Coupled Thermal-Hydraulic Analysis and Species Mass Transport in a Versatile Experimental Salt Irradiation Loop (VESIL) Using CTF

With the resurgence of interest in molten salt reactors, there is a need for new experiments and modeling capabilities to characterize the unique phenomena present in this fluid fuel system. A Versatile Experimental Salt Irradiation Loop (VESIL) is currently under investigation at Idaho National Laboratory to be placed in the Advanced Test Reactor (ATR). One of the key phenomena this proposed experiment plans to elucidate is fission product speciation in the fuel-salt and the subsequent effects this has on the fuel-salt properties, source term generation, and corrosion control. Specifically, noble gases (Xe & Kr) will bubble out to a plenum or off-gas system, and noble metals (Mo, Tc, Te, etc.) will precipitate and deposit in specific zones in the loop. This work extends the mass transfer and species interaction models in CTF (Coolant-Boiling in Rod Arrays—Two Fluids) and applies these models to give a preliminary estimation of fission product behavior in the proposed VESIL design. A noble metal–helium bubble mass transfer model is coupled with the thermal-hydraulic results from CTF to determine the effectiveness of this insoluble fission product (IFP) extraction method for VESIL. Amounts of IFP species extracted to the off-gas system and species distributions in VESIL after a 60-day ATR cycle are reported.

The Development of a Thermal-Hydraulic and Nonlinear Dynamic System for Molten Salt Reactors

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.

A new method for monitoring the redox potential of fuel salt based on the deposition of 95Nb on Hastelloy C276

To develop the application of 95Nb as an indicator of redox potential for fuel salt in molten salt reactor (MSR), the specific activity of 95Nb in FLiBe salt and its deposition of 95Nb on Hastelloy C276 have been studied. Experimental results indicated that the amount of 95Nb deposited on Hastelloy C276 resulted from its chemical reduction exhibited a positive correlation with the decrease of 95Nb activity in FLiBe salt and the relative deposition coefficient of 95Nb to 103Ru appeared a well correlation with 95Nb activity in FLiBe salt. Both correlations implied that the measurement of 95Nb activity deposited on Hastelloy C276 specimen might provide a quantitative approach for monitoring the redox potential of fuel salt in MSR.

INFLUENCE OF FUEL FLOW RATE VARIATION ON MOLTEN SALT REACTOR PERFORMANCE

Molten Salt Reactors are Gen-IV reactors that use liquid fuel. Fluid fuel allows continuous removal of fission gases as well as batch fuel reprocessing. With these control mechanisms the system can be sustained within the desired operating temperature range and required power output. These methods rely on the presence of a chemical processing plant on-site that adds complexity. This also creates a risk of processing plant unavailability due to faults, emergency downtime or maintenance. The work considers variation of fuel salt flow rate in Molten Salt Reactors as a means of controlling reactor operation without using reprocessing. The analysis is performed using the Molten Salt Fast Reactor as an example. An extended version of the SERPENT Monte-Carlo transport code coupled with OpenFOAM generic platform were used for capturing delayed neutron drift, decay heat, gaseous fission product removal, calculating fuel salt velocity vectors and the fuel temperature distribution. The two models were coupled via a script that accounted for reactivity insertion between time steps and the changes caused in the fission power. Results confirm that, while operating at constant power, the difference between fuel inlet and outlet temperatures increase as the flow rate decreases. Burnup analysis has shown that while the average fuel temperature continues to reduce with time, the difference between inlet and outlet temperatures can be controlled by varying the flow rate while maintaining constant power. Finally, the variation in the fuel flow rate has been shown to extend the reactor operating time with no insertion of additional fissile inventory.

Inhibition effect of ZrF4 on UO2 precipitation in the LiF–BeF2 molten salt

This study provides an effective solution for controlling and monitoring the nuclear fuel precipitation (UO2) in molten fluorides, which is of great importance for the safe operation and fuel salt design of molten salt reactor (MSR).

