NEUTRONIC FLUX AND POWER DISTRIBUTION IN A NUCLEAR POWER REACTOR USING WIMS-D4 AND CITATION CODES

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1 International Journal of Physics and Research (IJPR) ISSN Vol.2, Issue 2 ec TJPRC Pvt. Ltd., NEUTRONIC FLUX AN POWER ISTRIBUTION IN A NUCLEAR POWER REACTOR USING WIMS-4 AN CITATION COES MAHER N.SARSAM 1 & BASHAIR MOHAMME SAIE 2 1 Alsalam University Collee, Iraq 2 Badad University-collee of Education.Ibn Al-Haitham, Iraq ABSTRACT The flux and power distribution of a WWER- 440 nuclear power reactor usin CITATION code was calculated. The calculation of the neutronic cross sections values was done usin WIMS-4 code. The comparison between results and benchmark values of the power distributions shows a very ood areement. KEYWORS: WIMS-4 Code, WWER- 440 Nuclear Power Reactor INTROUCTION The well known pressurized water reactors (PWR) WWER-440 [1] and WWER-1000 [2] are widely used in Russia and other east countries. The core of this type of reactors includes 349 hexaonal fuel assemblies with pitches of 14.4 cm, 37 of them are control and reulatin assemblies havin the same external dimensions as the fuel assemblies. The control and reulatin assemblies contains a very hih neutron absorber in its upper part, while their compositions are identical in its lower parts as for the fuel assemblies. Each fuel assembly is a cluster of 126 operatin fuel pins, the claddin of the fuel elements is a tube of a special zirconium alloy, composed of zirconium alloy with 1% niobium, havin outer diameter of 9.2 mm and inner diameter equal to 7.72 mm. As the other PWR, the UO2 fuel has three different initial enrichments in U-235 mainly 1.6%, 2.4% and 3.6%. The start- up loadin of the reactor includes 100 fuel assemblies of 3.6% enrichment, 133 fuel assemblies of 2.4% enrichment and 116 fuel assemblies of 1.6% enrichment. The core represents the heat source of a nuclear power plant and the ordinary water is used simultaneously as a moderator as well as coolant in order to remove the heat from the core. The outlet water temperature is 301, while the inlet temperature is 268.Two independent closed heat exchaner exist in this type of reactors, in the primary cycle the temperature is reduced from the operatin temperature 301 to 288, with water pressure of 123 kf/cm2, while in the secondary cycle the temperature is reduced to 268 with a water pressure of 47 kf/cm2 [3]. The most important physical characteristics of the core are: Thermal power Electric power 1375 MW 440 MW Averae enrichment of fresh fuel 2.5% Number of control rods 37 Number of fuel assemblies 312 Pressure 123 bars Outlet temperature 301

2 Neutronic Flux and Power istribution in a Nuclear Power Reactor Usin WIMS-4 and Citation Codes 25 THE IFFUSION EQUATION The one velocity diffusion equation for neutrons can be derived by considerin the causes of the increase and decrease of the neutron density. The chane in neutron density in any volume element of a specified medium results from the flow of neutrons across the surface of the element and the absorption of neutrons by the medium and also the production of neutrons by sources. The net flow of neutrons per unit volume per unit time is iven by Fick s law, which states that, the current density vector is proportional to the neative radient of the flux, i.e.: div J=div( rad ) where: J is the net current, is the diffusion coefficient and, is the neutron flux. If we define S as the number of neutrons emitted by the source per unit volume and unit time, then the time rate of chane of neutron density n is equal: Where is the macroscopic cross section and is the number of neutrons absorbed per unit volume per unit time. If we consider the diffusion coefficient () as independent of position, then for a homoeneous medium, the above equation becomes;[4,5]: For a flux time independent, the equation is reduced to: This equation is known as the steady state diffusion equation for one roup enery. In real situations, neutrons chanes their eneries due to scatterin, they have then different eneries. The one roup diffusion equation must be then substitute by the multiroup diffusion equation it can be obtained by considerin other terms, which must taken into account. The eneral form of multiroup diffusion equation is [6]: With:, =1,2, N : scatterin cross section, which is equal to transport cross sections from roup h to roup : transfer (or removal) cross section for the inelastic scatterin from roup to roup h ( <h ). is the fraction of neutrons appearin in the th roup.

