COMPARATIVE ANALYSIS OF WWER-440 REACTOR CORE WITH PARCS/HELIOS AND PARCS/SERPENT CODES

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1 COMPARATIVE ANALYSIS OF WWER-440 REACTOR CORE WITH PARCS/HELIOS AND PARCS/SERPENT CODES S. Bznuni, A. Amirjanyan, N. Baghdasaryan Nuclear and Radiation Safety Center Yerevan, Armenia G. Baiocco, A. Petruzzi Nuclear and Industrial Engineering (NINE), Borgo Giannotti 19 (LU), Italy P. Kohut Brookhaven National Laboratory Upton, NY, USA J. Ramsey Nuclear Regulatory Commission, Rockville, Maryland, USA Abstract Currently predominantly deterministic 2D lattice codes (HELIOS, CASMO, WIMS) are used for few group macroscopic cross-section libraries generation for further reactor core 3D static and dynamic performance analysis by nodal diffusion codes, like PARCS, DYN3D etc. However recent developments in high performance computing and new generation of Monte Carlo codes makes it possible to get rid of all shortcomings of 2D lattice codes and generate crosssection using 3D geometry of fuel assembly. In this work the homogenized constants were applied to the WWER-440 core model developed by PARCS, using the deterministic lattice code HELIOS and the Monte Carlo code Serpent. Several parameters, such as k eff, axial and radial power were compared. 1. INTRODUCTION 2D deterministic lattice codes like HELIOS, CASMO, WIMS, TRITON are the major mature analytical tools widely used both by industrial and regulatory communities for generation of the few-group macroscopic cross-sections and associated neutronics data (yields, fractions of delayed neutrons, precursor decay constants, inverse velocities etc.) for nodal diffusion codes. Availability of high performance computational infrastructure, capacity of handling big memory footprints and massive parallelisation feature of Monte Carlo method allowed to create new generation of Monte Carlo codes for reactor physics applications that have some fundamental advantages over deterministic lattice codes: use of continuous energy neutron-nucleus interaction cross-section libraries instead of multi-group cross-section libraries in deterministic codes. This is a big advantage, because the self-shielding effects are automatically accounted for, without relying on various approximations used by deterministic codes [8]; no-geometric limitations, allowing modelling of any 2D and 3D fuel assembly, group of assemblies (super-cell), reflector or entire reactor core; more precise modelling of the isotopic composition, especially close to the reflector regions. In deterministic analyses to calculate isotopic compositions in the fuel assembly due to axial burnup distributions, set of 2-D calculations are carried out for discrete burnup values. However, this method is acceptable only for the fuel nodes far from axial reflectors where there is no significant axial gradient of the neutron flux. Moreover, the neutron flux spectrum is softened near to the top and bottom nodes of the fuel assembly, due to predominantly leakage of fast neutrons and reflecting back of thermal neutrons, and this effect is not captured by 2D codes [1]. 1

2 IAEA-CN-96 To evaluate maturity of the Monte Carlo reactor physics codes 2D deterministic HELIOS-2 [2] and 3D Monte Carlo Serpent codes [3] were used for the generation of the 2-group macroscopic cross-sections library and associated neutronics data for WWER-440 fuel. Selection of the HELIOS-2 code is due to its wide verification and validation for WWER type of fuel by both us [4] and internationally [5]. Cross-sections generated by HELIOS and Serpent were applied to the WWER-440 reactor core model. 2. COMPARATIVE ANALYSIS OF THE CROSS-SECTIONS As mentioned earlier Monte Carlo reactor physics codes requires high computational power and big memory footprint, therefore the burnup of the WWER-440 fuel assembly was limited up to 20 MWt*day/kgU with following discretization: 0, 0.15, 1, 3, 6, 10, 12, 14, 16, 18, 20. To properly take into account axial moderator density gradient influence on the cross-sections 3 history cases were set up corresponding to the WWER-440 reactor core coolant inlet, average and outlet temperatures at steady state operation (see Table 2.1). For each history 27 branching points were used (see Table 2.1): TABLE 2.1. Cross-sections parametrisation Moderator density, g/cm 3 Boron concentration, Fuel ppm temperature, History K Moderator temperature, K Brunches

3 In the HELIOS-2 (version ) model 1/12 symmetry of the WWER-440 fuel assembly with uniform 3.6% enrichment was taken into account (see Fig. 2.1). The CCCP (current coupling and collision probabilities) neutron transport equation solution method of HELIOS-2 was used in the model. 49 group neutron and 18 group gamma group cross-sections library [2] was used in HELIOS calculations. FIG WWER-440 fuel assembly model developed by HELIOS-2 code Fuel assembly model consists of space elements that are coupled with each other and with the boundaries by interface currents, while the properties of each space element are obtained independently. Furthermore, space elements subdivided into flat-flux regions and their periphery subdivided into flat-current straight-line and arc segments. Inside these regions, the cross-sections are assumed to be constant. k=3 coupling order was used in the model which is the best practice among WWER HELIOS community [6]. In Serpent, a full model of the WWER-440 fuel assembly was developed (see Fig. 2.2). A total of 891 Monte Carlo criticality simulations were run, using the automated burnup sequence of Serpent, with twenty million neutron histories per case. FIG WWER-440 fuel assembly model developed by Serpent code The Serpent calculations were run using the continuous energy ACE format cross-section library based on ENDF/B-VII data files. The B1 methodology was adopted in order to obtain cross-sections consistent with HELIOS. In this work version of Serpent was used for the two-group cross-section data generation. In both models a) the thermal cutoff energy is 1.84 ev to ignore upscattering; b) fission yields and delayed neutron data were calculated according to the recommendations of the [7] to be consistent with PARCS code; c) the cross-sections were generated using reflective boundary conditions. In the Table 2.3, as an example, differences between HELIOS-2 and Serpent calculated B1-corrected k inf values for Hystory-1 reference case are shown. The maximum difference is 250 pcm. For brunches difference varies between 0.67pcm to 587pcm. TABLE 2.3. Differences in k inf Burnup, B1_Kinf-Serpent B1_Kinf-Helios Difference, pcm MWt*d/kgU

