Investigation of uncovered-bundle heat transfer under high-pressure boil-off conditions

Uncovered-bundle heat transfer tests under high-pressure boil-off conditions were conducted in ranges of pressure from 3 to 12 MPa, heat flux from 3.3 to 18 W/cm² and vapor Reynolds number from 10000 to 62000. There existed two boil-off patterns depending on bundle power, low-power boil-off and high-power boil-off. Though existing steam cooling heat transfer correlations fitted approximately the low-power boil-off test data, they underpredicted considerably the high-power boil-off test data. Intermittent droplet cooling effect in the froth region was fairly larger in high-power boil-off than in low-power boil-off. The FLECHT correlation was modified to account for the intermittent droplet cooling effect in the froth region by using the test data. The modified correlation was in approximate agreement with the high-power boil-off test data as well as the lower-power boil-off test data.     

Matrix calculation of radiant heat transfer in LWR fuel bundles under accident conditions

A computer code was developed for calculating the radiant heat transfer in a LWR fuel bundle under accident conditions. The calculation method is a modular one: a fuel bundle or its part is divided into unit cells, each of which is composed of a coolant subchannel surrounded by several segments of solid or imaginary faces. The view factor matrix in each cell is expanded over the whole bundle using the concept of ‘boundary face’ between cells, and the resultant heat transfer equations are simultaneously solved for solid wall temperatures. The geometrical flexibility of this method is suitable for treating various simulation experiments for accidents. The method is also effective for repeated calculations of the radiant heat transfer reflecting state or material property changes when analyzing fuel rod behaviour under accident conditions.     

A theoretically based critical heat flux prediction for rod bundles at PWR conditions

A theoretically based procedure developed for round tubes has been applied to the prediction of DNB heat fluxes in rod bundles at PWR conditions. State-of-the-art subchannel analysis procedures were used to determine local flows and enthalpies. Very good comparison between DNB predictions and experimental observations are found for rod bundles which both uniform and non-uniform axial heat fluxes.     

Critical heat flux prediction in rod bundles under upward low mass flux densities

The analysis of experimental data and results of calculations for heat transfer crisis in heated channels under low upward coolant mass flux densities is presented. This analysis allows the determination of the basic features of the boiling crisis phenomenon. It is shown that the methods currently used for critical heat flux (CHF) prediction have insufficient accuracy in the given range of parameters. A new relationship for the CHF calculation is presented. It should be used for the water–water energy reactor (WWER) and uran–graphite channel reactor—Chernobyl-type (RBMK) rod bundles, and is verified by the test data. The comparison of results obtained by a new CHF correlation and the relationship used in RELAP5/MOD3.1 Code is presented. It is shown that the latter overpredicts the CHF values at atmospheric pressure and for x_cr>0.4 and does not provide conservative estimations for the RBMK fuel bundles.     

Void-fraction distribution under high-pressure boil-off conditions in rod bundle geometry

Void fractions in a simulated pressurized water reactor (PWR) core rod bundle geometry were measured under boil-off conditions covering pressures from 3 to 12 MPa and mass fluxes from 5 to 100 kg m⁻² s⁻¹, with a particular interest in void fractions at higher pressures and relatively high mass fluxes. Test results showed that the Chexal-Lellouche model predicts best the present (volume-averaged) void-fraction data among correlations and models examined in this study. The volume-averaged void fractions obtained from differential pressure measurements are systematically smaller than the chordally averaged void fractions obtained from gamma densitometer measurements. Local void fractions were measured in the same bundle for non-heated steam-water two-phase flow of 3 MPa by using an optical void probe. It was found that the difference between the volume-averaged and chordally averaged void fractions mentioned above can be explained qualitatively by a local void-fraction distribution in the bundle measured in the latter tests.     

