Dynamic Analysis on the Safety Criteria of the Conceptual Core Design in MTR-type Research Reactor

Abstract

One of the high-priority research activities in BATAN is designing a new MTRtype research reactor with a new fuel. The core follows a compact core model that consists of 16 fuels and 4 control rods. The increasing heat flux at the fuel will cause the temperature of the fuel and cladding to increase so that the coolant flow rate needs to be increased. However, the coolant flow rate in the fuel element is limited by the thermal-hydraulic stability in the core. Therefore, dynamic analysis is important in evaluating the design and operation of nuclear reactor safety. The objective of this research work is to carry out a dynamic analysis for a conceptual MTR research reactor core fuelled with the low-enrichment U9Mo-Al dispersion. The calculations were performed using WIMSD-5B, Batan-2DIFF, Batan-3DIFF, POKDYN, and MTRDYN codes. Steady-state thermal-hydraulic parameters and dynamic analysis were determined using the MTRDYN code. The calculation results show that the maximum temperatures of the coolant, cladding, and fuel meat with the uranium density of 3.96 g cm-3 are 76.01 °C, 192.02 °C, and 196.24 °C, respectively. The maximum value of fuel meat temperature for safety limit is 210 °C, which means that the maximum temperatures fulfill the design limit, and therefore the reactor operates safely at the nominal power. The dynamic analysis shows that inherent safety can protect the reactor operation when insertion of reactivity occurs in the core.

Transient calculation on TRIGA 2000 of plate-type fuel element using RELAP5, EUREKA2/RR, and MTR-DYN codes

“…Ther mal hy drau lic anal y sis con ducted us ing sev eral com puter codes is of ten used to eval u ate the safety of re ac tor oper a tion. To achieve this pur pose, a cal cu la tion of RIA and LOFA sce nar ios by us ing some nu mer i cal sim u lation codes is nec es sary [11,12]. Var i ous com puter codes have been fre quently used to cal cu late core param e ters dur ing steady-state and tran sient op er a tion ei ther in re search re ac tors or nu clear power plants.…”

Analyses of Neutron Absorber Materials on the Safety Parameters in the RSG-Gas Reactor

“…Calculation of changes in control rod reactivity and their effects on operating safety must be predicted and compared with calculated results aculately or measured parameters to ensure that there is sufficient reactivity margin at all times. So that, it can be ascertained that the reactor can be shutdown safely [6,7]. The RSG-GAS reactor core shutdown system currently uses AgInCd neutron absorber material.…”

Transient Analysis of Simultaneous Lofa and Ria in RSG-Gas Reactor After 32 Years Operation

“…The importance of LOFA and RIA analysis are for safety precaution during operation of research reactor as well as nuclear power plant. Several researches related to flow analysis, LOFA or reactivity in research reactor and nuclear power plant that supporting to this research such as preliminary accident analysis for a conceptual design a 10 MW research reactor [12], dynamic analysis for conceptual core design [14], thermal hydraulic analysis improvement for IEA-R1 reactor [15], characterization of oxide fuel element of RSG-GAS [16], accident safety analysis in JRR-3M [17], analysis of temperature effect on control rod worth in TRR [18], and transient analysis in a downward cooling pool-type material testing reactor [19] reactivity feedback effect on LOFA in PWR [20] and validation using standard code on reactor transient condition [21].…”