Analysis of VVER-1200 thermal characteristics using analytical calculation validated by PCTran simulator; Discover Mechanical Engineering; Vol. 4

Analysis of VVER-1200 thermal characteristics using analytical calculation validated by PCTran simulator; Discover Mechanical Engineering; Vol. 4

Մատենագիտական մանրամասներ
Parent link:Discover Mechanical Engineering.— .— Berlin: Springer Nature
Vol. 4.— 2026.— Article number 79, 21 p.
Հիմնական հեղինակ: Korotkikh A. G. Aleksandr Gennadievich
Այլ հեղինակներ: Odii Ch. J. Christopher Joseph
Ամփոփում:Title screen
An analytical method was developed to predict the thermal parameters of the VVER-1200 pressurized water reactor (PWR). The peak temperatures of the fuel, outer fuel surface temperature, inner cladding surface temperature, outer cladding surface temperature, and coolant temperature were evaluated along and across the fuel rod, assuming that the hottest fuel rod is located at the reactor core center. This method analyzed the steady-state temperature distributions and was validated against a corresponding steady-state and time-dependent COMSOL Multiphysics model. Additionally, validation was conducted using the VVER-1200 PCtran Simulator for peak fuel temperature and the departure from nucleate boiling ratio (DNBR). The analytical method was employed to increase the reactor’s rated power from 100% to approximately 106% to assess the peak fuel temperature and the minimum departure from nucleate boiling ratio (MDNBR), both of which are crucial safety parameters for the operation of the VVER-1200. Given that the VVER-1200 initiates a scram slightly above 106% rated power, we evaluated the peak fuel temperature and DNBR at 100%, 102%, 104%, and 106% rated powers to determine the safety limits for both peak fuel temperature and MDNBR. The results from both the analytical method and simulations were found to be comparable
Текстовый файл
AM_Agreement
Լեզու:անգլերեն
Հրապարակվել է: 2026
Խորագրեր:
электронный ресурс
труды учёных ТПУ
Thermal power
Fuel rod
Temperature
Pressurized water reactor
Departure from nucleate boiling ratio
Critical heat flux
Առցանց հասանելիություն:https://doi.org/10.1007/s44245-025-00175-7
Ձևաչափ: Էլեկտրոնային Գրքի գլուխ
KOHA link:https://koha.lib.tpu.ru/cgi-bin/koha/opac-detail.pl?biblionumber=684611

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