Critical heat flux prediction in vertical bottom-closed rod bundles

This paper presents the analysis of experimental data and calculational relationships for heat the transfer crisis in LWR rod bundle with closed bottom. A new relationship for critical heat flux prediction in the rod bundle with closed bottom based on the improved drift model is described. The comparison of critical heat flux values given by different correlations (including Groeneveld's algorithm used in RELAP5/MOD3.1 Code) and those obtained from the tests in the wide range of regime and geometric parameters is presented.     

Critical heat flux under flow oscillation of water at low-pressure, low-flow conditions

In this work, the effect of flow oscillations on critical heat flux (CHF) is investigated for water flow in vertical round tubes at low-pressure, low-flow (LPLF) conditions. An experimental study has been conducted to investigate the difference in CHF between forced and natural circulations, and between stable and oscillating flow conditions with three vertical round tube test sections (5.0 mm ID×0.6 m in length, 6.6 mm ID×0.5 m in length, and 9.8 mm ID×0.6 m in length) for mass fluxes below 400 kg m⁻² s⁻¹ under near atmospheric pressure. It is found that flow oscillations can drastically reduce the CHF, in particular for natural-circulation conditions. In addition to the experiments, CHF correction factors for flow oscillation effects are developed for forced and natural circulations, respectively, based on the experimental data of the present work and others.     

Stress biaxiality in high-burnup PWR fuel cladding under reactivity-initiated accident conditions

In order to promote a better understanding of failure mechanisms of high-burnup pressurized water reactor (PWR) fuels under reactivity-initiated accident (RIA) conditions, stress biaxiality in cladding has been estimated for the pellet-cladding (PC) mechanical interaction (PCMI) phase. The estimation was based on an analysis of the transient elongations of a pellet stack and a cladding tube measured in RIA-simulating experiments in the nuclear safety research reactor (NSRR) using the RANNS code. Stress biaxiality in the high-burnup PWR fuel cladding during the PCMI phase has been estimated to be 0.7–0.8, on average, at the mid-wall of the cladding. A comparison with fresh fuel test results and a sensitivity analysis showed that the effects of burnup and pulse width on cladding stress biaxiality are less than 10% for the investigated range. The present analysis also indicated that PC friction is strong, and that the cladding constraint on pellet stack elongation is significant irrespective of burnup. Therefore, it is recommended that strong PC friction be assumed, which is similar to the mechanical bonding condition, and that fuel pellets be treated as deformable materials in models of fuel behavior during the PCMI phase.     

Critical heat flux for finned and unfinned geometries under low flow and low pressure conditions

This paper presents the experiment and analysis for the critical heat flux (CHF) in a vertical annulus with finned and unfinned geometries under low flow and low pressure conditions. To consider the fin effect on CHF, the tests were performed on both finned heater and unfinned heater having same dimension as finned heater without fins. An analytical model was applied to estimate the heat flux and temperature distributions along the periphery of the finned geometry. The physical phenomena observed during the experiments are discussed and the parametric trends of the obtained data are examined to investigate the CHF characteristics for the finned geometry. A new correlation is proposed to predict the CHF for both finned and unfinned geometries at low flow and low pressure conditions. The developed correlation predicts the experimental data with an RMS error of 13.7%.     

The effective thermal conductivity of crud and heat transfer from crud-coated PWR fuel

Water-filled crud on the surface of PWR fuel could offer resistance to the flow of heat, which might be expected to cause higher clad temperatures, and probably more fuel failures, than are actually observed. However, there is some evidence from post-irradiation inspection that the crud is penetrated by pores large enough to permit vapour formation, and it is believed these provide a mechanism for ‘wick boiling’ to occur, which modifies, and indeed can under some circumstances actually improve, heat transfer. This phenomenon is investigated using a two-dimensional coupled multi-physics model, accounting for the flow of water, heat and dissolved species within the crud. The fuel thermal performance is characterized in terms of an effective crud thermal conductivity derived from the use of this model, and the non-linear dependence this effective thermal conductivity has on parameters such as crud thickness and pore density is determined.     

