A Comparison of MCNPX/WIMS-D5 Burnup Calculation with SAS2H/KENO-v for the IRIS Reactor

Transcription

1 A Comparison of MCNPX/WIMS-D5 Burnup Calculation with SAS2H/KENO-v for the IRIS Reactor E. A. Amin a, I. I. Bashter b, N. M.Hassan b, and S. S. Mustafa b. a Nuclear & Radiological Regulatory Authority, Cairo, Egypt b Faculty of Science, Zagazig University, Zagazig, Egypt Received: 5/2/16 Accepted: 18/4/16 ABSTRACT International Reactor Innovative and Secure (IRIS) reactor is a compact power reactor designed with special features. It contains Integral Fuel Burnable Absorber (IFBA). The core is heterogeneous both axially and radially. This work provides the full core burn up analysis for IRIS reactor using MCNPX and WIMDS-D5 codes. Criticality calculations, radial and axial power distributions and nuclear peaking factor at the different stages of burnup were studied. Effective multiplication factor values for the core were estimated by coupling MCNPX code with WIMS-D5 code and compared with values at different stages of burnup. The two calculation codes show good agreement and correlation. The values of radial and axial powers for the full core were also compared with published results given by (at the beginning and end of reactor operation). The behavior of both radial and axial power distribution is quiet similar to the other data published by. The peaking factor values estimated in the present work are close to its values calculated by. Keywords: IRIS Core, Axial Power, Radial Power, Peaking Factor, MCNPX 1. INTRODUCTION IRIS reactor (International Reactor Innovative and Secure) is an integral pressurized water reactor designed to generate a 1000 MW thermal power. This type of reactors belongs to the Generation IV nuclear reactors which have specific considerations in their design as proliferation resistance, enhanced safety, economic competitiveness and reduction of radiation wastes. The uranium enrichment and boron distribution in the reactor core are two important factors which play an important role in determining the k-eff value and the power distribution shape. The uranium enrichment used is 4.95 % for central assemblies and 2.6% for peripheral fuel assemblies. The Boron- 10 used is in the form of a thin layer ZrB 2 coating on fuel pellets. The presence of this layer of ZrB 2 (IFBA) around the fuel rods in some core assemblies leads to the flattening of power distribution shape inside the core. The reactor core calculations have been carried out using the MCNPX Monte Carlo code (1-5). Several researchers studied the burn up calculations of IRIS reactor core. Ječmenica et al. (6) calculated the effective multiplication factor by using three-dimensional Monte Carlo computer code (KENO-VI) of CSAS26 sequence of SCALE-4.4 code system. They compared their effective multiplication factor value with the effective multiplication factors achieved by HELIOS/NESTLE, CASMO/SIMULATE, and modified CORD-2 nodal calculations. A good agreement was verified between the results of the above codes. Pevec et al. (7) performed the IRIS Core design calculations using the modified CORD-2 code package. They had also used WIMSD-5B code for cell and cluster calculations with two different 69-

