Rod Bundle Critical Heat Flux Research Articles

Liquid level monitoring and quenching front tracking for SMR rod bundle CHF tests under low pressure, low flow, high quality conditions

In most Light Water Reactor (LWR) based Small Modular Reactors (SMRs), one of the critical factors that might affect Critical Heat Flux (CHF) and other thermal–hydraulic phenomena under low flow, low pressure, high-quality conditions is the existence of high water column over a relatively short fuel core in the majority of the integral design SMR reactor vessels. With the existence of this water column over the relatively short and high-power peaking fuel bundle, the potential of a reflood quenching effect is highly plausible under low pressure, low flow and high-quality conditions as the location of the CHF event will be much closer to the exit plenum, especially in the case of exit power peaking. The results from CFD modeling and flow visualization tests presented in this paper demonstrate the importance of liquid level as well as its impacts on rod bundle CHF. Such a critical factor has often been overlooked by the prevailing SMR designs as many of the rod bundle CHF tests performed for SMR, including those with the licensing data, have either neglected the effects of the exit quenching effect, or were performed with non- prototypical heater rod designs or surrogate fluid that may not include such exit quenching phenomena.In this paper, the phenomenon of exit quenching on SMR rod bundle CHF has been introduced with the development of the proprietary On-line, Real-time Liquid Level Monitoring System (OR_LLMS) and On-line, Real-time dynamic Quenching Front Tracking Systems (OR_QFTS), which allow accurate measurements of rod bundle CHF as a function of both liquid level and degree of downward quenching under various test conditions. The future application is also presented, which will aid in the development of rod bundle CHF correlations that account for and take advantage of different water levels over the reactor core.

Numerical simulation prediction of critical heat flux for 5 × 5 rod bundle with multiple grid spacers and different cladding materials

The critical heat flux (CHF) is one of the safety standards for core thermal design and is crucial for the normal operation of the reactor. The CHF problem of rod bundles with grid spacers is a hot topic in related research. The successful design of grid spacers can improve channel CHF by affecting the flow of downstream coolant, and this detection analysis can be achieved through numerical simulation methods. In addition, most simulation studies have not considered the solid region of the cladding, so there is little research on the effect of cladding materials on CHF. The role of grid spacers in rod bundle CHF is the focus of this paper, and the influence of cladding materials is another discussion direction. In this paper, following meticulous mesh generation and computational model validation, a study on the CHF of the 5 × 5 rod bundle with multiple grid spacers and different cladding materials is carried out by using CFD software. The verification results indicated that the physical model established in this study can reasonably predict the generation location and critical power value in the 5 × 5 rod bundle with grid spacers. Furthermore, the study identified that the grid spacer exerts a significant influence on coolant flow within the channel, while the weakening of the mixing effect of the grid spacer on the coolant flow is the main factor to lead the boiling crisis upstream of the grid spacer. By comparing the effects of thermal properties of different materials on CHF, it was found that when the thermal conductivity ratio of the two materials is 1.08, CHF is almost identical under the condition of a reference heat flux of 0.1 MW/m2 as a step size. In addition to thermal properties, further attention may be paid to the influence of surface characteristics of materials on CHF in future work.

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.

Investigation on the DNB-type critical heat flux in a fuel assembly with 5 by 5 rods

In this paper, Eulerian two-phase model coupled with wall boiling model was employed to simulate the departure from nucleate boiling (DNB) in a 5 × 5 fuel assembly. The numerical simulation accurately predicted the critical heat flux (CHF) and also its location. Compared with the calculated results from W-3 formula and FLICA subchannel code, the CFD (Computational Fluid Dynamics) calculations showed good agreement against experimental measurements with the deviation of 5%, but presented more detailed flow distribution. Based on CFD results, the method of CHF prediction was discussed with an eye on wall superheat, void fraction and vapor convective heat flux in rod surfaces. The distribution of flow field and liquid–vapor fraction affected by mixing vane were further studied. Mixing vane can continually influence the downstream thermal hydraulic characteristics, forming the diagonal circulation and square circulation with the flow development. As a result, the bubble aggregation firstly occurred on the surface of corner hot rod instead of central hot rod, which induced the local heat transfer crisis. The CFD calculations presenting the evolution process of physical parameters can provide guidance for rod bundle CHF test and fuel assembly design.