Critical heat flux under zero flow conditions in vertical annulus with uniformly and non-uniformly heated sections

The experimental study of water CHF (critical heat flux) under zero flow conditions has been carried out in an annulus flow channel with uniformly and non-uniformly heated sections over a pressure range of 0.52–14.96 MPa. In the present boiling system, the CHFs occur in the upper region of the heated section, in contrast to the results in the experiments for boiling tubes conducted by several investigators. The general trend of the CHF with pressure is that the CHF increases up to a medium pressure of about 6–8 MPa and decreases as the pressure is further increased. A comparison of the present data with the existing flooding CHF correlations shows that the correlations depend greatly on the effect of the heat flux distribution. When the correction terms with the density ratio and the effect of the heat flux distribution proposed in the present work are used with the CHF correlation based on the Wallis flooding correlation, it predicts the measured flooding CHF within an RMS error of 9.0%.     

An experimental and analytical investigation of uncovered core heat transfer under high pressure, low heat flux conditions

Experimental and analytical results are reported from two series of high pressure core uncovering experiments. It was determined that the uncovered core is cooled primarily by convection and radiation to dry steam and that droplets are confined to the immediate vicinity of the mixture level. Spacer grids substantially increased heat transfer at and downstream of the grid. A simple heat transfer model is presented which accurately predicts uncovered core heat transfer at modified wall Reynolds numbers greater than 2000. Results are expected to be use in modelling small break loss of coolant accidents.     

Energy transfer mechanisms in LMR rod bundles under mixed convection conditions

The energy transfer mechanisms operative under mixed convection conditions in heated rod arrays are identified. The relative importance of these various mechanisms of energy transfer has been assessed and is presented in regime maps which allow the designer and experimentalist to rapidly assess the characteristics of an operating regime and the capabilities needed in an analysis tool to treat such a regime. Azimuthal heater rod conduction is important for boron nitride insulated rods typically utilized in out-of-pile sodium tests. Therefore the comparison of energy transfer mechanisms necessitated the development of a correlation for azimuthal heater rod conduction. This correlation has been derived from a physical model formulated to describe this conduction effect. The proposed correlation has been tested against experimental data from out-of-pile tests and has been proven accurate for predicting the azimuthal heater rod conduction effect in LMR rod bundles.     

Critical heat flux under choking flow conditions: Part II – Maximum values of flow parameters attained under choking flow conditions

Under conditions of forced convective boiling at low pressures and high mass fluxes, beyond a certain quality, choking flow may occur at the exit of a heated channel. An experimental investigation carried out by Olekhnovitch et al. (Olekhnovitch, A., Teyssedou, A., Tye, P., Champagne, P., 2000. Critical heat flux under choking flow conditions. Part I — Outlet pressure fluctuations. Nucl. Eng. Des., this issue) has shown that the occurrence of choking flow does not radically influence the values of the critical heat flux (CHF). However, once the choking flow conditions have occurred, for a given mass flux and quality, the outlet pressure cannot be lowered below a certain value that is fixed by the flow itself. A model that allows this pressure to be determined and which must be used in conjuction with correlations for the prediction of the CHF is presented.     

Pressure drop and heat transfer in gas-cooled rod bundles

Extensive experimental and analytical investigations of fluid flow and heat transfer in gas-cooled rod bundles have been carried out. Different bundle geometries with partially or fully roughened rod surfaces were tested in a carbon dioxide loop. An advanced and comprehensive measuring control and instrumentation are important design features of this experiment. Comprehensive thermal hydraulic subchannel analysis computer codes have been developed in order to assist fuel element design calculation for gas-cooled reactors. The experiments, codes and their verification procedure are described and the results of comparisons between measured and calculated pressure and temperature distributions are given.     

