Study of temperature distribution of fuel, clad and coolant in the VVER-1000 reactor core during group-10 control rod scram by using diffusion and point kinetic methods

• Autorzy: Rahgoshay M. , Rahmani Y.
• Języki publikacji: EN

Abstrakty

EN

In this paper, through the application of two different methods (point kinetic and diffusion), the temperature distribution of fuel, clad and coolant has been studied and calculated during group-10 control rod scram, in the Bushehr Nuclear Power Plant (Iran) with a VVER-1000 reactor core. In the reactor core of Bushehr NPP, 10 groups of control rods are used of which, group-10 control rods contain the highest amount of injected negative reactivity in terms of quantity as compared to other groups of control rods. In this paper we explain impacts of negative reactivity, caused by a complete or minor scram of group-10 control rods, on thermoneutronic parameters of the VVER-1000 nuclear reactor core. It should be noted that through these calculations and by using the results, we can develop a sound understanding of impacts of this controlling element in optimum control of the reactor core and, on this basis, with careful attention and by gaining access to a reliable simulation (on the basis of results of calculations made in this survey) we can monitor the VVER-1000 reactor core through a smart control system. In continuation, for a more accurate survey and for comparing results of different calculation systems (point kinetic and diffusion), by using COSTANZA-R,Z calculation code (in which neutronic calculations are based on diffusion model) and using WIMS code at different areas and temperatures (for calculation of constant physical coefficients and temperature coefficients needed in COSTANZAR, Z code) for the VVER-1000 reactor core of Bushehr NPP, calculation of temperature distribution of fuel elements and coolant by using diffusion model is made in the course of group-10 control rods scram and afterwards.

Słowa kluczowe

EN

point kinetic; diffusion; COSTANZA-R,Z; VVER-1000; control rod; RELAP5

• Wydawca: Institute of Nuclear Chemistry and Technology
• Czasopismo: Nukleonika
• Rocznik: 2007
• Tom: Vol. 52, nr 4
• Strony: 159--165
• Opis fizyczny: Bibliogr. 8 poz., rys.

Twórcy

autor

Rahgoshay M.

autor

Rahmani Y.

• Department of Nuclear Engineering, Faculty of Engineering, Science and Research Branch, Islamic Azad University, Tehran, Iran, Tel.: [phone], Fax: [phone],

Bibliografia

• 1. ANS-5.1 STANDARD (1979) Decay heat power in light water reactors. American Nuclear Society, Illinois, USA
• 2. BUSHEHR VVER-1000 reactor (2003) Final Safety Analysis Report (FSAR), Chapter 4. Ministry of Russian Federation of Atomic Energy (Atomenergoproekt),Moscow
• 3. Lewis EE (1977) Nuclear power reactor safety. John Wiley & Sons, New York
• 4. RELAP/MODE3.2.2 Beta (1995) Idaho National Engineering Laboratory, Lockheed Idaho Technologies Company, Idaho Falls
• 5. Sonntag RE, Borgnakke C, Van Wylen GJ (1997) Fundamentals of thermodynamics, 5th ed. John Wiley & Sons, New York
• 6. Todress N, Kazimi MS (1982) Nuclear system I. Hemisphere Publishing Corporation, New York
• 7. Vincenti E, Clusaz A (1971) COSTANZA-R,Z Code. Joint Nuclear Research Center, Ispra Establishment, Italy
• 8. WIMS-D4 manual, Winfrith Improved Multigroup Scheme Code System, CCC-576WIMS-D4 (1991) Atomic Energy Establishment, Winfrith, Dorchester