Sökning: "thermal hydraulic system code"

Visar resultat 1 - 5 av 7 uppsatser innehållade orden thermal hydraulic system code.

  1. 1. Pressure loss characterization for cooling and secondary air system components in gas turbines

    Uppsats för yrkesexamina på avancerad nivå, Luleå tekniska universitet/Institutionen för teknikvetenskap och matematik

    Författare :Frida Isaksson; [2017]
    Nyckelord :Gas turbines; cooling and secondary air system; pressure losses; discharge coefficient; minor loss; one-dimensional modelling;

    Sammanfattning : There is a constant struggle to increase the efficiency in gas turbines, where one method is to have a higher inlet temperature to the turbine. Often, this results in temperatures higher than the critical temperature of the materials, which makes cooling of the components an important part of the turbine. LÄS MER

  2. 2. CFD-analysis of buoyancy-driven flow inside a cooling pipe system attached to a reactor pressure vessel

    Master-uppsats, Linköpings universitet/Mekanisk värmeteori och strömningsläraLinköpings universitet/Tekniska högskolan

    Författare :Jens Petersson; [2014]
    Nyckelord :CFD; OpenFOAM; LES; nuclear reactor pressure vessel;

    Sammanfattning : In this work a cooling system connected to a reactor pressure vessel has been studied using the CFD method for the purpose of investigating the strengths and shortcomings of using CFD as a tool in similar fluid flow problems within nuclear power plants. The cooling system is used to transport water of 288K (15°C) into a nuclear reactor vessel filled with water of about 555K (282°C) during certain operating scenarios. LÄS MER

  3. 3. TRACE Analysis of LOCA Transients Performed on FIX-II Facility

    Master-uppsats, KTH/Kärnkraftsäkerhet

    Författare :XIAO HU; [2012]
    Nyckelord :LOCA; TRACE; thermal-hydraulics transient; sensitivity analysis;

    Sammanfattning : 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. LÄS MER

  4. 4. Validation of TRACE Code against ROSA/LSTF Test for SBLOCA of Pressure Vessel Upper-Head Small Break

    Master-uppsats, KTH/Kärnkraftsäkerhet

    Författare :Mian Xing; [2012]
    Nyckelord :SBLOCA; LSTF; TRACE; thermal-hydraulics transient; safety analysis;

    Sammanfattning : 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. LÄS MER

  5. 5. Application of CFD to Safety and Thermal-Hydraulic Analysis of Lead-Cooled Systems

    Master-uppsats, KTH/Fysik

    Författare :Marti Jeltsov; [2011]
    Nyckelord :Coupled Codes; Verification Validation; CFD; System Thermal-Hydraulics; Lead Cooled sys-;

    Sammanfattning : 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. LÄS MER