2 group data libraries (ENDF/BVI and JEF- 2.2).Their results were compared with the results obtained from the same code (CORD-2) before the modification. Milosevic et al. (8) had setup a new SAS2H/KENO-V sequence for 3D full core depletion analysis using ENDF/B-V library of 238 energy groups. They had performed burnup analysis for IRIS core benchmark and many reactor parameters had been calculated as K-eff and power distribution. Milosevic et al. (9) used the SAS2H/KENO-V methodology for 3D full core depletion analysis and illustrated its capabilities by applying it to burnup analysis of the IRIS core benchmarks. This new SAS2H/KENO-V sequence combines a 3D Monte Carlo full core calculation of node power distribution and a 1D Wigner-Seitz equivalent cell transport method for independent depletion calculation of each of the nodes using ENDF/B-IV library of 27 energy groups. The SAS2H/KENO-V results for the asymmetric IRIS core benchmark were in a good agreement with the results obtained from the ALPHA/PHOENIX/ANC code system. Wollaber (10) had studied the effect of burnable poisons as zirconium di boride and erbium on the power distribution of IRIS reactor core using HELIOS methodology. The calculations were carried out for pin cell model, assembly model and the full core. K-eff, axial power, radial power, peaking factor and axial offset percentage were also calculated. In the present work, the capability of MCNPX with WIMS-D5 codes is used for burnup calculations of IRIS reactor benchmark 44 (8, 9) and studying many parameters as radial power, axial power, k-eff and peaking factor versus different time steps. The physical analysis of each calculated parameter is shown in the section of results and discussions. 2. The IRIS BENCHMARK REACTOR CORE The IRIS reactor core consists of 89 fuel assemblies (FAs). Each fuel assembly contains 4 rods arranged in matrix. The active fuel height is cm with axially uniform enrichment (either 2.6 or 4.95 w% U-235). The total core height, including top and bottom axial reflector regions, is cm. Radial reflector was modeled using reflector cells of the same dimensions as FA. The reflector is a homogeneous mixture of 50% water and 50% stainless steel. It plays an important role in reflecting a large number of neutrons to the reactor core. Some fuel assemblies contain Integral Fuel Burnable Absorbers (IFBA) in form of a thin layer ZrB 2 coating on fuel pellets; the loading of B-10 in these fuel assemblies is controlled by the concentration of B-10 in mg per cm of the fuel rod. The IFBA layer covers the central part of the fuel rod cm. This means that fuel rod is axially composed of: cm enriched (4.95 w% U-235) uncoated, cm enriched (4.95 w/o U235) coated, and cm enriched (4.95 w% U-235) uncoated UO 2 (8-12). The arrangement of core assemblies and non-uniform axial boron distribution are illustrated via horizontal and vertical cut of the IRIS Benchmark 44 core in figures (1and 2) respectively. 3. METHODOLOGY AND MATERIALS MCNPX (MCNP extended) is a Fortran90 (F90) Monte Carlo radiation transport computer code that transports nearly all particles at nearly all energies. The version MCNPX has been used for all the core calculations using ENDF71X library (13). WIMS-D5 code (Winfrith Improved Multigroup Scheme Code System) is a general lattice cell program which uses the transport theory to calculate the atom densities and masses of all fuel and poison isotopes at different time steps (14). The idea of coupling MCNPX with WIMS-D5 is based on calculating the concentrations of fuel isotopes and burnable poison at the required irradiation value of the full core. Then the new atom densities are transferred to MCNPX file to calculate the desired parameters as k-eff and power distribution at the stage under study.

3 Fig. (1): Horizontal and vertical cut of the IRIS Benchmark 44 core 4.95%+1.5 IFBA 4.95%+1.2 IFBA 4.95%+0.9 IFBA 4.95%+0.6 IFBA 4.95%+NO IFBA 2.6%+NO IFBA 4.95%+0.5 IFBA 4.95%+0.75 IFBA 4.95%+1 IFBA 4.95%+1.25 IFBA Reflector Fig. (2): Vertical cut of the IRIS Benchmark 44 core and assemblies enrichment

