An Improved Steady-State and Transient Analysis of the RSG-GAS Reactor Core under RIA Conditions Using MTR-DYN and EUREKA-2/RR Codes

Abstract

Steady-state and transient analysis of reactor core under Reactivity-Initiated Accident (RIA) conditions are important for reactor operation safety. The reactor dynamics are influenced by neutronic and thermal-hydraulic aspects of the core. In this study, steady-state and transient analysis under RIA conditions of the RSG-GAS multipurpose reactor was carried out using MTR-DYN and EUREKA-2/RR programs. Neutronic calculations were performed using a few group cross-sections generated by Serpent 2 with the latest cross-section data ENDF/B-VIII.0. Steady-state conditions were carried out with a nominal power of 30 MW, while transient under RIA conditions occurred because the control rod was pulled too quickly while the reactor operated. These transient RIA conditions were performed for two cases, during start-up with an initial power of 1 W, and within power range with an initial power of 1 MW. Thermal-hydraulic parameters considered in this study are reactor power, the temperature of the fuel, cladding, and coolant. The calculated maximum fuel temperature at a steady state is 126.02°C. Meanwhile, the calculated maximum fuel temperature during RIA conditions at the initial power of 1 W and 1 MW are 64.38°C and 137.14°C, respectively. There are no significant differences in thermal-hydraulic parameters between each used program. The thermal-hydraulic parameters such as the maximum temperature of the coolant, cladding, and fuel under this postulated RIA condition are within the acceptable reactor operation safety limits.

An Efficient and Robust ILU(k) Preconditioner for Steady-State Neutron Diffusion Problem Based on MOOSE

“…The detailed implementation can be found in Ref [23]. In MOOSE, the nonlinear PDE equation Equation ( 5) is solved by the preconditioned JFNK algorithm, as shown in Equation (6). The JFNK method is a powerful nonlinear solver featured with two iteration layers, including the Newton iteration as the outer layer and the Krylov subspace iteration as the inner layer.…”

“…[23]. In MOOSE, the nonlinear PDE equation Equation ( 5) is solved by the preconditioned JFNK algorithm, as shown in Equation (6).…”

“…After the discretization of these nonlinear PDEs, it usually leads to a large-scale nonlinear algebraic equation problem [3]. Compared with the traditional methods, such as the operator splitting method and Picard iteration method [4][5][6], the Jacobian-free Newton-Krylov (JFNK) algorithm [7] is a powerful solver for the nonlinear equations due to its high-order convergence rate. The JFNK algorithm has been widely used in the newly developed nuclear reactor simulator [8][9][10][11], such as the MOOSE platform [12] and LIME platform [13,14].…”

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

“…The RSG-GAS core configuration [18] Figure 2. The RSG GAS standard fuel element [7] Figure 3. The RSG GAS controls element with absorber blades (7).…”

“…In order to assure that the reactor is safely operated, at the beginning of the operation cycle (BOC), some important nuclear parameters are measured such as excess reactivity and control rod worth [3][4][5][6]. 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].…”