Burn up Analysis for Fuel Assembly Unit in a Pressurized Heavy Water CANDU Reactor A. A. EL-Khawlani a, Moustafa Aziz b, M. Ismail c and A. Y. Ellithi c a Physics Department, Faculty of Science, High Education, Sana a, Yemen. b Nuclear and Radiological Regulatory Authority, Cairo, Egypt c Faculty o f Science, Cairo University, Cairo, Egypt Received: 1/7/2014 Accepted: 25/8/2014 ABSTRACT MCNPX code has been used for modeling and simulation of an assembly of CANDU Fuel bundle. The assembly is composed of a heterogeneous lattice of 37- element natural Uranium fuel, heavy water moderator and coolant. The fuel bundle is burnt in normal operation conditions of CANDU reactors. The effective multiplication factor (K eff) of the bundle is calculated as a function of fuel burnup. The flux and power distribution are determined. Comparing the concentrations of both Uranium and Plutonium isotopes are analyzed in the bundle. The results of the present model with the results of a benchmark problem, a good agreement was found. Key Words: CANDU Reactor, Multiplication Factor, MCNPX Code. INTRODUCTION A Pressurized Heavy-Water Reactor (PHWR) or CANDU is a nuclear power reactor, commonly using natural uranium as its fuel, that uses heavy water (deuterium oxide D 2O) as its coolant and moderator. The heavy-water coolant is kept under pressure, allowing it to be heated to higher temperatures without boiling as in a PWR. While heavy water is significantly more expensive than ordinary light water, it yields greatly enhanced neutron economy, allowing the reactor to operate without fuel-enrichment facilities (mitigating the additional capital cost of the heavy water) and generally enhancing the ability of the reactor to efficiently make use of alternate fuel cycles. The use of heavy-water moderator is the key to the PHWR system, enabling the use of natural uranium as fuel (in the form of ceramic UO 2), which means that it can be operated without expensive uranium enrichment facilities. Additionally, the mechanical arrangement of the PHWR, which places most of the moderator at lower temperatures, is particularly efficient because the resulting thermal neutrons are "more thermal" than in traditional designs, where the moderator normally runs hot. This means that a PHWR is not only able to "burn" natural uranium and other fuels, but tends to do so more efficiently as well (1-4). Natural Uranium is a mix of isotopes, mostly Uranium-238 and 0.72% (by weight) of fissile uranium-235. A nuclear reactor aims to sustain a steady rate of fission reactions over time, a state known as "criticality". During the moderating process, it helps to separate the neutrons and Uranium, since 238 U has a large affinity for intermediate-energy neutrons ("resonance" absorption), but can easily undergo fission by the few energetic neutrons above ~1.5-2 MeV (5-7). The CANDU- 6 fuel bundle consists of 37 elements, arranged in circular rings. Each element consists of natural uranium in the form of cylindrical pellets of sintered uranium dioxide contained in a zircaloy 4 sheath closed at each end by an end cap. The 37 elements are held together by end plates at each end to form the fuel bundle (8). Lattice Assembly Model: The lattice assembly geometry is shown in fig (1. a,b), the bundle contains 37 fuel rod. They are arranged in the hexagonal shape, a center rod surrounded by three rings of fuel rods, the inner ring contains 6 rods, the intermediate ring contains 12 rods and the outer ring contains 18 rods. The length of a rod is 48 cm and the lattice pitch is 1.488 cm. The fuel rods (composed of natural UO2 pellets Corresponding author E-mail: moustafaaai@yahoo.com 94
