Transient analysis for liquid-fuel molten salt reactor based on MOREL2.0 code

Abstract

SummaryA molten salt reactor (MSR) is characterized by simultaneously using liquid fuel salt as both the nuclear fuel and coolant. The redistribution of delayed neutron precursors (DNPs) makes the transient behavior of MSRs different from traditional solid-fuel reactors. In this study, a 3D coupled neutronics/thermal hydraulics code, MOREL2.0, was employed to analyze a liquid-fuel Thorium Molten Salt Reactor (TMSR-LF) under perturbations of fuel pump start-up and coast-down and by overheating and overcooling the inlet fuel temperature. Some transient processes were simulated to provide guidance for the future design and optimization of TMSR-LFs. In response to the perturbations, reactivity was lost and gained in the pump start-up and coast-down, respectively.Overheating the inlet fuel temperature introduced negative reactivity, and TMSR-LF maintained a safety state, while overcooling the inlet fuel temperature resulted in positive reactivity. Overcooling by 70 K produced a supercritical transient condition and a rapid increase in power within a short period, which was followed by a decrease in power due to negative temperature feedback. The transient results demonstrate that the negative temperature feedback coefficients guarantee TMSR-LF inherent safety and the variation range of temperature stay within the safety margin.

Functional Requirements for the Modeling and Simulation of Advanced (Non-LWR) Reactor Mechanistic Source Term

“…Other efforts have also completed safety assessments using methodologies published from IAEA's International Project on Innovative Nuclear Power Reactor and Fuel Cycles (INPRO) [30]. Published safety analyses of MSRs using an assortment of nuclear codes, often coupled with thermal hydraulic codes, are available in refs [31][32][33][34][35][36][37][38][39]. Some of these efforts have shown that source term functionality can be added to these codes by tracking specific fission product isotopes with the flow [40][41][42][43].…”

Survey and Assessment of Computational Capabilities for Advanced (Non-LWR) Reactor Mechanistic Source Term Analysis.

“…The code is validated with MSRE data, proves to be of similar accuracy as the previous version, and has increased speedups for both steady-state and transient scenarios by considerable amounts. They later modeled the power and temperature as a function of time for scenarios such as a pump driven transient, an unprotected fuel pump startup and coast-down at zero and normal power, an inlet fuel temperature driven transient, and overheating and overcooling of inlet fuel [141]. This work shows the importance of system modeling for source term modeling, as time and space dependent temperature modeling during transient scenarios has a very important and direct impact on radionuclide speciation, transport, and many other phenomena that affect its evolution during a severe accident.…”