A methodology for CHF prediction in VVER rod bundles

The critical heat flux (CHF) is an important thermal–hydraulic parameter that must be determined during the design of water-cooled reactors to ensure safety of operation. In this paper, a methodology is proposed to predict the CHF in VVER11VVER is a transliterated acronym for “vodo-vodyanoi energetichesky reaktor” which literally means “water-water power reactor” (Nuclear Energy Agency, 2021). rod bundles. A whole-reactor model of VVER-1000 is implemented in RELAP-SCDAPSIM, and, based on the bundle-average approach, the parameters of the hot channel are used to assess four CHF predictors (i.e., the 2006 Groeneveld look-up table (LUT), and the Bowring, W-3, Biasi, and OKB-Gidropress correlations) upon comparison with the 2011 Bobkov LUT. A brief critique is given of the correction factors for both the Groeneveld and Bobkov LUTs. The effects of channel diameter and non-uniform axial heat flux on the predicted values of CHF in subcooled conditions are also studied. The performance and time complexity of the standard and Bourke trilinear interpolation algorithms are assessed for use with LUTs in safety codes. It has been concluded that the non-uniform axial heat flux does not affect the CHF in subcooled conditions and that the Groeneveld diameter factor exponent gives better predictability of the CHF in VVER rod bundles over the Tanase and Wong exponent correlations. Additionally, the Groeneveld LUT has been found to nearly reproduces the Bobkov LUT CHF values when used in conjunction with the Bobkov heated length factor. Values of MDNBR equal to 1.83 and 1.96 can be used as thermal limits at 12% overpower when the Bowring and W-3 correlations, respectively, are used in subcooled conditions. The standard and Bourke algorithms nearly exhibit the same average CPU time. The Bourke algorithm, however, is less affected by noise. Further research is needed to ensure the smoothness of the Bobkov LUT and to derive reactor type-dependent correction factor formulas for use with the Groeneveld LUT.

Development of a dimensionless rod-bundle CHF correlation based on stepwise regression method, Part II: determination of correction factors accounting for the effect of unheated guide tube and non-uniform axial heat flux profile

Correlation for rod-bundle critical heat flux (CHF) is of vital importance for design and assessment of a Pressurized Water Reactor (PWR). In our previous study, the basic form of a dimensionless rod-bundle CHF correlation has been developed based on stepwise regression with CHF data obtained in test bundles of uniform axial heat flux. It proves that stepwise regression is an effective method to develop simple, but reliable CHF correlation consisting of only dimensionless parameters with rigorous physical meanings. As a continuation of the previous investigation, revision of the basic form was conducted. Dimensionless correction factors accounting for the cold-wall effect due to unheated guide tube and the effect of non-uniform axial heat flux were developed. For the data set of 486 rod-bundle CHF points obtained in 7 test bundles, the average ratio of the measured to predicted CHF is 1.002 with a standard deviation from the mean of 5.51%.

