstructed axial power shapes using GMDH method are compared to the reference power shapes. The results show that the proposed method is very robust and accurate compared with the spline fitting method.

A Feasibility Study of Load-Follow Operation Without Boron Adjustment for CPR1000

This paper presents mathematical modeling of dynamic phenomena in large pressurized water reactors to study load-follow capability. One of the main reactor types in China’s national nuclear development, CPR1000, uses a mode G control method, with G banks, N banks, R banks, and soluble boron to adjust reactor power changes and the axial power shape. In this paper, a new control mode is adopted that can follow the daily variation of power demand without changing the boron concentration. The control banks are regrouped to realize reactivity/temperature control by M banks and axial offset control by an AO bank. A two-node dynamic core model is constructed, taking into account the coupling coefficient and the mutual influence. The transient parameters are obtained by steady-state calculation of a single channel using the original design and operation parameters of CPR1000. Then, to adopt a control mode without soluble boron adjustment, the optimal control implementation is connected to the core simulation platform. Simulation results show that this optimal control policy can provide the capability for the CPR1000 to follow a daily load curve.

Extension for load-follow operation of PWR core by using single- or multi-variable control of state-feedback

The single- or multi-variable control of full-state feedback with a full-order observer is applied to design a nonlinear Pressurized Water Reactor (PWR) core load following control system for regulating the core power level and axial power shape. A two-point based nonlinear PWR core model without boron and with the power rod and Axial Offset (AO) rod is built. The linearized single-variable (multi-variable) core model under Case 1 (Case 2) classified by two movable regions of power rod is constructed. Based on a linearized core model, the state feedback control is implemented by utilizing the robust pole assignment method with an additional integrator for Case 1 and devising an integral decoupling control system with a dynamic controller for Case 2; a Kalman filter with robustness is designed as an observer for each case. The integration of the state feedback and the observer structures a controller of the nonlinear core model of each case. For Case 1 (Case 2), the nonlinear core model and the state feedback single-variable (multi-variable) control construct the nonlinear PWR core load following control system. Finally, the control system of each case is simulated and the simulation results show that the control system is effective.

Uniform versus Nonuniform Axial Power Distribution in Rod Bundle CHF Experiments

Rod bundle experiments with axially uniform and nonuniform heat fluxes are examined to explore the potential limitations of using uniform rod bundle CHF data for CHF correlation development of light water reactors with nonuniform axial power distribution (APD). The case of upstream burnout is presented as an example of unique phenomena associated with nonuniform rod bundle CHF experiments. It is a result from combined effect of axial nonuniform power shape and different interchannel mixing mechanisms. In addition, several key parameters are investigated with respect to their potential impacts on the thermal-hydraulic behaviors between rod bundles with uniform and nonuniform APDs. This type of misrepresentation cannot be amended or compensated through the use of correction factors due to the lack of critical information in the uniform rod bundle CHF testing as well as the fundamental difference in the underlining driving mechanisms. Other potential issues involved with the use of uniform rod bundle CHF data for nonuniform APD system applications also present strong evidence concerning the limitations and inadequacy of using uniform rod bundle CHF data for the correlation, prediction, and design limit calculation for safety analysis.

Accuracy and Uncertainty Analysis of PSBT Benchmark Exercises Using a Subchannel Code MATRA

In the framework of the OECD/NRC PSBT benchmark, the subchannel grade void distribution data and DNB data were assessed by a subchannel code, MATRA. The prediction accuracy and uncertainty of the zone-averaged void fraction at the central region of the 5 × 5 test bundle were evaluated for the steady-state and transient benchmark data. Optimum values of the turbulent mixing parameter were evaluated for the subchannel exit temperature distribution benchmark. The influence of the mixing vanes on the subchannel flow distribution was investigated through a CFD analysis. In addition, a regionwise turbulent mixing model was examined to account for the nonhomogeneous mixing characteristics caused by the vane effect. The steady-state DNB benchmark data with uniform and nonuniform axial power shapes were evaluated by employing various DNB prediction models: EPRI bundle CHF correlation, AECL-IPPE 1995 CHF lookup table, and representative mechanistic DNB models such as a sublayer dryout model and a bubble crowding model. The DNBR prediction uncertainties for various DNB models were evaluated from a Monte-Carlo simulation for a selected steady-state condition.