Sökning: "lower plenum"

Visar resultat 1 - 5 av 7 avhandlingar innehållade orden lower plenum.

1. 1. The Effective Convectivity Model for Simulation and Analysis of Melt Pool Heat Transfer in a Light Water Reactor Pressure Vessel Lower Head

   Författare :Chi Thanh Tran; Tomas Lefvert; Harri Tuomisto; KTH; []
   Nyckelord :TEKNIK OCH TEKNOLOGIER; ENGINEERING AND TECHNOLOGY; light water reactor; hypothetical severe accident; accident progression; accident scenario; core melt pool; heat transfer; turbulent natural convection; heat transfer coefficient; phase change; mushy zone; crust; lower plenum; analytical model; effective convectivity model; CFD simulation; Thermal energy engineering; Termisk energiteknik;

   Sammanfattning : Severe accidents in a Light Water Reactor (LWR) have been a subject of intense research for the last three decades. The research in this area aims to reach understanding of the inherent physical phenomena and reduce the uncertainties in their quantification, with the ultimate goal of developing models that can be applied to safety analysis of nuclear reactors, and to evaluation of the proposed accident management schemes for mitigating the consequences of severe accidents. LÄS MER

3. 2. Development, validation and application of an effective convectivity model for simulation of melt pool heat transfer in a light water reactor lower head

   Författare :Chi Thanh Tran; Truc-Nam Dinh; Florian Fichot; KTH; []
   Nyckelord :NATURVETENSKAP; NATURAL SCIENCES; light water reactor; hypothetical severe accident; accident progression; accident scenario; core melt pool; heat transfer; turbulent natural convection; heat transfer coefficient; phase change; mushy zone; crust; lower plenum; analytical model; effective convectivity model; CFD simulation.; Other physics; Övrig fysik;

   Sammanfattning : Severe accidents in a Light Water Reactor (LWR) have been a subject of the research for the last three decades. The research in this area aims to further understanding of the inherent physical phenomena and reduce the uncertainties surrounding their quantification, with the ultimate goal of developing models that can be applied to safety analysis of nuclear reactors. LÄS MER

4. 3. Input Calibration, Code Validation and Surrogate Model Development for Analysis of Two-phase Circulation Instability and Core Relocation Phenomena

   Författare :Viet-Anh Phung; Pavel Kudinov; Cristophe Demaziere; KTH; []
   Nyckelord :NATURVETENSKAP; NATURAL SCIENCES; Reactor system code; input calibration; code validation; surrogate model; two-phase circulation flow instability; in-vessel core relocation; genetic algorithm; artificial neural network; Reaktorsystemkod; indata kalibrering; kodvalidering; surrogatmodell; två-fas cirkulationsflöde; instabilitet; härdrelokering i reaktortanken; genetisk algoritm; artificiella neurala nätverk; Fysik; Physics;

   Sammanfattning : Code validation and uncertainty quantification are important tasks in nuclear reactor safety analysis. Code users have to deal with large number of uncertain parameters, complex multi-physics, multi-dimensional and multi-scale phenomena. LÄS MER

5. 4. Numerical Investigations on Debris Bed Coolability and Mitigation Measures in Nordic Boiling Water Reactors

   Författare :Zheng Huang; Weimin Ma; Florian Fichot; KTH; []
   Nyckelord :Severe accident; Debris bed; Coolability; Quench; Oxidation; MEWA code; Coupled analysis; Svåra haverier; Grusbädd; Kylbarhet; Kylning; Oxidering; MEWA program; Kopplad analys; Physics; Fysik;

   Sammanfattning : This thesis is aiming at coolability assessment of particulate debris beds formed in hypothetical severe accidents of Nordic boiling water reactors (BWRs) which may employ either lower drywell flooding or control rod guide tubes (CRGT) cooling as severe accident management strategies. For this purpose, quench and cooling limit (dryout) of debris beds after their formation from fuel coolant interactions were investigated by numerical simulations using the MEWA code. LÄS MER

7. 5. Thermo-mechanical Assessment of Reactor Pressure Vessels of Light Water Reactors During Severe Accidents

   Författare :Hongdi Wang; Walter Villanueva; Sevostian Bechta; Weimin Ma; Artem Kulachenko; Michael Fitzpatrick; KTH; []
   Nyckelord :Severe accident scenario; reactor pressure vessel; in-vessel melt retention; failure analysis; finite element analysis; validations; Scenario för svåra haverier; reaktortank; in-vessel retention; feleffektsanalys; Finite Element Analyser; validering; Physics; Fysik;

   Sammanfattning : The reactor pressure vessel (RPV) is one of the crucial safety barriers designed to isolate the reactor core, safeguarding against potential radioactive releases into the environment during a severe accident. The assessment of RPV behaviour and its failure is necessary to predict the characteristics of melt release into the reactor pit and succeeding ex-vessel accident progression. LÄS MER