Investigate the fatigue life of the PWR nuclear reactor vessel

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dc.contributor.advisorHuri, Dávid
dc.contributor.authorKalthoum, Mhd Wafik
dc.contributor.departmentDE--Műszaki Kar
dc.date.accessioned2024-06-20T07:48:08Z
dc.date.available2024-06-20T07:48:08Z
dc.date.created2024
dc.description.abstractInternational design regulations can be used to test and design a reactor pressure vessel. These codes give engineers proactive steps that can be taken to prevent any major catastrophes. Through the use of the ANSYS finite element program, this project examines the efficacy of the elastic-plastic limit based design approaches of the primary International Design Codes (ASME and BS). For the analysis, a standard pressurized water reactor (PWR) with a 300 MW capacity was chosen. SA-508 Gr.3 Cl.1, a nuclear grade steel, was utilized as the RPV material in the comparison. It has been determined that the design by analysis approach can be utilized to eliminate the needless caution that results from using the design by rule approach, which is covered in BS-5500 section 3. This study suggests that by applying the design by analysis approach as outlined in the ASME code, the maximum allowed pressure of the RPV may be increased by up to 26.37%.
dc.description.correctorLB
dc.description.courseMechanical Engineeringen
dc.description.degreeMSc/MA
dc.format.extent52
dc.identifier.urihttps://hdl.handle.net/2437/374045
dc.language.isoen
dc.rights.accessHozzáférhető a 2022 decemberi felsőoktatási törvénymódosítás értelmében.
dc.subjectFEA
dc.subjectCAD
dc.subject.dspaceEngineering Sciences
dc.titleInvestigate the fatigue life of the PWR nuclear reactor vessel
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