Experimental investigation of heat transfer from a 2×2 rod bundle to supercritical pressure water

SUMMARYIn the fuel assemblies in reactor core, the distribution of temperature is not uniform in the fuel rod along the circumferential direction. The circumferential temperature difference could be neglected for conventional assembly design but increases significantly in a tight lattice.To investigate the circumferential non-uniformity of heat transfer in rod bundles, this paper firstly establishes mathematical model and makes theoretical analyses, based on which influential non-dimensional numbers are summarized. Then, a semi-empirical correlation describing this phenomenon is proposed based on theoretical results and CFD sensitivity analyses. Finally, a three-dimensional model for cylindrical fuel is introduced in the thermal-hydraulic code to predict circumferential heat transfer non-uniformity in different geometry and flow conditions. To validate this model, some results are compared with experimental data of bundle test for subcritical and supercritical water.The three-dimensional fuel model for sub-channel code is a new numerical tool to predict the peak cladding and fuel temperature and can be applied in thermal-hydraulic analysis of fuel assemblies of tight lattice design.

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“…Figure 9. Test section and configuration of thermocouples [23] Circumferentially non-uniform fuel model T. Yang, X. Liu and X. Cheng…”

Section: Discussionmentioning

confidence: 99%

“…Comparison between different heat transfer correlations is reviewed in [33]. (2) Similar experiments are carried out for water-cooled 2 × 2 square bundle in reference [23]. The test facility is quite similar to [32] except for the round-corner assembly box, as shown in Figure 9.…”

Section: Rementioning

confidence: 99%

A circumferentially non-uniform fuel model and its application to thermal-hydraulic code

Yang

Liu

2017

Int J Energy Res

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“…Figure 6 shows the variation of HTC and T w vs. T b for leading correlations against 2 sets of data collected at similar flow conditions for 2 × 2 rod bundles. Figure 6 also compares the experimental heat transfer in a plain 2 × 2 rod bundle from the data by Wang et al [11] and the corresponding experimental heat transfer coefficient in a wire wrapped 2 × 2 rod bundle from the data by Wang et al [14]. Although more information is needed about the location of the wire wrap along the heated length, one observation can be reported here about the wire wrap effect on SCHT in the 2 × 2 rod bundle, which is the local increase in HTC just upstream of T pc .…”

Section: Rod Bundle Databasesmentioning

confidence: 98%

“…The correlation is applicable to G < 1200 kg m −2 s −1 and fluid temperatures 70-140°C. Wang et al [11] conducted an experimental study of SCHT to water flowing vertically upward through a 2 × 2 rod bundle; the 4 rods were inserted into a square channel with rounded corners. They assessed 8 correlations and reoptimized the numerical coefficients and dependencies of Jackson's [33] correlation for best fit of the collected rod bundle data.…”

Section: Rod Bundle Correlationsmentioning

confidence: 99%

“…Several reviews of convective heat transfer correlations at SC pressure have been published, for example, overviews and assessments of supercritical heat transfer (SCHT) correlations against both SC water and SC CO 2 data for tube [2][3][4][5][6][7], an assessment of correlations [8][9][10], and an updated assessment of correlations based on rod bundle water data [11].…”

Section: Introductionmentioning

confidence: 99%

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General Assessment of Convection Heat Transfer Correlations for Multiple Geometries and Fluids at Supercritical Pressure

Zahlan¹,

Leung²,

Huang³

et al. 2017

CNL Nuclear Review

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Calculation of the Heat Emission from Rod Bundles to Supercritical Pressure Steam

Kirillov

2015

At Energy

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The maximum temperature of fuel-element cladding in VVER-SKD reactors, where a triangular lattice with relative spacing x = 1.1 is used, can be determined using known relations, since the Prandtl number Pr ~ 1 for superheated steam at supercritical pressure.A large number of computational and experimental studies are devoted to the determination of the coeffi cient of heat transfer to water fl ow at supercritical pressure in round tubes [1,2]. A particularity of heat transfer at such pressures is a complex nonlinear temperature dependence of the thermophysical properties of water. The properties of water vary most sharply at pseudocritical temperature 658.05 K [3]. The temperature range of such variation of the thermophysical properties of water is defi ned as (T pc -ΔT) < T < (T pc -ΔT) and at pressures P < 26 MPa equals about 3-5 K ( Fig. 1). Three regions of the states of water at supercritical pressure are singled out:• high coolant density (liquid-like state); • near-critical or near-pseudocritical region near T pc ; and • low coolant density (gaseous state) [4]. One of the main thermohydraulic characteristics is the maximum admissible temperature of the fuel-element cladding. The VVER-SKD fuel assemblies comprise bundles of fuel elements in the form of 10.5 mm in diameter round rods with 0.55 mm thick walls, arranged in a triangular packing with relative spacing x = s/d ~ 1.1 [5]. In most cases, the heat-emission coeffi cient in VVER-SKD rod bundles is determined according to the correlations developed for water fl ow at supercritical pressure in round tubes [1, 2, 6]. These correlations do not take account of all particularities of heat exchange in fuel assemblies and can be used only for the normal (degradation-free) heat exchange [7].Two-way motion of water is used in the concept developed for a fast-resonance VVER-SKD; it improves the conditions for cooling the fuel elements as a result of the higher coolant fl ow velocity. Mixing of the coolant in the bottom pressure tank reduces the nonuniformity of the coolant heating at the exit from a fuel assembly in the central part of VVER-SKD. This secures the required coolant temperature distribution over the height of the core at pressure 25 MPa, specifi cally, at the entrance into the reactor 553 K, at the exit 813 K, and in the bottom mixing chamber 663 K (Fig. 2). As a result of this design, the fuel-element cladding temperature can be lowered to an acceptable value. Another important advantage of the two-way scheme is that the region of pseudocritical coolant temperature is displaced into the top interior part of the core, which is characterized by relatively small heat fl uxes [8,9].Taking account of the nonuniform distribution of the energy release (heat fl ux density) over the height of the interior part of the core, the region of the highest temperature of the fuel-element cladding in VVER-SKD will be located in the top third of its height. The highest temperature of coolant in the form of superheated water vapor in this zone (close to the exit from t...

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Experimental investigation of heat transfer from a 2×2 rod bundle to supercritical pressure water

Cited by 70 publications

References 10 publications

A circumferentially non-uniform fuel model and its application to thermal-hydraulic code

A circumferentially non-uniform fuel model and its application to thermal-hydraulic code

General Assessment of Convection Heat Transfer Correlations for Multiple Geometries and Fluids at Supercritical Pressure

Calculation of the Heat Emission from Rod Bundles to Supercritical Pressure Steam

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