A nuclear fuel test loop (after below, FTL) is installed in IR1 of an irradiation hole in HANARO for testing neutron irradiation characteristics and thermo hydraulic characteristics of a fuel loaded in a light water power reactor or a heavy water power reactor. There is an in-pile section (IPS) and an out-pile section (OPS) in this test loop. When HANARO is normally operated, the fuel loaded in the IPS has a nuclear reaction heat generated by a neutron irradiation. To remove the generated heat and to maintain an operating conditions of the test fuel, a main cooling water system (MCWS) is installed in the OPS of the FTL. The MCWS is composed of a main cooler, a pressurizer, two circulation pumps, a main heater, interconnection pipe line and instruments. The interconnection pipeline is a closed loop which is connected respectively to an inlet and an outlet of the IPS. To absorb the nuclear reaction heat, there is a higher elevation pipeline with a reverse U letter type pipeline installed in upstream of the cooler. And there is another higher elevation pipeline for sucking a fluid by a vertically installed circulation pump with a top suction and a side discharge. Therefore, we predicted that it is difficult to continuously suck a fluid and pressurize the fluid due to air pockets in the two higher elevation pipelines under a high temperature operation. To verify the hot function flow characteristics of the MCWS including air pockets of the higher elevation pipelines, a flow net work analysis has been conducted under a high temperature operation. When the two higher elevation pipelines wholly filled with coolant, it was confirmed through the results that the pump pressurizes the coolant normally. And it was confirmed through the analysis results that the system hot function characteristics met the system design requirements.
Skip Nav Destination
17th International Conference on Nuclear Engineering
July 12–16, 2009
Brussels, Belgium
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-4353-6
PROCEEDINGS PAPER
The Hot Function Flow Characteristics of a Main Cooling Water System for the Nuclear Fuel Test Loop Installed in HANARO
Young-Chul Park,
Young-Chul Park
Korea Atomic Energy Research Institute, Daejeon, Republic of Korea
Search for other works by this author on:
Young-Seob Lee,
Young-Seob Lee
Korea Atomic Energy Research Institute, Daejeon, Republic of Korea
Search for other works by this author on:
Dae-Young Chi
Dae-Young Chi
Korea Atomic Energy Research Institute, Daejeon, Republic of Korea
Search for other works by this author on:
Young-Chul Park
Korea Atomic Energy Research Institute, Daejeon, Republic of Korea
Young-Seob Lee
Korea Atomic Energy Research Institute, Daejeon, Republic of Korea
Dae-Young Chi
Korea Atomic Energy Research Institute, Daejeon, Republic of Korea
Paper No:
ICONE17-75282, pp. 269-275; 7 pages
Published Online:
February 25, 2010
Citation
Park, Y, Lee, Y, & Chi, D. "The Hot Function Flow Characteristics of a Main Cooling Water System for the Nuclear Fuel Test Loop Installed in HANARO." Proceedings of the 17th International Conference on Nuclear Engineering. Volume 3: Thermal Hydraulics; Current Advanced Reactors: Plant Design, Construction, Workforce and Public Acceptance. Brussels, Belgium. July 12–16, 2009. pp. 269-275. ASME. https://doi.org/10.1115/ICONE17-75282
Download citation file:
2
Views
Related Proceedings Papers
Related Articles
Introducing Passive Nuclear Safety in Water-Cooled Reactors - Numerical Simulation and Validation of Natural Convection Heat Transfer and Transport in Packed Beds of Heated Microspheres
ASME J of Nuclear Rad Sci (January,0001)
COBRA-TF Simulation of DNB Response During Reactivity-Initiated Accidents Using the NSRR Pulse Irradiation Experiments
ASME J of Nuclear Rad Sci (July,2016)
An Innovative Falling Film Evaporative Cooling With Recirculation Driven by Low-Grade Heat
J. Thermal Sci. Eng. Appl (December,2009)
Related Chapters
Studies Performed
Closed-Cycle Gas Turbines: Operating Experience and Future Potential
Insights and Results of the Shutdown PSA for a German SWR 69 Type Reactor (PSAM-0028)
Proceedings of the Eighth International Conference on Probabilistic Safety Assessment & Management (PSAM)
New Generation Reactors
Energy and Power Generation Handbook: Established and Emerging Technologies