Ji-Young Jeong

ACOUSTIC LEAK DETECTION TECHNOLOGY FOR WATER/STEAM SMALL LEAKS AND MICROLEAKS INTO SODIUM TO PROTECT AN SFR STEAM GENERATOR

Abstract This technical note presents the results of an experimental study of the role of water in sodium leak noise spectrum formation and at various water/steam leak rates of <1.0 g/s. The conditions and ranges for the existence of bubbling and jetting modes in water/steam outflow into circulating sodium through an injector device were determined to simulate a defect in the wall of the heat-transmitting tube of a sodium-water steam generator (SG). Based on experimental leak noise data, the simple dependency of the acoustic signal level on the leak rate of a microleak and small leaks at different frequency bands was presented for the principal analysis to develop an acoustic leak detection methodology for a KALIMER-600, 600-MW(thermal) reactor (K-600) SG, with the operational experiences for noise analysis and measurements of the Bystry neutron (fast neutron) reactor BN-600. Finally, the methodology was tested with the Korea Atomic Energy Research Institute (KAERI) acoustic leak detection system using sodium-water reaction signals of the Institute of Physics and Power Engineering and background noise of the Prototype Fast Reactor (PFR) superheater for methodology development of KAERI, and it was able to detect a leak rate of under 1 g/s and a signal-to-background noise ratio of −22 dB, using this system and methodology.

Design methodology for insulating and cooling of a small modular reactor head by high temperature structural-thermal-fluid analysis

ABSTRACT The design of a sodium-cooled fast reactor (SFR) head can be complicated due to its shape and functions. The head is a component placed in the pressure boundary to shield nuclear radioactive radiation. At the same time, it needs to seal the reactor vessel, support penetrating components, and minimize heat losses. This paper presents a new insulating and cooling design concept of a small SFR head. For a new design, this study shows a comprehensive design approach considering fluid-thermal-structural computations. The interactive design approach refers to dependent simulation steps of three-dimensional (3D) thermal-structural, one-dimensional (1D) heat-transfer, and 3D computational fluid dynamics (CFD) analysis. This multi-domain approach was applied to the head of the large sodium integral effect test facility called sodium test loop for safety simulation and assessment (STELLA-2). And the STELLA-2 head design was proposed as a thick plate with a sandwich type of insulation, cooling the perimeter annulus of the round head-top surface. For the structural design, the ASME design code was utilized, and the head temperature of 346 °C was calculated as its initial design temperature target. In an axial heat-transfer mode from the in-vessel to the head, a 1D finite element model gave 57 and 75 mm insulation thicknesses with a thermal conductivity of 0.07 W/m/K. The cooling effectiveness of the proposed head design was shown through a commercial CFD package.

Experimental Evaluation of Helical-Type Sodium-to-Air Heat Exchanger Performance for Thermal Sizing Design Code Validation

Abstract The Sodium Test Loop for Safety Simulation and Assessment (STELLA) program for demonstration of decay heat removal performance of the Prototype Generation-IV Sodium-cooled Fast Reactor (PGSFR) is in progress at Korea Atomic Energy Research Institute. As the first phase of the program, the STELLA-1 facility has been constructed, and separate-effect tests for the sodium heat exchangers of the safety-grade passive decay heat removal system (PDHRS) have been conducted. A natural-draft sodium-to-air heat exchanger, one of the key heat exchangers of PDHRS, was tested for the performance demonstration and the design code verification and validation. Twenty-nine cases of experiments were conducted with 13 different test conditions for the selected operating and design conditions of PGSFR. Heat transfer rates were experimentally estimated based on the measured inlet/outlet temperatures and flow rates of both the shell side and the tube side. The experimentally obtained heat transfer rates were compared with the values calculated from the design code, which showed good agreement within a 12.6% error range. Finally, the average Nusselt number was obtained from the experimental results considering the convection mode.

Preliminary Performance Test of Mechanical Pump for a Stella-1

In the process of sodium-cooled fast reactor (SFR) design, it is very important to verify thermo-hydraulic performance of each component in the sodium environment. In KAERI (Korea Atomic Energy Research Institute) STELLA (Sodium Integral Effect Test Loop for Safety Simulation and Assessment) project is under a Mid- and Long-term Nuclear R&D Program. The STELLA project is composed of two stages. In the 1st stage the performance for heat exchangers such as DHX (Decay heat exchanger) and AHX (Air heat exchanger) and for PHTS (Primary heat transport system) mechanical pump will be evaluated. The detailed design of each component is based on that of a 600MWe demonstration reactor.Since full-scale components could not be installed in STELLA-1 [1], the model pump is designed to be scaled-down based on the scaling law. Various pump tests have been done in water environment by using model pump.In this study the design features of model pump were described and the scaling parameters were examined. The results of pump performance tests have been also introduced which is essential to perform safety analysis.Copyright