Abstract

The Molten Salt Reactor (MSR) is the most important system suggested by Generation IV for the future direction in the nuclear reactor field. For more development of the MSR reactor, the core system inside the tube is proposed by naturally circulating molten fuel salt. The nonlinear kinetic equations form a linearized function and are obtained in state-space form. Reactivity feedback and delayed neutrons are extremely important for reactor control. In this paper, a thermal-hydraulic system for the commercial computation dynamic model is proposed. Currently, there is no commercial software to simulate the natural circulation flow. The proposed method can be easily employed to detect faults and can provide a feasible overall system performance.

Highlights

  • The primary loop of Molten Salt Reactor (MSR) is simulated under the steady state in Matlab/Simulink

  • Many dynamic methods considering the thermal MSR have been presented in previous works

  • It was assumed that the outlet core temperature is the same as the average temperature of the higher lumped region

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Summary

INTRODUCTION

The reactor control mechanism or trip primary circuit may allow or produce fuel salt in the primary circuit, and circulation of the delayed-neutron precursor is generated in the core. MSRs are associated with solid fuel which is used in traditional reactors. The fuel salt is always held on the outer side of the MSR core. In [5, 6], the fuel region is modeled as a double lasting region, which is very important in order to improve the state space equation for the outlet core temperature. The outlet core temperature is measured more accurately. The stateaction space equation is used for the reactor, thermal-hydraulic modeling, energy balance, and controls all the volume of the core and the heat exchanger (HX) unit [8]. We explore and propose a novel thermalhydraulic method which give us more accurate and confident results without the ambiguity of the previous models

THE THERMAL-HYDRAULIC METHOD
MATHEMATICAL DERIVATION AND REACTOR NEUTRONS
Neutronic Calculation Model Development
RESULTS AND DISCUSSION
CONCLUSION
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