• Nuclear Engineering and Technology
• /
• 제38권2호
• /
• pp.185-194
• /
• 2006

Transient operations for an integral type reactor, SMART-P, have been experimentally investigated using a thermal-hydraulic integral test facility, VISTA (Experimental Verification by Integral Simulation of Transients and Accidents), in order to verify the system design and performance of the SMART-P, a pilot plant of SMART. The VISTA facility was subjected to various accident conditions such as feedwater increase and decrease, loss of coolant flow, and control rod withdrawal accidents in order to elucidate the thermal-hydraulic responses following such accidents and finally to verify the system design of the SMARTP. Full functional control logics have been implemented in the VISTA facility in order to control the required control action for an accident simulation. As one of the sensitivity tests to verify the PRHRS performance, the effects of the initial water level in the compensation tank are experimentally investigated. When the initial water level is 16%, the water is quickly drained and nitrogen gas is then introduced into the PRHR system, resulting in deterioration of the PRHRS performance. It is thus found that nitrogen ingression should be prevented to ensure stable PRHRS operation.

• Nuclear Engineering and Technology
• /
• 제52권12호
• /
• pp.2789-2802
• /
• 2020

Small Modular Reactors (SMRs) are attracting wide attention due to their outstanding performance, extensive studies have been carried out for lead-based fast reactors (LFRs) that cooled with Lead or Lead-bismuth (LBE), and small modular natural circulation LFR is one of the promising candidates for SMRs and LFRs development. One of the challenges for the design small modular natural circulation LFR is to master the natural circulation thermal-hydraulic performance in the reactor primary circuit, while the natural circulation characteristics is a coupled thermal-hydraulic problem of the core thermal power, the primary loop layout and the operating state of secondary cooling system etc. Thus, accurate predicting the natural circulation LFRs thermal-hydraulic features are highly required for conducting reactor operating condition evaluate and Thermal hydraulic design optimization. In this study, a thermal-hydraulic analysis code is developed for small modular natural circulation LFRs, which is based on several mathematical models for natural circulation originally. A small modular natural circulation LBE cooled fast reactor named URANUS developed by Korea is chosen to assess the code's capability. Comparisons are performed to demonstrate the accuracy of the code by the calculation results of MARS, and the key thermal-hydraulic parameters agree fairly well with the MARS ones. As a typical application case, steady-state analyses were conducted to have an assessment of thermal-hydraulic behavior under nominal condition, and several parameters affecting natural circulation were evaluated. What's more, two characteristics parameters that used to analyze natural circulation LFRs natural circulation capacity were established. The analyses show that the core thermal power, thermal center difference and flow resistance is the main factors affecting the reactor natural circulation. Improving the core thermal power, increasing the thermal center difference and decreasing the flow resistance can significantly increase the reactor mass flow rate. Characteristics parameters can be used to quickly evaluate the natural circulation capacity of natural circulation LFR under normal operating conditions.

• Nuclear Engineering and Technology
• /
• 제41권5호
• /
• pp.677-690
• /
• 2009

A previous study demonstrated that the two-row fuel assembly has much more favorable neutron-physical and thermal-hydraulic behavior than the conventional one-row fuel assemblies. Based on the newly developed two-row fuel assembly, an SCWR core is proposed and analyzed. The performance of the proposed core is investigated with 3-D coupled neutron-physical and thermal-hydraulic calculations. During the coupling procedure, the thermal-hydraulic behavior is analyzed using a sub-channel analysis code and the neutron-physical performance is computed with a 3-D diffusion code. This paper presents the main results achieved thus far related to the distribution of some neutronic and thermal-hydraulic parameters. It shows that with adjustment of the coolant and moderator mass flow in different assemblies, promising neutron-physical and thermal-hydraulic behavior of the SCWR core is achieved. A sensitivity study of the heat transfer correlation is also performed. Since the pin power in fuel assemblies can be non-uniform, a sub-channel analysis is necessary in order to investigate the detailed distribution of thermal-hydraulic parameters in the hottest fuel assembly. The sub-channel analysis is performed based on the bundle averaged parameters obtained with the core analysis. With the sub-channel analysis approach, more precise evaluation of the hot channel factor and maximum cladding surface temperature can be achieved. The difference in the results obtained with both the sub-channel analysis and the fuel assembly homogenized method confirms the importance of the sub-channel analysis.

• 한국기계기술학회지
• /
• 제13권4호
• /
• pp.85-91
• /
• 2011

Experimental analysis has been carried out to investigate thermal characteristics of hydraulic system in special vehicles. Hydraulic system performance is largely influenced by oil temperature, and there are considerable performance decline and malfunctions in the system for high temperature conditions caused by heavy load and continuous operation. Transient oil temperature and pressure variation are analyzed and heat generation rates in the several main system parts are compared for various flow rates. With the start of system operation oil temperature gradually increases, and viscosity deceases by about 70% as temperature increases from $20^{\circ}C$ to $80^{\circ}C$. Operation pressure in the hydraulic system decreases with oil temperature, and its variation rate becomes less steep as oil temperature increases. Heat generation rate in hydraulic pump also depends on the oil temperature, and it reaches maximum near $50^{\circ}C$. These results in this study can be applied to optimal design of efficient hydraulic system in special vehicles.

