Analysis of Heat Transfer Characteristics of Canadian SCWR Fuel Assembly Concept

2020 ◽  
Vol 6 (3) ◽  
Author(s):  
A. Nava Domínguez ◽  
Y. F. Rao ◽  
T. Beuthe
Keyword(s):  
Heat Transfer ◽  
Fuel Assembly ◽  
Heat Loss ◽  
Pressure Tube ◽  
Flow Tube ◽  
Code Model ◽  
Fuel Bundle ◽  
Tube Type ◽  
Gen Iv ◽  
Central Flow

Abstract Canada is participating in the Generation IV (Gen IV) International Forum with a main focus on the pressure-tube-type supercritical water-cooled reactor (SCWR) concept. The subchannel code ASSERT-PV modified for SCWR applications was used to design the SCWR fuel assembly, specifically the fuel bundle. Several assumptions were required to model the fuel assembly, including the perfect insulation of (i) the central flow tube (i.e., no heat transfer through the central tube) and (ii) the pressure tube (i.e., no heat loss to the moderator). These two assumptions were considered as conservative, but they were not analyzed or assessed for their validity or accuracy. ASSERT-PV was upgraded to model the heat loss to the moderator, and an external CATHENA system code model was coupled to ASSERT-PV to model the heat transfer to the central flow tube. This paper describes these additional heat transfer components, and presents an assessment of these two assumptions for their impact on the prediction of maximum fuel cladding temperature.

Further Details of a Numerical Analysis on the Thermal Hydraulic Effect of Wrapped Wire Spacers in Fuel Bundle

2020 ◽  
Vol 6 (3) ◽  
Author(s):  
Attila Kiss ◽  
Bence Mervay
Keyword(s):  
Heat Transfer ◽  
Fuel Assembly ◽  
High Performance ◽  
Transverse Velocity ◽  
Axial Direction ◽  
Optimal Choice ◽  
Strong Mixing ◽  
Fuel Rods ◽  
Ongoing Research ◽  
Fuel Bundle

Abstract The application of relatively simple and cheap wrapped wire spacer in the European supercritical water-cooled reactor (SCWR) (high-performance light water reactor (HPLWR)) has been proposed in order to provide enhanced heat transfer in the fuel assembly without unacceptable penalty in pressure loss. The wires cause twisting flow in the fuel assembly, which means the coolant not only flows straight in the axial direction but also has a significant transverse velocity component, and strong mixing between neighboring subchannels occurs. The aim of this ongoing research is to numerically investigate the effect of wrapped wire spacers on thermal hydraulics of the turbulent coolant flow and its heat transfer in a small bundle of four fuel rods. One bare and six-wired geometries with varying wire pitches (1–6 turn(s) of wires) have been studied. It was found that the wires generate significant amount of transverse velocity, decrease the wall temperature, and increase the heat transfer coefficient mostly in corner subchannels which were the hottest in bare geometry. Thus, the presence of wires enhances heat transfer where it is most needed. Temperature hot spots with moderate values have been identified on the cladding wall of fuel rods. Based on the results, a technically optimal choice of number of wire turns from thermal hydraulic sense has been proposed.

Simulation of load following mode of operation for a natural circulation pressure tube type BWR

Kerntechnik ◽  
2011 ◽  
Vol 76 (4) ◽  
pp. 237-243
Author(s):  
R. Kumar ◽  
A. J. Gaikwad ◽  
A. D. Contractor ◽  
A. Srivastava ◽  
H. G. Lele ◽  
...  
Keyword(s):  
Natural Circulation ◽  
Pressure Tube ◽  
Load Following ◽  
Mode Of Operation ◽  
Tube Type

Heat Transfer Correlation for Supercritical Carbon Dioxide Flowing in Vertical Bare Tubes

2013 ◽  
Author(s):  
Sahil Gupta ◽  
Eugene Saltanov ◽  
Igor Pioro
Keyword(s):  
Heat Transfer ◽  
Error Analysis ◽  
Nuclear Power ◽  
Nuclear Reactor ◽  
Large Set ◽  
Transfer Coefficients ◽  
Data Points ◽  
Institute Of Technology ◽  
Using Data ◽  
Gen Iv

Canada among many other countries is in pursuit of developing next generation (Generation IV) nuclear-reactor concepts. One of the main objectives of Generation-IV concepts is to achieve high thermal efficiencies (45–50%). It has been proposed to make use of SuperCritical Fluids (SCFs) as the heat-transfer medium in such Gen IV reactor design concepts such as SuperCritical Water-cooled Reactor (SCWR). An important aspect towards development of SCF applications in novel Gen IV Nuclear Power Plant (NPP) designs is to understand the thermodynamic behavior and prediction of Heat Transfer Coefficients (HTCs) at supercritical (SC) conditions. To calculate forced convection HTCs for simple geometries, a number of empirical 1-D correlations have been proposed using dimensional analysis. These 1-D HTC correlations are developed by applying data-fitting techniques to a model equation with dimensionless terms and can be used for rudimentary calculations. Using similar statistical techniques three correlations were proposed by Gupta et al. [1] for Heat Transfer (HT) in SCCO2. These SCCO2 correlations were developed at the University of Ontario Institute of Technology (Canada) by using a large set of experimental SCCO2 data (∼4,000 data-points) obtained at the Chalk River Laboratories (CRL) AECL. These correlations predict HTC values with an accuracy of ±30% and wall temperatures with an accuracy of ±20% for the analyzed dataset. Since these correlations were developed using data from a single source - CRL (AECL), they can be limited in their range of applicability. To investigate the tangible applicability of these SCCO2 correlations it was imperative to perform a thorough error analysis by checking their results against a set of independent SCCO2 tube data. In this paper SCCO2 data are compiled from various sources and within various experimental flow conditions. HTC and wall-temperature values for these data points are calculated using updated correlations presented in [1] and compared to the experimental values. Error analysis is then shown for these datasets to obtain a sense of the applicability of these updated SCCO2 correlations.

