Open AccessSub-channel thermal-hydraulic analysis for VVER- 1000 generation III PWR
Author(s) -
Alsherief M Almessallmy,
Mohamed Shaat,
Saeed A Hassanien
Publication year - 2020
Publication title -
iop conference series. materials science and engineering
Language(s) - English
Resource type - Journals
eISSN - 1757-899X
pISSN - 1757-8981
DOI - 10.1088/1757-899x/975/1/012019
Subject(s) - vver , coolant , nuclear engineering , scram , thermal hydraulics , transient (computer programming) , nuclear reactor core , decay heat , neutron flux , pressurized water reactor , mechanics , cladding (metalworking) , critical heat flux , nucleate boiling , heat flux , nuclear reactor , materials science , engineering , heat transfer , physics , nuclear physics , mechanical engineering , computer science , neutron , metallurgy , operating system
The sub-channel thermal-hydraulic analysis of a nuclear reactor is essential for assessing of its safety aspects. In this paper, the VVER-1000 has been selected as an example of the third generation reactors since it meets most of the international safety standards and because it has been taken as a base for designing the VVER-1200 which is belonging to the III+ generation. A steady state mathematical model has been proposed and solved to validate and assure that the hottest channel temperature limits are satisfied. The various temperature distributions, the critical heat flux and the departure from nucleate boiling ratio (DNBR) for the hottest channel were evaluated. Also, a transient state model has also been presented and solved using the finite difference method with the aid of MATLAB algorithm. An exponential loss of flow rate of the reactor core coolant was triggered from the steady state conditions. We assumed that the neutron flux and the generated power were unchanged during the postulated event. The average core coolant flow time constant was treated as a single parameter expressing the rapidity of the event. A value of 250 seconds time constant was assumed for slow transient, whereas 10 seconds was assumed for fast one. The reactor core was assumed to be protected through the reactor control system and mitigated according to the regular emergency operating procedures. The time dependent temperature distributions were calculated for the cladding of the hottest coolant channel. For each value of the temperature, the response time required for reaching unsafe conditions was evaluated, discussed and presented.