Wyniki wyszukiwania - BazTech

Ograniczanie wyników

2 2015

Znaleziono wyników: 2

Liczba wyników na stronie

Wyniki wyszukiwania

Wyszukiwano:
w słowach kluczowych: thermal hydraulics

Sortuj według:

Ogranicz wyniki do:

Neutronic and thermal-hydraulic coupling for 3D reactor core modeling combining MCB and fluent

Królikowski I. P., Cetnar J.

Nukleonika

2015

Vol. 60, No. 3, part 2

531--536

Three-dimensional simulations of neutronics and thermal hydraulics of nuclear reactors are a tool used to design nuclear reactors. The coupling of MCB and FLUENT is presented, MCB allows to simulate neutronics, whereas FLUENT is computational fluid dynamics (CFD) code. The main purpose of the coupling is to exchange data such as temperature and power profi le between both codes. Temperature required as an input parameter for neutronics is significant since cross sections of nuclear reactions depend on temperature. Temperature may be calculated in thermal hydraulics, but this analysis needs as an input the power profile, which is a result from neutronic simulations. Exchange of data between both analyses is required to solve this problem. The coupling is a better solution compared to the assumption of estimated values of the temperatures or the power profiles; therefore the coupled analysis was created. This analysis includes single transient neutronic simulation and several steady-state thermal simulations. The power profi le is generated in defined points in time during the neutronic simulation for the thermal analysis to calculate temperature. The coupled simulation gives information about thermal behavior of the reactor, nuclear reactions in the core, and the fuel evolution in time. Results show that there is strong influence of neutronics on thermal hydraulics. This impact is stronger than the impact of thermal hydraulics on neutronics. Influence of the coupling on temperature and neutron multiplication factor is presented. The analysis has been performed for the ELECTRA reactor, which is lead-cooled fast reactor concept, where the coolant flow is generated only by natural convection.

Thermal – hydraulic analysis of fuel block in high temepera-ture reactor

Królikowski I., Kępisty G., Orliński M.

Logistyka

2015

nr 4

9278--9285, CD3

Thermal hydraulic analysis of the reactor core is important since it allows to optimize the nuclear reactor operation and to avoid too high temperature in the fuel. Enhancement of the reactor core increases the safety and the efficiency of the reactor operation and it has positive impact on the logistic in the nuclear sector. The thermal analysis of the fuel block column of the high temperature reactor is presented. The 3D power density profile has been used in the thermal calculations to obtain the temperature field within the column of the nine fuel blocks. The hot spot for the critical power profile is found. Temperature profiles obtained in the analysis have been compared with the reference data to check the numerical model, which has been used in the CFD calculations. Obtained temperatures are consistent with the reference data, it proves that the numerical model is correct.