Tabular method of calculating the critical heat flux in water-cooled triangular rod assemblies

Bobkov, V. P.; Vinogradov, V. N.; Kirillov, P. L.; Smogalev, I. P.

doi:10.1007/bf02673204

Cited by 7 publications

(4 citation statements)

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“…The method (Bobkov 1999) was used to calculate Q c for the parameters implemented in the "steady-state cri-sis" experiment. To obtain a satisfactory correlation between the calculated data and the experimental data, a multiplying factor of 1.29 was determined.…”

Section: Discussionmentioning

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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Section: Discussionmentioning

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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“…The following have been used in developing a method for PWR, BWR, RBMK, and other assemblies with a square arrangement of fuel elements: 1) experience in constructing tables for pipes, ring channels, and assemblies with a triangular arrangement of fuel elements [4,5] and 2) developments, presented in [7], where it is shown that critical heat fluxes in assemblies with triangular and square arrangements of fuel elements differ substantially from one another. Experimental data arrays for assemblies from the database at the Physics and Power Engineering Institute were used [8].…”

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“…The tables for pipes are the core in describing a crisis in channels and are the basis for developing tabular methods for describing and calculating a crisis in many channels with a complicated cross-section and in fuel-element assemblies. Subsequent works performed at the Physics and Power-Engineering Institute developed methods for tabular description and calculation of critical heat fluxes in ring [2] and other single channels [3], in triangular assemblies of cylindrical fuel elements [4,5], and in the assemblies with a mixed arrangement of fuel elements [6].Tabular methods for calculating critical heat fluxes in channels have substantial advantages over conventional methods and are widely used in practical calculations using approaches based on an analytical (in the form of formulas) description of experimental data. Tabular methods make it possible to describe a wider range of parameters, and are based on a wider experimental base, give a continual description of critical heat fluxes in the entire field of parameters, and are convenient for estimating the aggregate influence of the parameters and for monitoring the results of calculations.…”

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Table for Critical Heat Fluxes in Square-Packed Fuel-Element Assemblies

et al. 2005

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This paper is devoted to the construction of a tabular method for calculating critical heat fluxes in fuel-element assemblies with a square arrangement. The method consists of a basic table and a system of correctional functions, describing the dependence of the critical heat flux on the important independent parameters. Experience in developing such a method for triangular assemblies is used. The method is compared with experimental data and other computational methods. The advantages of the new method are shown: maximum width of the range of parameters, possibility of taking account of the influence of the most important parameters, possibility of using in cell-by-cell (channelwise) calculations, and high accuracy of the description.The publication in 1997 of the international version of the tables for critical heat fluxes in pipes [1] laid the foundation for the development of similar tables for critical heat fluxes in other channels. The tables for pipes are the core in describing a crisis in channels and are the basis for developing tabular methods for describing and calculating a crisis in many channels with a complicated cross-section and in fuel-element assemblies. Subsequent works performed at the Physics and Power-Engineering Institute developed methods for tabular description and calculation of critical heat fluxes in ring [2] and other single channels [3], in triangular assemblies of cylindrical fuel elements [4,5], and in the assemblies with a mixed arrangement of fuel elements [6].Tabular methods for calculating critical heat fluxes in channels have substantial advantages over conventional methods and are widely used in practical calculations using approaches based on an analytical (in the form of formulas) description of experimental data. Tabular methods make it possible to describe a wider range of parameters, and are based on a wider experimental base, give a continual description of critical heat fluxes in the entire field of parameters, and are convenient for estimating the aggregate influence of the parameters and for monitoring the results of calculations.The Tabular Method of the Physics and Power Engineering Institute for Assemblies with a Square Arrangement of Fuel Elements. The following have been used in developing a method for PWR, BWR, RBMK, and other assemblies with a square arrangement of fuel elements: 1) experience in constructing tables for pipes, ring channels, and assemblies with a triangular arrangement of fuel elements [4,5] and 2) developments, presented in [7], where it is shown that critical heat fluxes in assemblies with triangular and square arrangements of fuel elements differ substantially from one another. Experimental data arrays for assemblies from the database at the Physics and Power Engineering Institute were used [8]. These data are published in [9][10][11][12][13][14][15][16][17]. Unfortunately, the extensive data of Columbia University [18-20] are inaccessible. The following parameters characterize the array used in the present paper: the number of experimenta...

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Comparative analysis of two tabular methods for calculating critical heat fluxes in rod assemblies

Bobkov¹

2000

At Energy

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Tabular method of calculating the critical heat flux in water-cooled triangular rod assemblies

Cited by 7 publications

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

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

Table for Critical Heat Fluxes in Square-Packed Fuel-Element Assemblies

Comparative analysis of two tabular methods for calculating critical heat fluxes in rod assemblies

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