Essays about: "thermal-hydraulics transient"
Showing result 1 - 5 of 6 essays containing the words thermal-hydraulics transient.
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1. Simulation of IB-LOCA in TRACE : A semi-blind study of numerical simulations compared to the PKL test facility
University essay from Uppsala universitet/Tillämpad kärnfysikAbstract : This thesis studied the performance of the thermal hydraulic software TRACE applied on an intermediate sized break (IB) happening on the cold leg in a pressurized water reactor (PWR), causing a loss-of-coolant accident (LOCA). The same accident has previously been simulated in the PKL Test Facility, which is a scaled version of a PWR and is used to simulate transients stemming from different accidents. READ MORE
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2. TRACE Analysis of LOCA Transients Performed on FIX-II Facility
University essay from KTH/KärnkraftssäkerhetAbstract : As a latest developed computational code, TRACE is expected to be useful and effective for analyzing the thermal-hydraulic behaviors in design, licensing and safety analysis of nuclear power plant. However, its validity and correctness have to be verified and qualified before its application into industry. READ MORE
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3. Validation of TRACE Code against ROSA/LSTF Test for SBLOCA of Pressure Vessel Upper-Head Small Break
University essay from KTH/KärnkraftssäkerhetAbstract : OECD/NEA ROSA/LSTF project tests are performed on the Large Scale Test Facility (LSTF). LSTF is a full-height, full-pressure and 1/48 volumetrically-scaled two-loop system which aims to simulate Japanese Tsuruga-2 Westinghouse-type 4-loop PWR. READ MORE
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4. TRACE Code Validation for Natural Circulation During Small Break LOCA in EPR-Type Reactor
University essay from KTH/KärnkraftssäkerhetAbstract : The PWR PACTEL test facility was built in Lappeenranta (Finland) to gain experience in thermal-hydraulics behavior of vertical steam generators used by EPR (European Pressurized Water Reactor) during SBLOCA (Small Break Loss of Coolant Accident) transient, which involves natural circulation phenomenon. The benchmark, which consisted of blind and open part, offered a unique opportunity for code users to improve and test their knowledge and skills in developing the input deck models and performing calculations. READ MORE
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5. Application of CFD to Safety and Thermal-Hydraulic Analysis of Lead-Cooled Systems
University essay from KTH/FysikAbstract : Computational Fluid Dynamics (CFD) is increasingly being used in nuclear reactor safety analysis as a tool that enables safety related physical phenomena occurring in the reactor coolant system to be described in more detail and accuracy. Validation is a necessary step in improving predictive capability of a computationa code or coupled computational codes. READ MORE