Skinfolds and resting heat loss in cold air and water: temperature equivalence

1975 ◽  
Vol 39 (1) ◽  
pp. 93-102 ◽  
Author(s):  
R. M. Smith ◽  
J. M. Hanna
Keyword(s):  
Heat Transfer ◽  
Water Temperature ◽  
Heat Loss ◽  
Air Temperature ◽  
Indirect Calorimetry ◽  
Cold Water ◽  
Heat Conductance ◽  
Air Temperatures ◽  
Body Core ◽  
Male Subjects

Fourteen male subjects with unweighted mean skinfolds (MSF) of 10.23 mm underwent several 3-h exposures to cold water and air of similar velocities in order to compare by indirect calorimetry the rate of heat loss in water and air. Measurements of heat loss (excluding the head) at each air temperature (Ta = 25, 20, 10 degrees C) and water temperature (Tw = 29–33 degrees C) were used in a linear approximation of overall heat transfer from body core (Tre) to air or water. We found the lower critical air and water temperatures to fall as a negative linear function of MSF. The slope of these lines was not significantly different in air and water with a mean of minus 0.237 degrees C/mm MSF. Overall heat conductance was 3.34 times greater in water. However, this value was not fixed but varied as an inverse curvilinear function of MSF. Thus, equivalent water-air temperatures also varied as a function of MSF. Between limits of 100–250% of resting heat loss the followingrelationships between MSF and equivalent water-air temperatures were found (see article).

THE COEFFICIENT OF HEAT TRANSFER FOR VERTICAL SURFACES IN STILL AIR

1937 ◽  
Vol 15a (7) ◽  
pp. 109-117
Author(s):  
R. Ruedy
Keyword(s):  
Heat Transfer ◽  
Natural Convection ◽  
Heat Loss ◽  
Temperature Difference ◽  
Plane Surface ◽  
Vertical Plane ◽  
Black Body ◽  
Average Temperature ◽  
Fourth Root

For a vertical plane surface in still air the coefficient of heat transfer, valid within the range of temperatures occurring in buildings, depends on the temperature and the height of the surface. If black body conditions are assumed for the heat lost by radiation, the coefficient is equal to 1.39, 1.50, 1.62, and 1.73 B.t.u. per sq. ft. per ° F. at 32°, 50°, 68°, and 86° F. respectively, the height of the heated surfaces being 100 cm. Convection is responsible for about one-third, and radiation, mainly in the region of 10 microns, for about two-thirds of the heat loss. Convection currents depend on the temperature difference, while radiation depends on the average temperature. When attempts are made to stop convection currents by placing obstacles across the surface, the loss of heat due to natural convection varies inversely as the fourth root of the height, providing that the nature of the flow of air remains unchanged.

Engineering Relation for Dependence of Threshold Stress Intensity Factor for Delayed Hydride Cracking on Crack Orientation

2008 ◽  
Author(s):  
Douglas A. Scarth ◽  
Gordon K. Shek ◽  
Steven X. Xu
Keyword(s):  
Stress Intensity Factor ◽  
Stress Intensity ◽  
Intensity Factor ◽  
Threshold Stress ◽  
Pressure Tube ◽  
Threshold Stress Intensity ◽  
Tube Material ◽  
Fuel Bundle ◽  
Threshold Stress Intensity Factor ◽  
Cold Worked

Delayed Hydride Cracking (DHC) in cold-worked Zr-2.5 Nb pressure tubes is of interest to the CANDU industry in the context of the potential to initiate DHC at an in-service flaw. Examples of in-service flaws are fuel bundle scratches, crevice corrosion marks, fuel bundle bearing pad fretting flaws and debris fretting flaws. To date, experience with fretting flaws has been favourable, and crack growth from an in-service fretting flaw has not been detected. However, postulated DHC growth from these flaws can result in severe restrictions on the allowable number of reactor Heatup/Cooldown cycles prior to re-inspection of the flaw, and it is important to reduce any unnecessary conservatism in the evaluation of DHC from the flaw. One method to reduce conservatism is to take credit for the increase in the isothermal threshold stress intensity factor for DHC initiation at a crack, KIH, as the flaw orientation changes from an axial flaw to a circumferential flaw in the pressure tube. This increase in KIH is due to the texture of the pressure tube material. An engineering relation that provides the value of KIH as a function of the orientation of the flaw relative to the axial direction in the pressure tube has been developed as described in this paper. The engineering relation for KIH has been validated against results from DHC initiation experiments on unirradiated cold-worked Zr-2.5 Nb pressure tube material.

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