Core Burn-Up Analysis of the RSG-Gas Research Reactor Using Deterministic and Stochastic Methodss

Abstract

Due to several characteristics, such as geometry, compact core, high coolant flow, and high neutron flux, the burn-up study of the RSG-GAS multi-purpose reactor provides challenges when employing a neutronic calculation. For the burn-up analysis, two calculating methodologies are used in the RSG-GAS: deterministic and probabilistic methods. The deterministic codes such as WIMSD-5B and Batan-FUEL are utilized, whereas the continuous-energy Monte Carlo code Serpent 2 is used for the stochastic method. WIMSD-5B is being used to produce a four-group cross-section that is needed by Batan-FUEL to do full core diffusion calculations. Burn-up calculations were performed at the whole fuel assemblies inside the core to see if the deterministic code, WIMSD-5B/Batan-FUEL, could accurately replicate the burn-up behavior of the RSG-GAS research reactor. The Serpent 2 calculation was also done with the exact models to provide a comparison. The results show that both Serpent 2 and WIMSD-5B/Batan-FUEL can perform the RSG-GAS burn-up analysis if appropriate treatments are made to the deterministic codes at both the assembly and core levels. There is a 5% difference in calculated fuel burn-up between deterministic and stochastic approaches.

Effect of New Cross Section Data File on the Criticality of the RSG-GAS Research Reactor

“…Some of the safety parameters were also evaluated by appropriate analytical tools [7][8][9]. The RSG-GAS reactor produces an average thermal neutron flux of 2×10 14 neutron cm -2 s -1 at a nominal power of 30 MW [10]. Several irradiation facilities are provided for radioisotope production and experiments namely one central irradiation position (CIP), four irradiation positions (IP), five rabbit systems, a power ramp test facility, six beam tubes, and neutron transmutation doping.…”

“…For that reason, at operation cycle no. 88 burnup parameter has been measured and then compared to the calculations [4,14]. The burnup measurement results core no.…”

Effect of Neutronic Parameters on the Cladding of the AP1000 MOX Fuel by Calculations Using MCNP Code

“…To find out how the effect of the fuel cladding material on the neutronic parameters in the reactor core using MOX [4] fuel, it is necessary to calculate the core parameter using a computer code. The computer code used is MCNP [5,6] using the Monte Carlo method. The parameters analyzed were the value of the keff core and excess core reactivity.…”