The effect of subcooled water parameters on thermal-hydraulic characteristics for VVER reactor; Thermal Science; Vol. 29, iss.1

The effect of subcooled water parameters on thermal-hydraulic characteristics for VVER reactor; Thermal Science; Vol. 29, iss.1

Bibliographic Details
Parent link:Thermal Science.— .— Belgrade: Laboratory for Thermal Engineering and Energy
Vol. 29, iss.1.— 2025.— P. 27-34
Main Author: Korotkikh A. G. Aleksandr Gennadievich
Other Authors: Odii Ch. J. Christopher Joseph
Summary:Title screen
Prediction of critical heat flux using empirical correlations for circular tubes modified for rod bundles with correction factors, is one of established method of critical heat flux evaluation. In this work, an analysis of the thermal-hydraulics of VVER heated core was carried out. A geometrical and thermal analysis of the heated core, including the analysis of the flow rate and mass flux for the assemblies with hot water, analysis of heat transfer densities of the hottest part of the core and proper thermo-fluid analysis of the coolant parameters such as enthalpy, temperature and steam equilibrium quality were carried out. The thermal hydraulic analyses were carried out at the pressures, inlet temperatures and thermal powers of 16.2 MPa, 298.2°C, and 3200 MWth, 15.7 MPa, 298.2°C, and 3000 MWth and 12.5 MPa, 262°C, and 1375 MWth to ascertain the axial changes in thermal parameters of the fuel rod. The critical heat flux was predicted using the OKB Gidropress and Levitan-Lantsman critical heat flux correlations for rod bundles under the ranges of parameters suitable for VVER reactor
Текстовый файл
AM_Agreement
Language:English
Published: 2025
Subjects:
электронный ресурс
труды учёных ТПУ
thermal
hydraulics
water coolant
fuel rod
nuclear reactor
heat flux
temperature
pressure
Online Access:https://doi.org/10.2298/TSCI240330151K
Format: Electronic Book Chapter
KOHA link:https://koha.lib.tpu.ru/cgi-bin/koha/opac-detail.pl?biblionumber=679805

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