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Investigation and Simulation of Overcooling Transients of The Bushehr's VVER-1000 Nuclear Power Plant with RELAP5/MOD3.3

Yousefi, Amir Hossein | 2018

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  1. Type of Document: M.Sc. Thesis
  2. Language: Farsi
  3. Document No: 51467 (46)
  4. University: Sharif University of Technology
  5. Department: Energy Engineering
  6. Advisor(s): Ghofrani, Mohammad Bagher
  7. Abstract:
  8. During the operation of a nuclear power plant, the reactor pressure vessel (RPV) is exposed to a variety of pressure and thermal stresses and neutron radiations. This will cause the loss of the initial strength of the reactor vessel component. During the occurrence of some of accidents, an excessive cooling of the coolant inside the RPV takes place which in the term is called overcooling transients. In addition, in some of these events, a break in a section of the circuit will reduce the water level at the core of the reactor. By reducing the water level, the existing emergency makeup water systems are activated and inject water into the reactor. The temperature of the added water is much lower than the temperature of the water inside the reactor. Because of the temperature drop in the reactor body due to the rapid reduction of pressure or because of the low temperature of the makeup water, a significant thermal stress is generated in the RPV wall. On the other hand, the presence of considerable pressure inside the reactor causes considerable pressure stresses, which ultimately results in increased stresses and damage to the reactor vessel. This phenomena is called the Pressurized Thermal Shock (PTS). In this situation, the strength of the body may be lost, and the composing metal’s property will change from ductile to brittle which in sever conditions will lead to a fracture in the reactor vessel and ultimately dysfunction of the plant. Therefore, it is very important to study the accidents associated with overcooling of the reactor. In this study, first, the technical data of Bushehr nuclear power plant are gathered and various components of the plant are modeled using the RELAP5 code. In order to evaluate the accuracy of modeling, by implementing the code in the steady state of the plant, the results obtained from the steady conditions are compared with the values in the power plant safety records, between which there exists a good agreement. Based on the scenario in the Final Safety Analysis Report (FSAR) of the Bushehr nuclear power plant, some of the events related to overcooling of the reactor are modeled and the results are compared with FSAR reports. Investigating and analyzing the results of the accident simulations, i.e. Large-Break Loss Of Coolant Accident (LB LOCA) in the primary circuit and the Main Steam Line Break accident (MSLB), show that the temperature drop in the reactor vessel reaches a minimum temperature at which the properties of the metal will not change into brittle and hence, the strength of the vessel and its resistance to the thermal shock can be assured
  9. Keywords:
  10. Bushehr Nuclear Power Plant ; Thermal Shock ; RELAP5 Code ; Large Break Less-of-Coolant Accident (LB-LOCA) ; Pressurized Thermal Shock (PTS) ; Main Steam Line Break Accident (MSLBA) ; Overcooling Transients

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