A quantitative estimate on the heat transfer in cylindrical fuel rods to account for flux depression inside fuel

In a nuclear reactor, the power is limited by thermal rather than by nuclear considerations. The reactor core must be operated at a power level that the temperatures of the fuel and cladding anywhere in the core must not exceed safety limits to avoid damages in the fuel elements.Heat transfer from fuel pins can be calculated analytically by using a flat power density in the fuel pin. In actual practice, the neutron flux distribution inside fuel pins results in a smaller effective distance for the heat to be transported to the coolant. This inherent phenomenon gives rise to a heat transfer benefit in fuel pin temperatures.In this research, a quantitative estimate for transferring heat from cylindrical fuel rods is accomplished by considering a non-uniform neutron flux, which leads to a flux depression factor. This, in turn, alters the temperature inside the fuel pin. A theoretical relationship combining the flux depression factor and a ratio of temperature gradients for uniform and non-uniform is derived, and a computational program, based on finite volume method and energy balance, is developed to validate the considered approximation.     

Liquid metal heat transfer in heat exchangers under low flow rate conditions

The present paper describes the liquid metal heat transfer in heat exchangers under low flow rate conditions. Measured data from some experiments indicate that heat transfer coefficients of liquid metals at very low Péclet number are much lower than what are predicted by the well-known empirical relations. The cause of this phenomenon was not fully understood for many years. In the present study, one countercurrent-type heat exchanger is analyzed using three, separated countercurrent heat exchanger models: one is a heat exchanger model in the tube bank region, while the upper and lower plena are modeled as two heat exchangers with a single heat transfer tube. In all three heat exchangers, the same empirical correlation is used in the heat transfer calculation on the tube and the shell sides. The Nusselt number, as a function of the Péclet number, calculated from measured temperature and flow rate data in a 50 MW experimental facility was correctly reproduced by the calculation result, when the calculated result is processed in the same way as the experiment. Finally, it is clarified that the deviation is a superficial phenomenon which is caused by the heat transfer in the plena of the heat exchanger.     

Measurement of homogeneity of U and Pu isotopes in PWR spent fuel powder mixture

We present a feasibility study of the homogenization of pressurized water reactor spent nuclear fuel (SNF) powder through a mechanical mixing process. Because burn-up of the SNF depends on the position in the SNF assembly, concentrations of uranium, plutonium, and fission products are distributed differently according to the burn-up profile. The heterogeneity of the material elements affects the selection of a representative sample for quantitative analysis. Homogenization process improvement to reduce the sampling error is thus required to precisely determine the amount of uranium, plutonium, and fission products in the SNF. In this study, fine powders (<70 μm) extracted from one SNF rod were mixed, and the degree of homogenization was determined as a function of the mixing time indicating the relative standard deviations of the ¹³⁴Cs/¹³⁷Cs, Pu, and U isotope ratio measurements.     

Critical heat flux of water in vertical round tubes at low pressure and low flow conditions

An experimental study on critical heat flux (CHF) has been performed for water flow in vertical round tubes under low pressure and low flow (LPLF) conditions to provide a systematic data base and to investigate parametric trends. Totally 513 experimental data have been obtained with Inconel-625 tube test sections in the following conditions: diameter of 6, 8, 10 and 12 mm; heated length of 0.31.77 m; pressure of 106951 kPa; mass flux of 20277 kg m⁻² s⁻¹; and inlet subcooling of 50654 kJ kg⁻¹, thermodynamic equilibrium critical quality of 0.3231.251 and CHF of 1081598 kW m⁻². Flow regime analysis based on Mishima & Ishii’s flow regime map indicates that most of the CHF occurred due to liquid film dryout in annular-mist and annular flow regimes. Parametric trends are examined from two different points of view: fixed inlet conditions and fixed exit conditions. The parametric trends are generally consistent with previous understandings except for the complex effects of system pressure and tube diameter. Finally, several prediction models are assessed with the measured data; the typical mechanistic liquid film dryout model and empirical correlations of (Shah, M.M., 1987. Heat Fluid Flow 8 (4), 326–335; Baek, W.P., Kim, H.G., Chang, S.H., 1997. KAIST critical heat flux correlation for water flow in vertical round tubes, NUTHOS-5, Paper No. AA5) show good predictions. The measured CHF data are listed in Appendix B for future reference.