Basınçlı Su Reaktör Kabının Favor (Fracture Analysis Of Vessels Oak Ridge) Kodu ile Yapısal Analizinin Yapılması

Abstract

Pressure vessels are the most important components of nuclear reactors in the view of protecting its integrity. In the past 30 years there have been a great development and innovative studies about the fracture mechanics and risk based methods to analyze the nuclear reactor pressure vessels. These improvements have been integrated into the developed computer codes which perform fracture mechanics analysis of pressure vessels. The decision of whether to extend or not to extend the operational license of the expired or near to be expired nuclear reactors and whether it is necessary to make maintenance to the pressure vessels or not are made by using these codes. The codes which make fracture mechanics analysis of pressure vessels perform their analysis with deterministic and probabilistic approximations. In Fracture Analysis of Vessels-Oak Ridge (FAVOR) there are algorithm modules which perform probabilistic analysis. The analyses are concentrated on the belt line of the pressure vessel which is exposed to fast neutrons along its lifetime. Previously, the FAVOR code was only used for performing analysis of PTS (Pressurized Thermal Shock), however, currently it is possible to make probabilistic fracture mechanics analysis of the transients such as start-up, cool-down conditions and some coolant leakage accidents with the updated forms of FAVOR. In this study, the Steam Generator Tube Rupture (SGTR) accident and the case during which the Pilot Operated Relief Valve (PORV) opens at 20th second were analyzed by using the thermal hydraulic code named as RELAP5 and the data obtained from thermal hydraulic analyses were used in FAVOR code as input data. In the analyzed cases, the initiation of cracks threatening the reactor pressure vessel and therefore, causing the deterioration of the integrity of the reactor pressure vessel have not been observed; this is a result which is expected, since in the literature, it is stated that all of the nuclear reactor pressure vessels in U.S. do not deteriorate and reach below the safe limits which have been defined by Nuclear Regulatory Commission (NRC). In FAVOR code which was opened to use by Oak Ridge National Laboratory 3 reactor pressure vessels with their flaw files were inputted. These are the reactors in U.S with the most damaged vessels since they were exposed to the fast neutrons during their lifetimes. In this study, the Beaver Valley 3-Loop Westinghouse type nuclear reactor has been analyzed and for the Steam Generator Tube Rupture Accident (SGTR) the flaws data have been worsened in a way that they may threat the pressure vessel’s integrity and the results obtained from these cases have been analyzed. The same analyzes have been performed for the accident case in which the Pilot Operated Relief Valve (PORV) was opened at 20th second.