3 26 Maher N. Sarsam & Bashair Mohammed Saied is the averae number of neutrons emitted by fission due to neutron havin eneries in roup h. S (ext) is the external source of neutrons. the fission cross section in roup h THE CALCULATION PROCEURE The reactor lattice code WIMS-4 [7] was used to calculate the initial neutron flux distribution. WIMS has its own library of materials, it consists of 69 enery roups in the enery interval between 0 and 10 MeV. The code solves the neutron transport equation over a specified reion of the reactor lattice, a unit cell in our case. The calculated neutron flux is used to et a set of macroscopic cross sections homoenized over a chosen sub reions and in a chosen broad roup structure. Four enery roups are chosen in our case one for fast enery (10 MeV MeV), 2 roups for epithermal ( MeV KeV and KeV ev ), and one thermal (0.625 ev ev ). The calculation of the flux distribution was done usin the multidimensional CITATION code[8] which is desined to solve the neutron diffusion theory by finite difference method. RESULTS AN ISCUSSIONS Usin the cluster eometry option of Wims-4, the four roups macroscopic cross sections of the three types of fuel and reulatin assemblies and moderator was obtained (tables 1-7). The CITATION code which solves the neutron flux distribution in three dimensions by finite difference techniques was used to obtain the neutron flux in the four considered enery roups in radial and axial coordinates as shown in fiures 1 and 2 respectively. The calculated power distribution across the reactor core was compared with benchmark values as shows in fiure 3. This result indicate clearly the validation of these two codes for WWER reactor type flux and power calculation. REFERENCES 1) Fundamental of WWER Type Pressurized Water Reactors. Technical Research Center of Finland Nuclear Enineerin Laboratory. Helsinki Finland 2) Chatterjee,B. et al. Brief escription of VVER-1000 Reactor Annals of Nuclear Enery 37, ( 2010 ) 3) Technical Proposal and Information Materials for NPP of 880 MW Capacity With WWER-440 ATOMENERGOEXPORT-USSR-(1986) 4) Nuclear Reactor Analysis. J.G.uderstand and J.G.Hamilton John Wiley & Sons (1986) 5) Neutron Physics by Paul Reuss Edn science (paris-france) (2008) 6) etermination of the Neutron Flux in the Reactor Zones with the Stron Neutron Absorption and Leakae V.Ljubenov and M. Milosevic Seberian Journal of Electrical Enineerin vol.1 No 3, ( 2004) 7) Roth M.J. The Preparation of Input ata for the Lattice Code WIMS- 4 AEER-538 Rev.7( 1987) 8) CITATION, Nuclear Reactor Core Analysis Code System. T.B. Fowler,.R.Vondy and G.W.Cunninham ORNL-TM-2496 Rev 2 ( 1999)

4 Neutronic Flux and Power istribution in a Nuclear Power Reactor Usin WIMS-4 and Citation Codes 27 APPENICES Table 1: The Macroscopic Cross Section of the Fuel Assemblies Containin Fuel of 1.6 % Enrichment Table 2: The Macroscopic Cross Section of the Fuel Assemblies Containin Fuel of 2.4 % Enrichment Table 3: The Macroscopic Cross Section of the Fuel Assemblies Containin Fuel of 3.6 % Enrichment Table 4: The Macroscopic Cross Section of the Control and Reulatin Assemblies Containin Fuel of 1.6 % Enrichment Table 5: The Macroscopic Cross Section of the Control and Reulatin Assemblies Containin Fuel of 2.4 % Enrichment

5 28 Maher N. Sarsam & Bashair Mohammed Saied Table 6: The Macroscopic Cross Section of the Control and Reulatin Assemblies Containin Fuel of 3.6 % Enrichment Table 7: The Moderator Macroscopic Cross Section Fi.1: The Four Neutron Enery Group Flux istribution in the Reactor Core

6 Neutronic Flux and Power istribution in a Nuclear Power Reactor Usin WIMS-4 and Citation Codes 29 Fi.2: Relative Vertical istribution of the Four Enery Group Flux Fi.3: Averae Relative Power istribution Across the Reactor Core