4 The fuel burnup analysis for IRIS cores is performed via assuming a core average linear power (.7107 MW/tHM) which corresponds to a total power of 1000 MWt (12). MCNPX and WIMS-D5 calculations were done at the following burnup stages: 150, 6650, 12650, 24650, and MWd/tHM. Moreover, the MCNPX code has been used for all reactor core calculations by normalizing all the power results to the reactor power of (10 9 watt) is divided by the sum of power results of 89 assemblies. The calculations are carried out at 600 active cycles, 150 skipped cycles and 10 5 neutron histories. The core is divided into 16 radial nodes and 14 equal axial nodes. For axial power calculation, 14 equal axial nodes are used for the stages 150 and MWd/tHM. For the other burnup stages, 4 axial equal nodes are used in the axial power distribution and comparing the results with other published results for the same number of axial nodes. 3- RESULTS AND DISCUSSION In this part, the power distribution results at different irradiation values are compared with other published results performed by the. At 150 MWd/ton, the values of powers in MW/ton HM of the six radial assemblies corresponding to the radial distance are plotted in figure (3). The radial distance for each assembly equals cm. At this stage, there is a remarkable decrease in the value of power for the center core assembly due to the high content of boron i.e. the depleting of B-10 is not remarkable (the depleting at 7.24 days). The fourth assembly of the highest enrichment (4.95%+ NO IFBA) has the highest power which is nearly twice the power value of the sixth assembly of enrichment (2.6% +NO IFBA). The power distribution shaped decreased at the sixth assembly due to the presence of leakage neutrons and the low enrichment of this assembly (2.6% U-235). Figure( 4) shows the axial power at the first stage. The power values are plotted for 14 equal axial nodes. It is noted that the power decreases from 30 cm to 213 cm due to previous mentioned reason (high load of boron 1.5 mg/cm). A gradual increase from 213 cm to 396 cm is noticed due to the lower content of boron for this segment (1.25 mg/cm).the power has a maximum value at the segment beginning from 396 to 426 because of the high enrichment of U-235 (4.95%) without boron- 10. Core average radial power (MW/ton HM) Radius (cm) Fig. (3): Radial power distributions at 150 MWd/tHM (first stage) for present work by MCNPX code and

5 Core average axial power (MW/ton HM) Height (cm) Fig. (4): Axial power distributions at 150 MWd/tHM (first stage) for present work by MCNPX code and for 14 equal axial nodes Figures (5 and 6) show the second stage of burnup, at 6650 MWd/tHM. The Power values of the radial assemblies increase as the radial distance increases till the fourth assembly. This is due to the remarkable depleting of boron (depleting at 321 days) and the power values start to approach to each other. The power decreases at the fifth and sixth assemblies due to depleting high amount of U- 235 besides the low enrichment of the sixth assembly (2.6% U-235). For the axial power (Fig. 6) a slight increase for the power from 30 cm to 150 cm because this segment represents the central axial part of the core which corresponds to the radial assembly. A good agreement is found between the present investigation and the published work. Core average radial power (MW/ton HM) Radius (cm) Fig. (5): Radial power distributions at 6650 MWd/tHM (second stage) for the present work by MCNPX code and

6 Core average axial power (MW/ton HM) Height (cm) Fig. (6): Axial power distributions at 6650 MWd/tHM (second stage) for the present work by MCNPX code and for 4 equal axial nodes For the third stage of burnup, MWd/tHM (Figs 7& 8). As the time passes, boron-10 is burned more and the radial power shape begins to be flattened. For the axial part for this stage, the value of power from 30 cm to 213 cm is higher than the value of power from 213 cm to 396. This means that the B-10 is obviously depleted in the central axial part (depleting at 610 days).table 1 shows the two code results with percent error for radial power results. Table 2 shows the two code results with error percentage for axial power results Table (1): Comparison between MCNPX and KENO radial power results at MWd/ton Radius KENO MCNPX Percent error = (KENO- MCNPX)/KENO*100%

7 Core average radial power (MW/ton HM) Radius (cm) Fig. (7): Radial power distributions at MWd/tHM (third stage) for the present work by MCNPX code and Core average axial power (MW/ton HM) Height (cm) Fig. (8): Axial power distributions at MWd/tHM (third stage) for the present work by MCNPX code and for 4 equal axial nodes Table (2): Comparison between MCNPX and KENO axial power results at MWd/ton Height MCNPX KENO Percent error = (KENO- MCNPX)/KENO*100%

