Liquid-metal Reactor Core Research Articles

Simulation of the EBR-II Loss-of-Flow Tests Using the MARS Code

As part of the development of a safety analysis methodology for a liquid-metal reactor (LMR) in Korea, the Multidimensional Analysis of Reactor Safety (MARS) code was selected as a system transient safety analysis code. The Korea Atomic Energy Research Institute developed the MARS code to analyze safety and thermal-hydraulic phenomena related to a two-phase flow in the transients of water reactors a decade ago. The addition of thermal-hydraulic models related to liquid metal as a coolant and reactivity feedback models associated with the kinetics calculation of an LMR core is required for the application of the MARS to the transients of an LMR design. A table for various properties of liquid sodium, several heat transfer coefficients according to flow regimes and geometries, and the models for a pressure drop due to the wire spacers of the LMR core were newly implemented. The improved MARS code was verified through the analysis of three shutdown heat removal tests (SHRT)-17, -39, and -45 conducted in the Experimental Breeder Reactor (EBR)-II reactor. The SHRT-17 test involved a simultaneous loss of electrical power to all pumps and a reactor scram from 100% power and flow. Thus, the test simulated a thermal-hydraulic transition from a forced convection to the totally passive decay heat removal due to a natural circulation. SHRT-39 and SHRT-45 are loss-flow tests without a reactor scram. However, the pump coastdown periods and initial states of the plant are different from each other. Simulated results for the flow rate and temperature for an instrumented subassembly agree well with the experimental data.

An experimental study on LMR core seismic behavior with fluid couplings between closely spaced hexagons

In LMR (Liquid Metal Reactor) core seismic behavior, the fluid coupling between closely spaced hexagons is known to have a very strong effect on the natural frequencies and the seismic responses. In this paper, an experimental study is carried out to investigate the fluid coupling effects with reduced scale mock-ups with three types of hexagonal duct systems and three types of input seismic motions such as the ElCentro 1940 time history, the artificial time history (Reg. 1.60), and the Kobe time history by using a one-dimensional shaking table facility. The results of the experimental studies are compared and discussed with those of the analyses performed by using the SAC-CORE code, which implements a numerical algorithm of the CFAM (Consistent Fluid Added Mass) matrix proposed in a previous paper. From the results, it is found that the fluid coupling between closely spaced hexagons becomes stronger with an increasing number of the neighboring hexagonal ducts, and this causes the natural frequency to be significantly changed and it affects the seismic time history responses too.

Uncertainty in the Burnup Reactivity Swing of Liquid-Metal Fast Reactors

The uncertainty in the burnup reactivity swing {sigma}k{sub b} attributable to nuclear data uncertainties is analyzed using depletion-dependent sensitivity coefficients for single- and multicycle equilibrium depletion. Four systems are analyzed with design features that encompass many of the design options considered for current U.S. advanced liquid-metal reactor cores. These systems, while characterized by very different {sigma}k{sub b} values in the range from {minus}0.22 to 3.87% {Delta}k, exhibit much smaller differences in their {sigma}k{sub b} uncertainties, which range from 0.18 to 0.33% {Delta}k. The {sigma}k{sub b} uncertainties depend primarily on the design choices of core size and fissile fuel type, as well as whether the analysis represents multicycle effects. For all reactors analyzed, the burnup swing uncertainty is dominated by the {sup 238}U capture reaction. In this paper the potential for reducing uncertainties by a factor of 3 by use of available integral experiment results is demonstrated.

An Analysis of Gamma-Ray Energy Deposition in a Heterogeneous Liquid-Metal Fast Breeder Reactor

Gamma-dose analysis was performed on an absolute-value basis at the zero-power plutonium reactor (ZPPR)-13B/4 critical assembly, which was one of the benchmark radially heterogeneous 650-MW (electric) liquid-metal reactor cores, to validate the current data and methods applicable to power reactor design calculations. Discussions of particular aspects inherent to the critical measurements, such as heterogeneity in the ZPPR plate cells and delayed fission product gamma effect, are included. The results show that the gamma-energy deposition distribution was reproduced within an accuracy of {approximately} 10% in cores that had complicated fuel and internal blanket layouts.

Mixed Convection in Parallel Channels with Application to the Liquid-Metal Reactor Concept

In this paper mixed convection flow for parallel vertical channels connected at upper and lower plenums is studied. The one-dimensional conservation equations are formulated in dimensionless form using channel integral parameters. Based on this formulation, expressions are derived for stable flow and reversal of channel flow. The equations are then used to calculate the flow redistribution within a liquid-metal reactor core during natural circulation primary loop flow. A channel/plenum interaction phenomenon, which limits the applicability of using one-dimensional formulations, is modeled, and a correlation is formulated utilizing measured results to predict the onset of this behavior. Finally, the reversal of a heated channel from upflow to downflow, which cannot be predicted with a one-dimensional analysis, is described, and the channel/plenum interaction previously modeled is proposed as the mechanism that initiates this flow reversal.