Progress in rod bundle CHF in the past 40 years

Critical Heat Flux (CHF) prediction remains one of the most important tasks in thermal hydraulics of Light Water Reactors (LWR’s). During recent years, tremendous progress has been made in understanding and predicting rod bundle CHF. However, due to the unique features associated with rod bundle subchannel systems, including complex and compact anisotropic geometry, non-uniform axial/radial heating, open channel interactions, periodic (positional) mixing vane grids (MVG’s) mixing (Yang et al., 2014), and a broad range of operation conditions (Yang and Han et al., 2019), a precise prediction of CHF in rod bundles, entirely based on analytical methods without the support of any extensive experimental database, is still beyond the present modelling and computational capabilities. For Pressurized Water Reactors (PWR’s), CHF occurs at a relatively low coolant quality and is primarily driven by the Departure from Nucleate Boiling (DNB) type event, which can be triggered by liquid sublayer dryout, bubble crowding, homogeneous nucleation, or other mechanisms causing local starvation of liquid under subcooled flow boiling or low quality saturated boiling conditions, and the dryout event, which describes CHF under a high void situation when liquid film on the fuel surface dries out, usually happens in a high quality saturated annular flow. Boiling Water Reactors (BWR’s) operate at a relatively lower pressure and can reach much higher coolant qualities that correspond to annular two-phase flow conditions. This dominant CHF mechanism, dryout, is based on the gradual thinning and drying out of the liquid film.This paper provides an overview of the progress made in the experimental and analytical investigations of CHF in both PWR and BWR rod bundles during the past 40 years. A complete comprehensive concept map is developed to outline most of the key mechanistic CHF models covering the entire range of flow regimes. The limitation and the potential approaches for the application of mechanistic models to rod bundle CHF prediction are reviewed. This includes the recently proposed, rod bundle specific NHNM (Non-uniform heater Homogeneous Nucleation Model [Yang, 2017]). Various industrial empirical correlations are reviewed together with their corresponding subchannel codes and required rod bundle CHF experiments. The need for clear guidelines and standards for benchmarking new rod bundle CHF test facilities is discussed.Various approaches on investigating, understanding, modeling, and predicting the occurrence of rod bundle CHF are reviewed from mechanistic models, semi-empirical and empirical approaches to advanced CFD simulation. The achievements and limitations in those methods are discussed with suggestions for future advancement.

Development and assessment of a new CHF mechanistic model for subcooled and low quality flow boiling

Under subcooled and low quality flow boiling conditions, a general critical heat flux (CHF) mechanistic model is essential to the thermal-hydraulic design and safety analysis, especially at the conceptual design stage of a new pressurized water reactor (PWR). In this study, an overview of the physical mechanisms behind the most common CHF models was carried out and the additional effects to be considered in a new CHF model were indicated. Then, based on the mechanism of superheated liquid layer depletion, the net mass transfer from the bulk liquid to the superheated liquid layer caused by turbulent velocity fluctuations of flowing bubbles was determined and a new calculation method of superheated liquid layer thickness was developed. After matching with other constitutive equations, a novel CHF mechanistic model for subcooled and low quality flow boiling conditions has been developed. Furthermore, the present CHF mechanistic model was validated with 5 × 5 rod bundle CHF experimental data of Nuclear Power Institute of China (NPIC) and compared with other CHF models and CF-DRW correlation. Finally, the distribution of M/P was evaluated and the advantage of the present CHF mechanistic model was discussed. The results show that the present CHF model performs as well as the CF-DRW correlation and is superior to other well established CHF models without any correction factors or empirical constants from experimental data.

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

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

Development of a dimensionless rod-bundle CHF correlation based on stepwise regression method, Part I: determination of the basic form with CHF data obtained with uniform axial heat flux profile

Accurate prediction of rod-bundle critical heat flux (CHF) is a main challenge in thermal-hydraulic design and safety assessment of Pressurized Water Reactor (PWR). The most widely applied approach is empirical correlation developed based on rod-bundle experiments carefully tested with geometry and power distribution representative to PWR working conditions. The standard procedure is to correlate the experimentally obtained CHF values as empirical functions of local thermal-hydraulic conditions obtained with subchannel analysis. However, up-to-date empirical correlations developed according to this manner are in general quite complicated with a number of dimensional coefficients and correction factors with less or non-physical meanings. In this study, stepwise regression method was used to develop a novel, dimensionless rod-bundle CHF correlation covering typical PWR working conditions. First, various dimensionless parameters affecting CHF were selected as candidate independent variables. With stepwise regression, the form and coefficients of the proposed CHF correlation were dynamically optimized and determined.

