Neutronic, Kinetic, and Thermal-Hydraulic Calculation of Accelerator Driven Target-blanket; Cross-section Libraries Testing

article in a collection out of WoS and Scopus

English

Calculation and computer modelling is very important for the search and evaluation of physical characteristics of every new reactor concept, such as accelerator driven system. Some neutronic calculation of ADS model blanket based on fluoride salts of actinides and fission products was performed (neutron flux parameters, multiplication factor dependencies etc.). Thermal-hydraulic analyses of the ADS reactor core with fluid fuel, and kinetic calculation of subcritical reactor with external neutron source and fluid fuel (which is one of the issues coupled with accelerator driven systems important for safety studies) was also proceeded. Nuclear data and various cross-section libraries influence to multiplication factor of ADS model blanket was separately studied. Neutronic calculation was computed by general Monte Carlo transport code MCNP. To study nuclear data influence is necessary to convert libraries from ENDF format to ACE format for MCNP, code NJOY was used to do that. Main world-widely used cross-section libraries were tested, such us ENDF/B, JEF, JENDL, BROND, CENDL and also available high energy libraries. Effects of various code versions (MCNP and NJOY) were studied too. For kinetics study of an ADS blanket with external neutron source and with fluid fuel based on fluorine salts were used point-kinetics equations modified by leakage of delay neutron precursor. For numerical solution of this equations was created code Bokin 2000 which was next applied to several selected transients of subcritical reactor system. From preliminary calculations can be seen that using molten fluoride salts acting as fuel and coolant simultaneously can cause a new type of transient effect during the fuel pump failure. The slowdown of the fuel flow decreases the leakage of delayed neutrons and thus β increases. However, the response of the system is mostly determined by the value of thermal feedback coefficient. Preliminary calculations of a radial power density distribution in different modifications of blanket have been done up today. To calculate a solution to the Navier-Stokes equations for liquids, with volume changes as functions of the temperature and movable heat sources, modern computer techniques and fitting simulation codes, which are based on suitable numerical methods are needed. They are called the CFD computer programmes and two of them were used: PHOENICS 3.2.0 based on the finite volume method, and a module FlowPlus for the programme package COSMOS 2.5, which is based on the finite element method.

ADS; ADTT; Accelerator Driven Systems; Nuclear Data; TH; Thermal Hydraulics; Neutronics

KATOVSKÝ, K.; KOBYLKA, D.; KŘEPEL, J.

2. 9. 2002

Accelerator Applications Division, American Nuclear Society

555 North Kensington Avenue, La Grange Park, IL 60526, United States

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