CORE DESIGN OF BREED & BURN MOLTEN SALT FAST REACTOR

Previous designs of once-through solid-fuelled breed-and-burn (B&B) reactor and the conventional molten salt reactor (MSR) concepts suffer from material limitation of neutron irradiation damage and chemical corrosion. A novel breed-and-burn molten salt reactor (BBMSR) concept uses separate molten salt fuel and coolant in a linear assembly core configuration. Similar to Moltex Energy Stable Salt Reactor (SSR) design, the configuration with fuel salt contained in fuel tubes and coolant salt in pool type reactor vessel has been previously studied. The study confirmed that breed-and-burn operation is feasible in principle, however with a low neutronic margin. The objective of this paper was to seek improvements of the neutronic margin with a metallic natural uranium blanket design. A parametric study was performed for the natural uranium blanket design. BBMSR neutronic performance simulation was modelled using Serpent, a Monte Carlo reactor physics code, with a single 3D hexagonal channel containing a single fuel tube in an infinite lattice with reflective radial and vacuum axial boundary conditions. The addition of a metallic natural uranium blanket inside the fuel tube, which increases the natural uranium metal to fuel salt ratio (ϒ) of the BBMSR, was shown to significantly increase the neutronic performance of the BBMSR.

A Theory and Model of Molten Salt Reactor Xenon Behavior After the Solubility Limit

Due to the circulating nature of the fuel, there is a qualitative difference between xenon behavior in a molten salt reactor (MSR) compared to a solid fuel reactor. Therefore, the equations that describe 135Xe behavior in a molten salt reactor must be formulated differently. Prior molten salt reactor xenon models have focused on behavior below a solubility limit in which the 135Xe is partially dissolved in the fuel salt. It is foreseeable that a molten salt reactor may operate with a concentration of gas dissolved in the salt sufficiently high such that no further gas may dissolve in the fuel salt. This paper introduces a theory of molten salt reactor xenon behavior for a reactor operating above the solubility limit. A model was developed based on this theory and analyses performed are discussed. Results indicate: (1) steady-state xenon poisoning is not monotonic with respect to gas egress rate, (2) a increase in gas ingress rate leads to a characteristic increase which is followed by a new steady-state in xenon poisoning, and (3) given a sufficient rate of gas egress, it is possible to remove the iodine pit behavior.

Modeling the Molten Salt Reactor Experiment With the GOTHIC Code

GOTHIC 8.3(QA) includes capabilities for modeling advanced, non-light water cooled reactors. Important capabilities introduced in GOTHIC 8.3(QA) include fluid property tables for various molten salts, an enhancement to the tracer tracking module to allow radioactive decay energy to be released locally in the carrier fluid and other improvements to the neutron kinetics module. With these new capabilities in place, GOTHIC is used to benchmark steady-state and transient conditions in the Molten Salt Reactor Experiment (MSRE), which operated at Oak Ridge National Laboratory from 1965 to 1969. In this experimental reactor, UF4 fuel was dissolved in molten fluoride salt, and criticality could be achieved only in the graphite moderated core. An air-cooled radiator transferred fission and decay heat to the environment. The design thermal output of the MSRE was 10 MWt, but the radiator design limited the output to 8 MWt. The original design parameters neglected the impact of decay heat on system temperatures. GOTHIC is used to benchmark system operating parameters at both the 10 MWt design condition and the 8 MWt operating condition, both with and without decay heat. The cases that include decay heat apply 7% of the nominal thermal output using the eleven decay heat precursors from ASB 9-2 as tracers. The results of the benchmark exhibit good agreement with design and operating data and demonstrate heat-up due to decay heat in the fuel salt outside the core. In the MSRE, delayed neutron precursors are not confined to the core because the fuel and fission products flow through the system. As a result, there are different values for (effective) delayed neutron fraction with and without flow, and the decay of delayed neutron precursors outside the core under full-flow conditions reduces reactivity by 0.212 % δk/k. Zero power physics testing included fuel salt pump start-up and coast-down transients with a control rod automatically moving to maintain criticality. The control rod motion calculated by GOTHIC is a reasonable match to measured data from these transients. Low power testing included a natural convection transient with no control rod motion such that reactor power was responding to heat load demand from the radiator. The reactor power and fuel salt and coolant salt temperatures calculated by GOTHIC exhibit good agreement with measured data.