Development and assessment of a new rod-bundle CHF correlation for China fuel assemblies

One of the main objectives of subchannel analysis for a pressurized water reactor (PWR) core or fuel assembly is to predict the critical heat flux (CHF) which usually employs the correlation based on the local parameters. The effects of spacer grid, cold wall, and non-uniform heat flux are essential for the sake of accurate calculation of local parameters. The subchannel code CORTH is applied to determine the local thermal-hydraulic parameters at the occurrence of departure from nucleate boiling (DNB) for each set of bundle CHF data of China fuel (CF) assemblies. In order to develop a new CHF correlation for CF assembly, the minimum DNB ratio (MDNBR) point and the burn-out (BO) point methods are both employed. The three-step form was adopted for the correlation which considers the effects of spacer grids, cold walls, and non-uniform heat flux. The coefficients of the correlation were determined based on the nonlinear regression method. The analysis and assessment results against the experimental data indicate that the correlation can predict the CHF with high prediction accuracy and correct parametric trends. In comparison with MDNBR point method, the BO point method is recommended to predict the risk of DNB because it is more reasonable and conservative and has a better predication rate. When implemented with the subchannel code CORTH, the developed correlation can predict the thermal-hydraulic performances of CF assemblies.

DNB type critical heat flux prediction in rod bundles with simplified grid spacer based on Liquid Sublayer Dryout model

Five published Liquid Sublayer Dryout (LSD) models are assessed using the Groeneveld Look-up table 2006 (LUT-2006) and critical heat flux (CHF) experimental data in circular tube under both subcooled boiling conditions and saturated boiling conditions. The flow regime transition from inverted annular flow to dispersed flow at post-CHF is postulated as the upper limit of the LSD model application range in saturated boiling region. Lee and Mudawar’s LSD model modified by W.X. Liu shows good agreement when void fractions is lower than 0.7. The modified L&M’s LSD model is coupled into subchannel code, COBRA-TF, to predict DNB type CHF in rod bundles. In present study, the gap clearance or thermal equivalent diameter of rod bundles is adopted as characteristic length in the modified L&M’s LSD model. For rod bundles with simplified grid spacers, Karman velocity distribution equation is utilized to calculate the velocity distribution normal to pin wall. The rod bundle CHF predictions by COBRA-TF coupled with LUT-2006, Bowring’s CHF correlation and the modified L&M’s LSD model, are compared with the CHF experimental data in 2 × 2 rod bundles with non-uniform power profile. Based on reasonable predictability on both the input power and axial position at CHF, the deviations of CHF predictions by the modified L&M’s LSD model, which adopts gap clearance instead of thermal equivalent diameter as characteristic length, and Bowring’s CHF correlation are within ±20% as void fraction approaching to 0.7.

Revision of a CHF correlation for PWR under low pressure conditions with only dimensionless parameters as independent variables

Abstract Accurate prediction of the critical heat flux (CHF) is one of the key tasks of PWR core design and safety assessment, for the maximal allowable heat flux in the reactor core is limited by CHF. Since CHF in rod bundle cannot be predicted analytically, up-to-date predictive approach is based on empirical correlations related to the local thermal-hydraulic conditions, geometry and power distribution. However, development of CHF correlation for PWR fuel assemblies under low pressure conditions (2–10 MPa) is constrained by limited amount of experimental data points, which builds up in statistics a typical problem of small sample amounts, but requiring simultaneously high prediction accuracy. In our previous study, stepwise regression method was applied to develop a dimensional, empirical CHF correlation for PWR under low pressure conditions, termed as the advanced low pressure CHF correlation (ALPC), which successfully solves the challenge of small sample problem. However, the ALPC correlation still uses dimensional independent variables with less physical meanings, which limits its physical interpretability. In the current study, stepwise regression method was used to develop a revised, dimensionless version of the ALPC CHF correlation. First, various dimensionless, two-phase thermal-hydraulic parameters that might influence CHF were selected as candidate independent variables. With stepwise regression, the form and coefficients of the revised CHF correlation were optimized in a dynamic manner. Compared to the current ALPC correlation, the revised version developed in this study possesses a similar simple form but a much higher prediction accuracy. Revision of the ALPC correlation demonstrates clearly the advantages of utilizing dimensionless parameters as independent variables in CHF correlation, which points out a new direction of developing rod-bundle CHF correlations for engineering purpose.