and cladded with thin Zircaloy-4 tubes) are surrounded by heavy water coolant, inside the pressure tube (ziconium-nibium alloy). The pressure tube is, in turn, surrounded by a clandria tube (Zicaloy-2), where the moderator (heavy water) fills the annular space between the two tubes. This model is designed by MCNPX code system. The fuel pin data and geometry are taken from table (1) (9). To calculate the neutronic behavior of the lattice assembly, MCNPX code package are used with burn up card, which is used to make tally calculation for the time operation to 14 time step starting from 1 day to 246 days corresponding to the burn up values which starts from 752 MWd/T to 8000 MWd/T, this operation calculations at average fission bundle power is 0.62 MW. The model evaluates the detailed flux, pin power ratio, axial flux and power distribution through the fuel bundle. ANALYSIS OF THE PROBLEM MCNPX computer code package were used to simulate the burn up of the typical three dimensional model of CANDU unit and to perform calculations for the neutronic parameters inside the lattice. MCNPX is a computer code which uses the Monte Carlo method to simulate the neutron history inside the medium (10, 11) and accumulate the tally of the problem. The entire energy range was divided into 5 neutron energy groups and two million neutron history are used to accumulate the neutrons tally. RESULTS AND DISCUSSION A computer model was designed to simulate a super cell of CANDU Reactor and analyze the effect of control rod (12), the results are compared with published data. Fig ( 2) illustrates the multiplication factor versus fuel burn up (MWd/MTU), the results are also compared with the reference values (9). The initial multiplication factor Predicted by MCNPX reduces from 1.0984 at startup to about 0.96635 at 8000 MWd/MTU (11) due to burn up of fuel and consumption of fissile isotopes. The results of the present model have a good agreement with the reference results (9). Fig ( 3) illustrates the lattice average flux as a function of the burn up. The average flux predicted by MCNPX builds up from 4.07E+14 at startup to about 4.62E+14 at 8000 MWd/MTU, the total flux should increase with the increase of the burn up to compensate for reduction and burn up of fuel and stabling the power. Fig (4) illustrates the concentration of 235 U for lattice (atom/barn.cm) as a function of burn up, 235 U decreases with burn up due to consumption and fission of the isotope. Fig (5) illustrates the concentration of Pu isotopes for lattice as a function of the burn up, the Pu isotopes concentration increases with burn up. 239 Pu absorbs neutrons to be transformed into 240 Pu, this isotope is not fissile, it captures another neutron to produce the fissile isotope 241 Pu, finally the 241 Pu may undergo fission or goes to 242 Pu (13). The fractional content of each isotope depends on the burn up of the fuel, so the Pu concentrations increase with burn up. Fig (6) shows the relation between axial power and axial distance, the axial power decreases in the boundary which is near the coolant because the thermal flux decreases and the axial power increases in the center of the assembly and outer the boundary due to the moderator. Fig ( 7) shows the relation between the axial flux and axial distance, as observed from the figure, the axial flux decreases in the boundary due to the leakage of the neutrons to the coolant at the boundary and it increases in the center of the assembly. Fig (8): illustrates the thermal flux distribution map through the lattice at startup, the thermal flux decreases at the center of the model, and it is clear that the thermal flux increases in the outer ring while decreases in the inner ring, due to self shielding of neutrons in the inner rings. The maximum thermal flux is 1.32E+14 and the minimum thermal flux is 0.914E+14. High thermal flux increases near the lattice periphery adjacent the water layers which thermalize the neutron flux. 95
Fig ( 9): illustrates the pin power ratio distribution map through the lattice at startup, the pin power increases in the outer rings while decreases in the inner rings. The outer ring makes self shielding to the inner neutrons, so the thermal neutrons decreases, consequently, fission rate and finally the pin power decreases also.the maximum pin power through the lattice is 2.02E-2 and the minimum pin power is 1.3E-2. Fuel rods Moderator Coolant Moderator Pressure tube Fig. (1).a: horizontal cross section of MCNPX bundle. Moderator Fuel rods Coolant Fig (1).b: vertical cross section of MCNPX bundle. 96