8 Figures (9 and 10) show the fourth stage at MWd/tHM. For radial power distributions, it is clear that there is an increase in the value of power of central core assembly. Core average radial power (MW/ton HM) Radius (cm) Fig. (9): Radial power distributions at MWd/tHM (fourth stage) for the present work by MCNPX code and Core average axial power (MW/ton HM) Height (cm) Fig. (10): Axial power distributions at MWd/tHM (fourth stage) for the present work by MCNPX code and for 4 equal axial nodes At a burnup of MWd/tHM (fifth stage), the core power center decreases which causes flattening of power again (Figs 11&12). The axial power values at (30650 MWd/tHM) for the two middle nodes in the present work are closer to each other than those of the same two middle nodes performed by the SAS2H/KENO-V. This shows that the flattening of axial power by MCNPX code is also clearer and better than that of the due to the point cross section (ENDF71X) used by MCNPX.

9 Core average radial power (MW/ton HM) Radius (cm) Fig. (11): Radial power distribution at MWd/tHM (fifth stage) for the present work by MCNPX code and Core average axial power (MW/ton HM) 160 SAS2H/KENO-V Height (cm) Fig. (12): Axial power distribution at MWd/tHM (fifth stage) for the present work by MCNPX code and for 4 equal axial nodes. At the end of cycle, MWd/tHM (sixth stage), the radial and axial power distribution deeps at the center; this means that the fuel burnup at the core periphery exceeds the fuel burnup at the center. Figures (13&14) show that behavior 89

10 Core average radial power (MW/ton HM) Radius (cm) Fig. (13): Radial power distribution at38650mwd/thm (sixth stage) for the present work by MCNPX code and Core average axial power (MW/ton HM) Height (cm) Fig. (14): Axial power distribution at MWd/tHM (sixth stage) for the present work by MCNPX code and for 14 equal axial nodes. The differences between burnup results obtained by coupling with MCNPX with WIM-D5 codes and results by the are expected due to the following reasons, (1) The method of reflector modeling and reflector thickness. (2) Probability of using different cross section libraries while, the present work used point cross section (ENDF71X), the used 238 multigroup library generated from ENDF/B-V. (3) The method of calculation in the present work depends on the Monte Carlo method, but is a probabilistic Monte Carlo methodology with multi scheme energy groups. (4) Modeling of ZrB 2 by depends on smearing boron with clad, but in the present work, ZrB2 is modeled as a separate ring coating the fuel, of certain radius according to the concentration of boron-10 in mg/cm. 89

11 The variation of K-eff with burnup for full core (Fig 15) is due to main factors (1) the depletion of U-235 which provides negative reactivity (2) the depletion of boron which provides positive reactivity. The contribution of both factors gives the variation of K-eff with time. For the variation of power peaking factor with burnup (Fig 16), at the beginning, at 150 MWd/ton, it has a high value. This is due to the high value of power of the assembly which has 4.95% U-235 and no boron. Since the peaking factor is defined as the ratio between the maximum rod power to the average core power, then peaking will be higher for the highest enriched assembly. As the burnup proceeds, the value of power of highest enriched assembly drops from 41 MW/t HM at 150 MWd/tHM to nearly 32 MW/tHM at 6650 MWd/tHM. This will be accompanied by decreasing the value of peaking factor especially at nearly 8 MWd/tHM. From MWd/tHM, the peaking factor begins to increase till it has a maximum value at MWd/tHM. This is due to increasing the power of core central assembly. After MWd/tHM, the peaking decreases with burnup due to the deeping of power at the center at the end of the cycle of the reactor operation. Effective multiplication factor Burnup (MWd/tHM) Fig. (15): Effective multiplication factor versus different stages of burnup for IRIS full core performed by MCNPX and WIMS-D5 codes Peaking factor Burnup (MWd/tHM) Fig. (16): Total peaking factor against different stages of burnup for IRIS full core performed by MCNPX and WIMS-D5 codes 88