Heat Loss Simulations in Rod Bundle Tests

Nuclear fuel performance is generally derived from large-scale rod bundle tests, which mainly include rod bundle critical heat flux (CHF) tests and rod bundle mixing tests under conditions of high pressure, high mass flux, and high power. Though few differences exist in the test section configuration and operation conditions, CHF tests and mixing tests generally share the same test loop, the same pressure-bearing housing, and similar flow channel designs. In rod bundle tests, the test section consists of an inner flow path with an outer surrounding chamber (outer chamber) filled with stagnant coolant. Depending on the operating conditions, the massive coolant in the outer chamber may become a significant source of heat loss/heat gain resulting from natural circulation in the outer chamber. The complex structure of the test section and variable conditions appear to be obstacles to heat loss/gain estimation.In the past, most of the calculations and discussions for heat loss/gain correction were based on a steady-state assumption without exploring the impacts of true heat loss/gain resulting from potential transient operation. That is, in rod bundle tests, in order to determine the heat loss characteristics, only single-phase heat loss tests were conducted under different temperature steady-state conditions. With the limited parameter measured, the transient behaviors of the outer chamber are not reflected. In addition, potential heat gain conditions are neglected in single-phase heat loss tests. Based on heat loss test data, a correlation between heat loss and temperature difference of the test section inlet and exit is often derived. Given this situation, in this paper, lumped-parameter analyses are conducted to evaluate the heat loss/gain trend. A system code is applied to simulate the test section and surrounding systems under typical CHF test conditions to explore heat loss/gain transient behaviors, respectively. Methods to minimize the heat loss/gain effect through the design and operating scheme are discussed at the end.

New models of droplet deposition and entrainment for prediction of CHF in cylindrical rod bundles

In this paper, we present a new set of model of droplet deposition and entrainment in cylindrical rod bundles based on the previously proposed model for annuli (effectively a “one-rod” bundle) (2016a). These models make it possible to evaluate the differences of the rates of droplet deposition and entrainment for the respective rods and for the outer tube by taking into account the geometrical characteristics of the rod bundles. Using these models, a phenomenological model to predict the CHF (critical heat flux) for upward annular flow in vertical rod bundles is described. The performance of the model is tested against the experimental data of Becker et al. (1964) for CHF in 3-rod and 7-rod bundles. These data include tests in which only the rods were heated and data for simultaneous uniform and non-uniform heating of the rods and the outer tube. It was shown that the predicted CHFs by the present model agree well with the experimental data and with the experimental observation that dryout occurred first on the outer rods in 7-rod bundles. It is expected that the methodology used will be generally applicable in the prediction of CHF in rod bundles.

Experimental investigation on repeatability of CHF in rod bundle with non-uniform axial heat flux distribution

Critical heat flux (CHF) is important to thermal-hydraulics design and safety analysis. In the purpose to validate the CHF data acquired from LS-THTF of NPIC, experimental investigation on repeatability of the CHF data is carried out with full length non-uniform axial heat flux distribution rod bundle. The repeatability of the CHF data from LS-THTF itself is examined by comparing the data from the same run but different time and data from different runs. The rod bundle powers of CHF data acquired under well controlled parameters are directly compared with reference measured data from HTRF of Columbia University. The rod bundle power ratio of CHF data is employed to describe the repeatability of CHF data from LS-THTF to HTRF, and the variations of power ratio with inlet and outlet parameters show a reasonable trend. The point to point differences of CHF power between the two facilities of every single point are found to be within 5%. The result demonstrates good repeatability between data from LS-THTF and data from HTRF. LS-THTF is capable of acquiring reliable and repeatable CHF data to verifying thermal-hydraulics performance of PWR fuel assemblies.

