Research & Development

MODELLING AND SIMULATION OF HYDROGEN EFFECT ON THE MECHANICAL BEHAVIOR OF ZIRCALOY-4 ALLOY

A code named HUNEM-1.0 fuel performance code is developed and the cladding is modeled as a one-dimensional cylindrical tube. In the model, the time-dependent diffusion coefficient and mass gain of oxygen were first calculated and then the oxide displacement model was developed by taking the oxide metal interface temperatures into account. Oxide thickness was calculated by modelling the effects of the heat flux, temperature gradient, fast neutron flux and coolant chemistry.

(CEM TOKER) TURKEY - NUCLEAR ENERGY AND ITS PEACEFUL USAGE TURKISH

Cem Toker, Nuclear Engineer, Turkey: This article is written by a recent Turkish nuclear engineering graduate in Turkish to promote public understanding by people in Turkey. We look for more articles in any language to promote public understanding of nuclear energy, nuclear medicine, radioisotopes and radiation and why they are important for people and the environment. Congratulations Cem Toker.

COMPUTATIONAL THERMAL ANALYSIS FOR A NEW CONCEPTUAL CANISTER DESIGN OF A DEEP GEOLOGICAL DISPOSAL OF SPENT FUEL by Mehmet Cem Toker 

     After decades of development on safe, environmentally acceptable options for the long-term management of radioactive materials, it is important to verify how spent fuels will behave and impact on final storage. To help verifying the thermal behavior of high-level or spent nuclear fuels in the final geological field, a computational thermal analysis has been performed for a new conceptual copper canister design that contains fuel assemblies and fuel rods simultaneously and is proposed for high-level or spent nuclear fuel from nuclear power plants. 

       In the new model, in addition to the four separate fuel assemblies loaded into the container designed in the KBS-3 concept of SKB (Swedish Nuclear Fuel and Waste Management Company), which is accepted as a reference, a copper container with 68 fuel rods added to four separate areas was designed into the container. A new vessel design was made by increasing the spent fuel density of the KBS-3 concept and the performance of this vessel within the thermal limits within the specified time was analyzed. Heat transfer in the copper vessel is carried out by heat conduction, and all these processes are completed with a finite element analysis software. 

INVESTIGATION OF NUCLEAR FUEL ROD RELIABILITY UNDER LARGE BREAK LOSS OF COOLANT ACCIDENT IN VVER-1000 TYPE PRESSURIZED WATER REACTORS

In this study; It is aimed to determine the reliability of the nuclear fuel rod in Russian Made VVER-1000 type Pressurized Water Reactors under the conditions of a large-scale coolant loss accident, which is known to pose a risk in terms of nuclear fuel rod integrity deterioration. FRAPTRAN-1.4 fuel rod transition regime performance code was used to examine the nuclear fuel rod performance in the mentioned accident conditions. In the FRAPTRAN-1.4 analysis, the boundary conditions required for the determination of the fuel rod behavior and the response of the system to the transition regime were calculated with the codes RELAP5 Mod3.4 and FRAPCON-3.4a. The thermal-hydraulic simulation of the loss of refrigerant accident was carried out using the code RELAP5 Mod3.4 on an accident scenario based on an 830 mm guillotine pipe fracture in the cold leg where the fluid is sent to the reactor pressure vessel. The normal operating and pre-accident conditions of the fuel rod are modeled and simulated with the code FRAPCON-3.4a. Required data for FRAPTRAN-1.4 are used to perform transition regime simulation and analyze fuel rod reliability after obtaining with FRAPCON-3.4a and RELAP5 Mod3.4.