12 CONCLUSION MCNPX and WIMS-D5 codes are used for the depletion analysis of the full IRIS reactor core. From the obtained results, it is obvious that the IRIS reactor operates as a supercritical reactor at the beginning of life. Variation of K-eff with burnup is due to the depletion of U-235(negative reactivity) and B-10 (positive reactivity). The power peaking factor of this reactor at the start of reactor operation is around 3.5 which is close to the published result (4.5). This high value is expected due to the high power generated from the fifth assembly that contains 4.95% of U-235 only. The peaking factor firstly decreases with time due to the depleting of U-235 in the highest enrichment assembly. Then it decreases due to the high power generated from the central assembly at the last stages of burnup. It is also observed that the power distribution varies widely both radially and axially. The oscillatory behavior of the power distribution described above pertains to a benchmark core and not to a realistic core. A realistic core will be designed to have a much more stable power shape. The benchmark problem was defined to be axially symmetric due to the non-uniform distribution of B-10. Finally, it can be concluded that the present work was very successful in accurately modeling and predicting the different IRIS reactor core parameters for the fresh core as well as the different reported burnup stages. REFERENCES (1) M.D. Carelli, and B. Petrovic, Next Generation Advanced Reactor, Nuclear Plant Journal, 19 (3) (01) (2) M.D. Carelli, IRIS: Generation IV Advanced Light Water Reactor for Countries with Small and Medium Electricity Grids" Proceedings of the 4th International Conference on Nuclear Option in Countries with Small and Medium Electricity Grids, Dubrovnik, Croatia, June 16- (02) 4, 5. (3) M.D Carelli, B. Petrovic, D.V. Paramonov, T. Moore, C.V. Lombardi, M.E.Ricotti; E.Greenspan, J. Vujic, N.E. Todreas, M. Galvin, K. Miller, A. Nagano, K. Yamamoto, H.Ninokata, J. Robertson, F. Oriolo, G. Proto, G. Alonso, and M.M. Moraes, IRIS: An Integrated International Approach to Design and Deploy a New Generation Reactor, IAEA International Seminar on Status and Prospects for Small and Medium Sized Reactors, Cairo, Egypt, May 27-31, (01) 3. (4) M.D. Carelli, B. Petrovic, H. Garkisch, D.V. Paramonov, C. Lombardi, L. Oriani, M. Ricotti; E. Greenspan, and T. Lou Trade-Off Studies for Defining the Characteristics of the IRIS Reactor Core, 16th International Conference on Structural Mechanics in Reactor Technology, Washington, DC, USA, August (01). (5) M.D. Carelli, L. E. Conway, B. Petrovic, D. V. Paramonov, M. Galvin, N. E. Todreas, C.V. Lombardi, F. Maldari, M. E. Ricotti and L. Cinotti, IRIS Reactor Conceptual Design, International Conference On the Back-End of the Fuel Cycle (Global 01), Paris, France, September 9-13 (01) (6) R. Ječmenica, K. Trontl, D. Pevec, D. Grgić, IRIS Core Criticality Calculations, University of Zagreb, Faculty of Electrical Engineering and Computing Unska 3, Zagreb, Croatia,03. (7) D. Pevec, D. Grgic, R. Jecmenica, and B. Petrovic, Core Design Calculations of Iris Reactor Using Modified Cord-2 Code Package, (02). (8) M.Milosevic, E.Greenspan, and J.Vujic, A Sas2h/Keno V Methodology for 3d Depletion Analysis, in Proc. International Conference on the New Frontiers of Nuclear Technology: Reactor Physics, Safety and High-Performance Computing, PHYSOR (02) pp (9) M. Milosevic, E. Greenspan, J.Vujic, and B. Petrovic, A Sas2h/Keno-V Methodology for 3d Full Core Depletion Analysis, (03). (10) A. B. Wollaber, Burnable Poison Design for the International Reactor, Innovative and Secure (Iris), A Thesis Presented for the Master of Science Degree, The University of Tennessee, Knoxville, August (03). 88

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