A modified table for calculating critical heat fluxes in assemblies of triangularly packed fuel rods

Bobkov, V. P.; Efanov, A. D.; Pomet'ko, R. S.; Smogalev, I. P.

doi:10.1134/s0040601511040045

Cited by 5 publications

(20 citation statements)

References 2 publications

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“…Its first version, the 1997 Bobkov LUT, comprises more than 4000 data points obtained from experiments on triangular assemblies performed at the Gidropress Experimental Design Office (OKB Gidropress), the State Scientific Center of the Russian Federation, the I.I. Leypunsky Institute Of Physics And Power Engineering (IPPE), as well as other organizations [8]. The 2011 version of the Bobkov LUT is improved over the 1997 version in two aspects:…”

Section: Bobkov Look-up Tablementioning

confidence: 99%

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A Methodology for CHF Prediction in VVER Rod Bundles

AbdulHameed,

Shaaban,

Gamal

et al. 2022

Preprint

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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 VVER 1 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.

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Section: Bobkov Look-up Tablementioning

confidence: 99%

“…3 if the heated length is taken into account. Examples of this type include the 2006 Groeneveld LUT [7], the 2011 Bobkov LUT [8], and the Biasi correlation [9]. 2.…”

Section: Introductionmentioning

confidence: 99%

A Methodology for CHF Prediction in VVER Rod Bundles

AbdulHameed,

Shaaban,

Gamal

et al. 2022

Preprint

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“…Owing to numerous experimental data, it became possible to describe, in more or less optimal form, heat fluxes in assemblies with different rod placement arrangements. The basic tables and sys tems of correction functions for different assemblies are presented in [8][9][10][11][12]. The formula for calculating CHF with the appropriate corrections for triangular assemblies looks as follows: q tr.ass = q tab F dh F Sd F L F kf F hh F f F nr , where q tab is taken from the basic tables developed for the basic structures [10], F dh = [d h /(9.36)] -1/3 is the correction function for the thermal diameter of each cell in a particular assembly (see Fig.…”

Section: Burnout In Eccentric Annular Channels the Notion Of Elementmentioning

confidence: 99%

“…The basic tables and sys tems of correction functions for different assemblies are presented in [8][9][10][11][12]. The formula for calculating CHF with the appropriate corrections for triangular assemblies looks as follows: q tr.ass = q tab F dh F Sd F L F kf F hh F f F nr , where q tab is taken from the basic tables developed for the basic structures [10], F dh = [d h /(9.36)] -1/3 is the correction function for the thermal diameter of each cell in a particular assembly (see Fig. 3); F Sd is the cor rection function for the relative pitch between the rods, at S/d ≤ 1.1 (in the given case, the parameter S/d takes into account the curvature, stretched shape, and nonuniformity of elementary cells composing the bundle macrocell), F Sd = 0.82-0.…”

Section: Burnout In Eccentric Annular Channels the Notion Of Elementmentioning

confidence: 99%

Critical heat flux for water boiling in channels. Modern state, typical regularities, unsolved problems, and ways for solving them (a review)

Bobkov

2015

Therm. Eng.

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Some general matters concerned with description of burnout in channels are outlined. Data obtained from experimental investigations on critical heat fluxes (CHF) in different channels, CHF data banks, the main determining parameters, CHF basic dependences, and a system of correction functions are discussed. Two methods for estimating the CHF description errors are analyzed. The influence of operating parameters, transverse sizes of channels, and conditions at their inlet are analyzed. The effects of heat trans fer surface shape and heat supply arrangement are considered for concentric annular channels. The notions of a thermal boundary layer and an elementary thermal cell during burnout in channels with an intricate cross section are defined. New notions for describing CHF in rod assemblies are introduced: bundle effect, thermal misalignment, assembly section averaged and local parameters (for an elementary cell), cell wise CHF analysis in bundles, and standard and nonstandard cells. Possible influence of wall thermophysical properties on CHF in dense assemblies and other effects are considered. Thermal interaction of nonequivalent cells and the effect of heat supply arrangement over the cell perimeter are analyzed. Special attention is paid to descrip tion of the effect the heat release nonuniformity along the channels has on CHF. Objectives to be pursued by studies of CHF in channels of different cross section shapes are formulated.

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“…In (Bobkov et al 2011), a method for calculating Q c using a table for bundles is described: this region was not marked as experimentally justified and the data were obtained by extrapolation. To date, several papers have been published, the parameters of which do not correspond to the above range of changes in the coolant mass flow rate.…”

Section: Introductionmentioning

confidence: 99%

Outcomes of the “steady-state crisis” experiment in the MIR reactor channel

Alekseev

Dreganov

Izhutov

et al. 2019

NUCET

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To license nuclear fuel for nuclear power plants, data on the behavior of fuel elements (FE) under design-basis accidents are required. These data are obtained during tests of fuel assemblies (FA) and single fuel elements in research reactor channels followed by post-test studies in protective chambers. A reactivity-initiated accident (RIA) with an unauthorized release of CPS rods from the reactor core leads to a pulsed channel power increase. This accident can proceed according to two scenarios: without a critical heat flux (CHF) on the fuel element jacket at the final stage and with a dry heat flux. To date, a series of experiments have been carried out according to the first scenario in the MIR reactor channel and the corresponding data on the behavior of fuel elements have been obtained. An urgent task for today is to prepare and conduct reactor experiments according to the second scenario. The main experimental parameter that determines the behavior and final state of the studied fuel elements is their temperature. No experimental data were found on the critical heat flux for the rod bundles in the low coolant mass flow rate region (experiments in the MIR reactor channel can be conducted in the range of 200–250 kg/(m2s)). The available data are in the extrapolation range. The “steady-state crisis” experiment was conducted to obtain data on the critical heat flux value within the specified coolant mass flow rate range in the MIR reactor channel. The test object was a jacket fuel assembly composed of three shortened VVER-1000 fuel rods with a length of 1230 mm (the fuel part length = 1000 mm) installed in a triangular grid at a pitch of 12.75 mm, which is a cell of the VVER-1000 core. This assembly configuration is used for in-pile tests to study the behavior of fuel elements under emergency conditions. The in-pile testing results are presented. The paper shows the possibility of detecting the start and development of a dry heat flux based on the readings of thermocouples located inside the FE kernel. As a result, the directly measured test parameters were used to determine the critical heat flux value.

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A modified table for calculating critical heat fluxes in assemblies of triangularly packed fuel rods

Cited by 5 publications

References 2 publications

A Methodology for CHF Prediction in VVER Rod Bundles

A Methodology for CHF Prediction in VVER Rod Bundles

Critical heat flux for water boiling in channels. Modern state, typical regularities, unsolved problems, and ways for solving them (a review)

Outcomes of the “steady-state crisis” experiment in the MIR reactor channel

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