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Background <p>The main conditions for the safe operation of fuel assemblies (FAs) in the core of the IVV-2M heterogeneous water-water nuclear reactor of the pool type include the absence of surface boiling on the fuel element cladding under the layer of deposits. The values of the coolant flow rate through the FA can be used to predict its temperature at the FA outlet for the campaign and justify the operating limit settings.</p> Aim <p>To determine the analytical dependence of the relative quantitative change in the coolant flow rate through the FA on the pressure drop in the core based on the obtained primary empirical data from hydraulic tests on an FA model.</p> Materials and methods <p>A&#xa0;methodology for measuring the coolant flow rate through FAs is developed. Hydraulic tests of an FA model were performed to determine a&#xa0;formula for calculating the coolant flow rate through the FA in the core depending on its pressure drop.</p> Results <p>The developed methodology is appropriate for determining the coolant flow rate through the FA of the IVV-2M reactor. Measurements carried out on an FA model were used to obtain an empirical dependence of the change in the flow rate on the pressure drop in the core.</p> Conclusion <p>The considered methodology increases safety during reactor operation, as it allows the temperature at the outlet of the FA, and therefore, on the fuel element cladding to be predicted. The analytical formula of obtained dependencies can be used to calculate the coolant flow through the FA in the reactor core and to analyze the measurement results for different core configurations and coolant temperatures. The next article “Study of the coolant flow rate through the fuel assemblies of the IVV-2M reactor core. Part&#xa0;2” will represent a&#xa0;description of the flow meter design and an analysis of flow rate measurements over several years.</p>

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Study of the coolant flow rate through the fuel assemblies of the IVV-2M reactor core. Part 1

  • 期刊论文

  • 发表日期：

    2025年7月1日

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  • I. M. Russkikh,
  • E. N. Seleznev,
  • A. A. Zyryanova,
  • Yu. V. Volchikhina,
  • N. M. Aristov,
  • N. S. Kalashnikov,
  • A. V. Goryachikh,
  • O. A. Kravtsova,
  • O. L. Tashlykov

摘要

Background

The main conditions for the safe operation of fuel assemblies (FAs) in the core of the IVV-2M heterogeneous water-water nuclear reactor of the pool type include the absence of surface boiling on the fuel element cladding under the layer of deposits. The values of the coolant flow rate through the FA can be used to predict its temperature at the FA outlet for the campaign and justify the operating limit settings.

Aim

To determine the analytical dependence of the relative quantitative change in the coolant flow rate through the FA on the pressure drop in the core based on the obtained primary empirical data from hydraulic tests on an FA model.

Materials and methods

A methodology for measuring the coolant flow rate through FAs is developed. Hydraulic tests of an FA model were performed to determine a formula for calculating the coolant flow rate through the FA in the core depending on its pressure drop.

Results

The developed methodology is appropriate for determining the coolant flow rate through the FA of the IVV-2M reactor. Measurements carried out on an FA model were used to obtain an empirical dependence of the change in the flow rate on the pressure drop in the core.

Conclusion

The considered methodology increases safety during reactor operation, as it allows the temperature at the outlet of the FA, and therefore, on the fuel element cladding to be predicted. The analytical formula of obtained dependencies can be used to calculate the coolant flow through the FA in the reactor core and to analyze the measurement results for different core configurations and coolant temperatures. The next article “Study of the coolant flow rate through the fuel assemblies of the IVV-2M reactor core. Part 2” will represent a description of the flow meter design and an analysis of flow rate measurements over several years.