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Thermal-hydraulic calculations for a fuel assembly in a European Pressurized Reactor using the RELAP5 code

Skrzypek M., Laskowski R.

Nukleonika

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2015

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Vol. 60, No. 3, part 2

537--544

EN

The main object of interest was a typical fuel assembly, which constitutes a core of the nuclear reactor. The aim of the paper is to describe the phenomena and calculate thermal-hydraulic characteristic parameters in the fuel assembly for a European Pressurized Reactor (EPR). To perform thermal-hydraulic calculations, the RELAP5 code was used. This code allows to simulate steady and transient states for reactor applications. It is also an appropriate calculation tool in the event of a loss-of-coolant accident in light water reactors. The fuel assembly model with nodalization in the RELAP5 (Reactor Excursion and Leak Analysis Program) code was presented. The calculations of two steady states for the fuel assembly were performed: the nominal steady-state conditions and the coolant flow rate decreased to 60% of the nominal EPR flow rate. The calculation for one transient state for a linearly decreasing flow rate of coolant was simulated until a new level was stabilized and SCRAM occurred. To check the correctness of the obtained results, the authors compared them against the reactor technical documentation available in the bibliography. The obtained results concerning steady states nearly match the design data. The hypothetical transient showed the importance of the need for correct cooling in the reactor during occurrences exceeding normal operation. The performed analysis indicated consequences of the coolant flow rate limitations during the reactor operation.

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Computer codes in the safety analysis for nuclear power plants. Computational capabilities of thermal-hydraulic tools, using the example of the RELAP5 code

Skrzypek E., Skrzypek M.

Journal of Power Technologies

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2014

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Vol. 94, nr 5

41--50

EN

Safety is a paramount concern of the Nuclear Power Program in Poland. To this end there is a need to investigate the design of the proposed reactor and its operation principles and perform multiple analyses both before the reactor start-up (The Pre-Construction Safety Report (PCSR) and during its operational life. In the worldwide nuclear community hundreds of people are involved in this complicated and complex process. Due to the sophistication of the phenomena occurring during operation and accidents, the number of analyses is increasing rapidly. Currently, much interest in this field is focused on the use of computer codes and high computational power.

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RELAP5 Capability to Predict Pressure Wave Propagation Phenomena in Single- and Two-Phase Flow Conditions

Sokolowski Ł., Koszela Z.

Journal of Power Technologies

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2012

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Vol. 92, nr 3

150--165

EN

Correct evaluation of the hydrodynamic loads induced by large and rapid pressure waves propagating with the speed of sound along the reactor piping systems and Reactor Pressure Vessel (RPV) is an important and difficult issue of nuclear power plant safety. The pressure shock transients and resulting hydrodynamic loads on the pipes and RPV structures are commonly calculated with one-dimensional thermo-hydraulic system codes such as RELAP5, TRACE, DRAKO and ROLAST. In Sweden, the most widely used computer code for this purpose is RELAP5. This code needs, therefore, to be assessed for its capability to predict pressure wave behavior. The conducted assessment involves simulations of single- and two-phase shock-tube problems and two-phase blowdown as well as water hammer experiments. The performed numerical experiments clearly show that RELAP5, with the proper time step and spatial mesh size, is capable of predicting the complex dynamics of single- and two-phase pressure wave phenomena with good to reasonable accuracy.

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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

Rahgoshay M., Rahmani Y.

Nukleonika

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2007

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Vol. 52, nr 4

159-165

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.