• Nuclear Engineering and Technology
• /
• 제40권3호
• /
• pp.199-212
• /
• 2008

A small-break loss of coolant accident (SB-LOCA) test with a break size equivalent to a 3-inch cold leg break of the APR1400 was carried out as the first transient integral effect test using the ATLAS (Advanced Thermal-hydraulic Test Loop for Accident Simulation). This was the first integral effect test to investigate the integral performance of the test facility and to verify its simulation capability for one of the design-basis accidents. Reasonably good thermal hydraulic data was obtained so that an integral performance of the fluid sub-systems was identified and control performance of the ATLAS was confirmed under real thermal hydraulic conditions. Based on the measured data, a post-test calculation was carried out using the best-estimate thermal hydraulic safety analysis code, MARS 3.1, and the similarity between the expected and actual data was investigated. On the whole, the post-test calculation reasonably predicts the major thermal hydraulic parameters measured during the SB-LOCA test. The obtained data will be used to enhance the simulation capability of the ATLAS and to improve an input model of the ATLAS for simulation of other target scenarios.

• 드라이브 ㆍ 컨트롤
• /
• 제18권2호
• /
• pp.1-8
• /
• 2021

In this work, a circuit with a hydraulic power unit is formulated as a means of predicting the behavior of the prefill valve in the future. The behavior of the prefill valve can be examined by the measurements of the configured power unit, and the performance is determined by using hydraulic pumps, relief valves, and hydraulic hoses that make up the power unit. In particular, pressure/flow pulsation generated by hydraulic pumps can cause instability in the prefill valve and cause noise-induced degradation of the overall performance and reliability of the hydraulic system containing the prefill valve. Therefore, to study the behavior and performance of the prefill valve in a relatively accurate manner, the prediction of the characteristics of the hydraulic power unit driving the prefill valve is very important. In this study, the pulsation characteristics of the hydraulic pump were analyzed to theoretically demonstrate its relationship with different settings of the power unit, such as relief valve pressure settings and the presence/absence of the hose.

• 드라이브 ㆍ 컨트롤
• /
• 제15권3호
• /
• pp.43-48
• /
• 2018

The hydraulic control valve, used in the CPB (cold-pad-Batch) cold dyeing system, passes through a pressurized material that absorbs the dye. The hydraulic control of the hydraulic control panel shall be driven in a uniform and precisely controlled manner, as it interferes directly with the dyschromatism. In this study, an acceleration test model was employed to verify the durability of the hydraulic control of the hydraulic control panel, which was manufactured by the scenic model, and the pre-roll angle was analyzed before the performance of acceleration test. Based on the change in the amount of deformation of the padder roll the durability of the padder roll was analyzed along with verification of the durability of the skin and the rubber coating in contact with the fabric. Furthermore, the accelerated test method used for hydraulic controlled multi-cell padder rolls was verified.

• 한국원자력학회:학술대회논문집
• /
• 한국원자력학회 1995년도 추계학술발표회논문집(1)
• /
• pp.377-382
• /
• 1995

An evaluation method of spacer grid thermal mixing performance in rod bundles is suggested based on hydraulic tests in a single phase flow. Heat transfer correlation was derived by the analogy between momentum and heat transfer. Three of major factors, such as blockage ratio of spacer grid, convective flow swirling, and turbulent intensity, were found to be significantly influential to the spacer grid thermal mixing performance. Local heat transfer near spacer grid was predicted for the hydraulic test of 6 ${\times}$ 6 rod bundles with neighboring different spacer grids.

• Nuclear Engineering and Technology
• /
• 제49권7호
• /
• pp.1388-1395
• /
• 2017

The Helically coiled tube Once-Through Steam Generator (H-OTSG) is a key piece of equipment for compact small reactors. The present study developed and verified a thermal-hydraulic design and performance analysis computer code for a countercurrent H-OTSG installed in a small pressurized water reactor. The H-OTSG is represented by one characteristic tube in the model. The secondary side of the H-OTSG is divided into single-phase liquid region, nucleate boiling region, postdryout region, and single-phase vapor region. Different heat transfer correlations and pressure drop correlations are reviewed and applied. To benchmark the developed physical models and the computer code, H-OTSGs developed in Marine Reactor X and System-integrated Modular Advanced ReacTor are simulated by the code, and the results are compared with the design data. The overall characteristics of heat transfer area, temperature distributions, and pressure drops calculated by the code showed general agreement with the published data. The thermal-hydraulic characteristics of a typical countercurrent H-OTSG are analyzed. It is demonstrated that the code can be utilized for design and performance analysis of an H-OTSG.

• Nuclear Engineering and Technology
• /
• 제49권7호
• /
• pp.1537-1546
• /
• 2017

Many thermal-hydraulic tests have been conducted at the Korea Atomic Energy Research Institute for verification of the SMART (System-integrated Modular Advanced ReacTor) design, the standard design approval of which was issued by the Korean regulatory body. In this paper, the contributions of these tests to the standard design approval of SMART are discussed. First, an integral effect test facility named VISTA-ITL (Experimental Verification by Integral Simulation of Transients and Accidents-Integral Test Loop) has been utilized to assess the TASS/SMR-S (Transient and Set-point Simulation/Small and Medium) safety analysis code and confirm its conservatism, to support standard design approval, and to construct a database for the SMART design optimization. In addition, many separate effect tests have been performed. The reactor internal flow test has been conducted using the SCOP (SMART COre flow distribution and Pressure drop test) facility to evaluate the reactor internal flow and pressure distributions. An ECC (Emergency Core Coolant) performance test has been carried out using the SWAT (SMART ECC Water Asymmetric Two-phase choking test) facility to evaluate the safety injection performance and to validate the thermal-hydraulic model used in the safety analysis code. The Freon CHF (Critical Heat Flux) test has been performed using the FTHEL (Freon Thermal Hydraulic Experimental Loop) facility to construct a database from the $5{\times}5$ rod bundle Freon CHF tests and to evaluate the DNBR (Departure from Nucleate Boiling Ratio) model in the safety analysis and core design codes. These test results were used for standard design approval of SMART to verify its design bases, design tools, and analysis methodology.