Effect of spacer grids on CHF in nuclear rod bundles

In nuclear reactor fuel assemblies, spacer grids are used in the rod bundles to provide mechanical support to the rods and to reduce flow-induced vibration. These are known to play a role in the local heat transfer as well as in the onset of the critical heat flux (CHF) condition. While considerable literature exists on these issues, a detailed study of the flow situation has not yet been reported. The objective of the present study was to investigate the effect of spacer grids on flow and heat transfer in rod bundles. To this end, computational fluid dynamics (CFD) based simulations have been carried out using the Eulerian–Lagrangian framework for calculating single-phase, vapour-only flow as well as the droplet-laden flow through a typical bundle geometry. The results show that the single-phase convective heat transfer coefficient increases locally by 30–40%, leading to a consequent decrease in the wall temperature, in the spacer grid region. Droplet trajectory calculations show that the droplet deposition rate is increased considerably in the spacer grid region and that about 15% of the droplet flux is deposited in the spacer grid. Estimates of the size of the re-entrained droplets have shown that these are much smaller than those of the undeposited droplets present in pre-CHF flow. It is hypothesized here that these smaller re-entrained droplets may cause enhancement of the evaporative heat transfer coefficient between the droplets and may help in bringing down the surface temperature further downstream of the spacer, which has been observed experimentally.

Development of a bundle correction method and its application to predicting CHF in rod bundles

A bundle correction method, based on the conservation laws of mass, energy, and momentum in an open subchannel, is proposed for the prediction of the critical heat flux (CHF) in rod bundles from round tube CHF correlations without detailed subchannel analysis. It takes into account the effects of the enthalpy and mass velocity distributions at subchannel level using the first derivatives of CHF with respect to the independent parameters. Three different CHF correlations for tubes (Groeneveld's CHF table, Katto correlation, and Biasi correlation) have been examined with uniformly heated bundle CHF data collected from various sources. A limited number of CHF data from a non-uniformly heated rod bundle are also evaluated with the aid of Tong's F- factor . The proposed method shows satisfactory CHF predictions for rod bundles both uniform and non-uniform power distributions.

The Current Status of Theoretically Based Approaches to the Prediction of the Critical Heat Flux in Flow Boiling

The phenomena governing the critical heat flux (CHF) in flow boiling in ducts and rod bundles vary with the flow pattern. Separate models are needed for dryout in annular flow, wall overheating in plug or slug flow, and formation of a vapor blanket in dispersed flow. The major theories and their current status are described for the annular and dispersed regions. The need for development of a theoretical approach in the plug and slug flow region is indicated.Current phenomenological models are capable of providing satisfactory predictions of rod bundle CHF in both the annular and dispersed flow areas. However, the precision obtained remains below that of empirical correlations.

Rod Bundle Critical Heat Flux at Low Pressure

Critical heat flux (CHF) tests were performed at low pressure in a close-packed rod bundle. The electrically heated test bundle had geometrical configurations the same as those of the Power Burst Facility nuclear core. Existing low-pressure CHF correlations, namely, those of Bernath and Lund, did not correlate well with the test data. The Bernath correlation overpredicts CHF in some cases by a factor of 5 when compared with measured values. Lund’s correlation overpredicts CHF at measured CHF values above 1.5 MW/m2, and underpredicts CHF at measured CHF values below 1.5 MW/m2. These CHF tests provided the first close-packed rod bundle data with a sufficient data base to develop a correlation. The study examined CHF with absolute coolant system pressures of 117 to 255 kPa, mass velocities of 1992 to 4830 kg/s· m2, and subcooling of up to 53°C, with a rod spacing of 1.02 mm. The effect of rod bowing was examined with the rod spacing reduced in varying degrees to a minimum of 0.0508 mm. Motion pictures of the rod bundle during CHF with nominal spacing and bowed rods show that CHF occurs in the rod gap and does not propagate azimuthally on the rod surface.A CHF correlation developed from the test data correlates the data with a standard deviation of 8.79%.