Sökning: "reactor transients"
Visar resultat 1 - 5 av 10 avhandlingar innehållade orden reactor transients.
Sammanfattning : Coupled calculations are important for the simulation of nuclear power plants when there is a strongfeedback between the neutron kineticsand the thermal-hydraulics. A general coupled model of the Ringhals-3 Pressurized Water Reactor has beendeveloped for this purpose. LÄS MER
2. Development of Effective Algorithm for Coupled Thermal-Hydraulics – Neutron-Kinetics Analysis of Reactivity TransientDetta är en avhandling från Stockholm : KTH
Sammanfattning : Analyses of nuclear reactor safety have increasingly required coupling of full three dimensional neutron kinetics (NK) core models with system transient thermal-hydraulics (TH) codes. To produce results within a reasonable computing time, the coupled codes use different spatial description of the reactor core. LÄS MER
3. Effective Spatial Mapping for Coupled Code Analysis of Thermal–Hydraulics/Neutron–Kinetics of Boiling Water ReactorsDetta är en avhandling från Stockholm : KTH Royal Institute of Technology
Sammanfattning : Analyses of nuclear reactor safety have increasingly required coupling of full three dimensional neutron kinetics (NK) core models with system transient thermal–hydraulics (TH) codes. In order to produce results within a reasonable computing time, the coupled codes use two different spatial description of the reactor core. LÄS MER
4. Development, validation and application of an effective convectivity model for simulation of melt pool heat transfer in a light water reactor lower headDetta är en avhandling från Stockholm : KTH
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
- Detta är en avhandling från Stockholm : KTH Royal Insitute of Technology
Sammanfattning : In this thesis, the feasibility to use a medium sized sodium cooled fast reactor fully loaded with MOX fuel for efficient transmutation of americium is investigated by simulating the safety performance of a BN600-type fast reactor loaded with different fractions of americium in the fuel, using the safety parameters obtained with the SERPENT Monte Carlo code.The focus is on americium mainly due to its long-term contribution to the radiotoxicity of spent nuclear fuel and its deterioration on core's safety parameters. LÄS MER