Engineering topics

Kim, Yeon-Sik

Publications and source records attributed to Kim, Yeon-Sik.

3 records

DOE OSTI2017

Korean Localization Program for Performance Verification Test of APR1400's RCP

The Korean government together with the Korean nuclear society completed the development of technologies for the design, manufacturing, and performance verification test of the APR1400's Reactor Coolant Pump (RCP) as one of the national programs for the localization of deficient nuclear technology. The RCP localization program was driven by KAERI, which was responsible for the development of the performance verification and evaluation technology, and by DOOSAN heavy industrial company, which was in charge of the design and manufacturing of the RCP. The test facility for the localized APR1400's RCP was constructed at the KAERI site at the end of 2012. The thermal hydraulic and electric capability of the RCP test facility (RCPTF) covers up to 18.5 MPa, 343 deg. C, 11.7 m{sup 3}/s, and 14.0 MW for the design pressure, temperature, flow rate, and maximum electric power, respectively. In 2013, a commissioning test was performed to verify its designed capability, followed by several modifications in the RCPTF including signal processing and control logic to enhance the verification and evaluation capability of the RCP performance. After finishing the commissioning and modification of the RCPTF, a type test for the new-type of RCP was performed successfully. In 2014 and 2015, eight production tests for the localized RCPs, which will be installed in a commercial nuclear power plant in Korea, were successfully completed. As a final stage of the localization program of the RCP, KAERI will perform a seal LOCA test to verify the leakage characteristics and produce experimental data on the localized seal assembly for a safety analysis during a station black out (SBO) condition. In this summary, several technical issues developed through a series of research processes for a performance verification test of the localized RCP are described.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

A Proposed Correlation for Critical Flow Rate of Steam/Water Flow

In the safety analysis of loss-of-coolant-accident (LOCA) and steam generator tube rupture (SGTR) scenarios of light water reactors, a modeling of a break is very important to predict the accident's result. For example, the design of a break simulator to simulate a LOCA in light water reactors requires accurate knowledge of the leakage flow through the break, whose shape can be assumed as an orifice or a nozzle according to the characteristics of the break shape. In the case of a small-break LOCA (SBLOCA) scenario, the critical flow mostly occurs by sub-cooled and saturated liquid water including a two-phase water flow at relatively high-pressure conditions. Among previous critical flow maps, Moody's maps are well known for their critical flow rate and pressure of a single component and two-phase mixture with respect to the upstream enthalpy and pressure. In Moody's model, a slip ratio defined by maximizing the critical flow rate with respect to the slip ratio was introduced, and thus its predictions tended to be too conservative. Later, Moody suggested other maps based on a homogeneous equilibrium model (HEM), and they were useful to predict a choked condition for a given upstream condition. However, their predictions of the critical flow rate seemed to show quite an underestimation compared to various test data. The author suggested more practicable critical flow maps for a critical mass flux and pressure of a steam-water flow with respect to the upstream stagnation conditions using an extended Henry-Fauske model. From a comparison with the selected test data, suggested maps showed conservative predictions within reasonable ranges. In this report, a new correlation for the critical flow rate was developed to simulate the suggested critical flow maps within a reasonable range of error. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

Code Validation of a SBLOCA test in SMART Integral Test Loop

Integrated small modular reactors (iSMRs) are favored because of their unique characteristics, such as enhanced safety without any large break loss of coolant accidents (LBLOCAs), low initial construction costs, and multi-use of desalination and electricity generation. Many iSMRs have been developed thus far. In Korea, an iSMR called SMART was developed after a thorough validation of the components and systems, and finally received its design certificate in 2012 for the first time in the world. To ensure the performance of its safety systems, an integral test loop, SMART-ITL, is constructed. It contains the primary system, secondary system, passive safety systems, and an auxiliary system. The passive safety systems include a passive residual heat removal system (PRHRS), passive injection system with a core makeup tank (CMT) and a safety injection tank (SIT), and an automatic depressurization system (ADS). The height is preserved while the diameter is scaled down to 1:7. A series of SBLOCA tests have been performed with 4 trains of PRHRS, 1 train of CMT and SIT, and ADS. In parallel with the experiments, simulation of a representative test has been conducted to track the underlying thermal hydraulics. SMART-ITL is an integral test loop for SMART and a series of SBLOCA tests for confirming the performance of passive safety systems have been conducted. In this paper, a system code simulation on a SBLOCA test using SMART-ITL is presented. The reactor vessel water level, break flow rate, and CMT and SIT water levels are predicted well. (authors)

42 ENGINEERING↗