1.2 1.1 keff keff ref 1 keff 0.9 0.8 0.7 0.6 0 1000 2000 3000 4000 5000 6000 7000 8000 9000 burnup MWd/MTU Fig. (2): k eff for lattice predicted by MCNPX as compared with reference versus burnup. 4.7 x 10 14 4.6 Total Flux (n/cm 2.sec) 4.5 4.4 4.3 4.2 4.1 4 0 1000 2000 3000 4000 5000 6000 7000 8000 9000 burnup (MWd/MTU) Fig. (3): Total average neutron flux for lattice predicted by MCNPX versus burnup. 97
1.8 x 10-4 1.6 U- 235 ( atom/barn.cm ) 1.4 1.2 1 0.8 0.6 0.4 0 1000 2000 3000 4000 5000 6000 7000 8000 9000 burnup MWd/MTU Fig. (4): U-235 Concentration for lattice predicted by MCNPX versus burnup. 6 x10-5 PU239 Pu- isotops (atom/barn.cm) 5 4 3 2 1 PU240 PU241 PU242 Pu 239 Pu 240 0 Pu 241 Pu 242 0 2 4 6 8 10 12 14 16 burnup (MWd/MTU) Fig. (5): Pu- isotopes concentrations for lattice predicted by MCNPX versus burnup. 98
4.5 x 10 2 4 3.5 axial power (W) 3 2.5 2 1.5 1 0.5 0-25 -20-15 -10-5 0 5 10 15 20 25 axial distance (cm) 4 X 10 14 3.5 axial flux (n/cm 2.sec) 3 2.5 2 1.5 1 0.5 0-25 -20-15 -10-5 0 5 10 15 20 25 axial distance (cm) Fig. (7): axial flux for lattice predicted by MCNPX vs. axial distance. 99
1.31 1.22 1.22 1.31 1.22 1.07 1.02 1.07 1.22 1.22 1.03 0.95 0.95 1.03 1.23 1.31 1.07 0.95 0.914 0.95 1.07 1.31 1.23 1.03 0.95 0.95 1.03 1.22 1.23 1.07 1.03 1.07 1.22 1.32 1.23 1.22 1.31 Fig (8): Thermal flux distribution for the lattice for fresh fuel, the results multiply by 10 14 n/cm 2.sec. 100
1.235 1.129 1.136 1.228 1.136 0.969 0.919 0.963 1.136 1.136 0.919 0.833 0.833 0.926 1.142 1.235 0.963 0.839 0.802 0.839 0.969 1.228 1.142 0.919 0.846 0.839 0.926 1.136 1.142 0.969 0.926 0.969 1.136 1.247 1.142 1.136 1.228 Fig (9): Pin Power ratio distribution in the lattice for fresh fuel. 101
Table (1). The Lattice Assembly Data model Material Total fuel mass per bundle Temperature Fuel Diameter of pellets Stack length (nominal) Number of rods per bundle Length of bundle Material Density Pressure Tube Temperature Inside diameter Wall thickness Material Density Calandria Tube Temperature Inside diameter Wall thickness Material Density Coolant Temperature Fission bundle power Material Moderator Density Temperature Natural UO 2 18.7 kg 700 o C 12.154 mm 480 mm 37 495 mm Zirconium 6.8775 g/cc 290 o C 103 mm 4.34 mm Zirconium 6.55 g/cc 680 o C 129 mm 1.4 mm 99.75 w/o D 2O 0.804 g/cc 290 o C 0.62 Mw 99.75 w/o D 2O 1.0858 g/cc 68 o C CONCLUSION 1- MCNPX code is used to design computer model to simulate the fuel rod assembly in typical condition of pressurized heavy water reactor. MCNPX code used to calculate the multiplication factor, the variation of flux, power and isotopic compositions with fuel burn up. 2- The results indicated 235 U concentration decreases with burn up while Pu isotope concentrations increase with burn up and the pin power increases in the outer (near the moderator) and the outer ring acts as self shielding for the thermal neutrons, so they decrease first toward the center and increase again, fission rate and finally the pin power decreases also with burn up. REFERENCES (1) B. Rouben, "Basic CANDU Design", University Network for Excellence in Nuclear Engineering, 2005. (2) U.S. Nuclear Industry Capacity Factors (1971-2010)", Nuclear Energy Institute, 2010. (3) CANDU Lifetime Performance, Canadian Nuclear Society, September 30, 2009. (4) I. Fairlie, "Tritium Hazard Report: Pollution and Radiation Risk from Canadian Nuclear Facilities", Greenpeace, June 2007. (5) J. Whitlock, "NPD Historical Plaque", Canadian Nuclear Society, 22 February 2002. (6) J. Gibbons, "Darlington Re-Build Consumer Protection Plan", Ontario Clear Air Alliance, pg. 3, 23 September 2010. 102
(7) Ontario s Stranded Nuclear Debt: A Cautionary Tale", Ontario Clean Air Alliance, 1March 2011. (8) CANDU-6 Technical Summary, Atomic Energy of Canada Limited (AECL), Canada. May 2005. (9) IAEA, In-core fuel management benchmarks for PHERs, VIENNA, June 1996. (10) MCNP, A General Monte Carlo N-Particle Transport Code, Version 5, Los Alamos National Laboratory, 2003. (11) MCNPX, Version 26E, Los Alamos National Laboratory, November 17, 2007. (12) A.A. EL-Khawlani, Moustafa Aziz and A. Y. Ellithi, Analysis of Neutronic Parameters for Supercell of CANDU Reactor using MCNPX Code, Journal of Materials Science and Eng. B3 (8) (2013) 550-553. (13) J.R. Lamarsh, Introduction to nuclear engineering, Addison-Wesley publishing company, New York, January 1975. 103