4 IAEA-CN Fig show percentage of relative differences between HELIOS and Serpent calculated 2-group transport, absorption, fission, ν-fission, κ-fission macroscopic, Xe and Sm microscopic cross sections and ADFs at different burnups for Hystory-1 reference case. FIG Percentage of relative differences between HELIOS and Serpent calculated cross-sections and ADF at 0, 0.15, 1, 3, 6 MWt*d/kgU burnups

5 FIG Percentage of relative differences between HELIOS and Serpent calculated cross-sections and ADF at 10, 12, 14, 16, 18, 20 MWt*d/kgU burnups As we can see, for the cross-sections relative difference varies within 1-6%. Differences in the transport crosssections is due to simplified approach (out-scatter approximation) [8] used in Serpent in calculating transport cross-section which is prone to significant errors for light nuclides like, hydrogen bound in water. For macroscopic absorption, ν-fission, κ-fission, fission and microscopic Xe and Sm cross-sections there is clear tendency of increasing discrepancies between HELIOS and SERPENT with increasing of the burnup. Here potential reason could be absence of leakage (B1) correction to the transmutation cross sections during burnup calculation [8]. This could significantly influence on isotopic composition of burned fuel since fuel depletion is strongly dependent on spectral changes. k inf differences could be explained by approximations [8] used in B1 leakage correction in case of using reflective boundary condition in Serpent as well as by above mentioned reasons. Fig. 2.5 shows percentage of relative differences between HELIOS and Serpent calculated inverse velocities, fission yields of I, Xe and Sm as well as delayed neutron fractions and precursors decay constants. FIG Percentage of relative differences between HELIOS and Serpent calculated auxiliary neutronics data Differences in delayed neutron fractions and precursors decay constants mainly are due to methodological differences of their calculations. It s obvious that application of Serpent for transient analysis could bring significant errors due to big discrepancies, especially for Xe fission yield and delayed neutron data. 3. TEST OF CROSS-SECTIONS WITHIN FULL CORE MODEL HELIOS and SERPENT cross sections were applied to the WWER-440 reactor core model. Since WWER-440 core has -degree rotational symmetry in the PARCS [9] model fuel types (59 fuel and 2 reflector types) were modeled. Fuel types radial configuration is presented in Fig

6 IAEA-CN-96 FIG Assembly types used in ANPP core model White nodes correspond to fuel assemblies, yellow and red nodes to control assemblies, blue nodes to reflector. Each of fuel types axially subdivided into 43 axial nodes (compositions). Top and bottom axial nodes belong to corresponding top and bottom reflectors. PARCS model coupled to the RELAP model via PVM. The coupling of the PARCS neutronic nodes to RELAP thermal-hydraulic nodes was accomplished through the assignment of mapping weights between the respective nodes. These mapping weights, with values between 0 and 1, determine the distribution of neutronic power in the thermal-hydraulic and heat structure components, as well as the calculation of thermohydraulic feedback in the neutronic nodes. Mapping weight consists from radial and axial weights. Since in the RELAP and PARCS core models the same radial nodalization has been used for each axial level, in the core radial weights are equal to 1 for all nodes. Taking into account axial nodalization of PARCS and RELAP models axial weights were calculated as each RELAP node coupled to 4.1 PARCS nodes. For top and bottom reflectors, axial weights equal to 1 has been used. Entire set of mapping weights is constructed by multiplying each radial weight by each axial weight. Radial burnup distribution of the core is shown in the Fig FIG. 3.2 Radial burnup distribution of the core PARCS calculated k-eff values using HELIOS and SERPENT generated cross-sections are in good agreement: difference is 1.7 pcm. 4. CONCLUSIONS Helios and Serpent 2-group cross sections shows fair agreement, so SERPENT generated could be used for WWER-440 reactor core steady state analysis. However due to large discrepancies in delayed neutron data and fission yields of important poisons application of the Serpent cross-sections for RIA analysis could be questionable. Full core calculations have shown good agreement on k eff. 5. REFERENCES [1] DeHart M., Gaulda I., Suyamab K., Three-dimensional depletion analysis of the axial end of a Takahama fuel rod, International Conference on the Physics of Reactors Nuclear Power: A Sustainable Resource, Casino- Kursaal Conference Center, Interlaken, Switzerland (2008). [2] HELIOS-Methods, Studswik-Scandpower (2010). [3] Aufiero, M., Cammi A., Fiorina C., Leppänen, J. Luzzi L., Ricotti, M. An extended version of the Serpent-2 code to investigate fuel burn-up and core material evolution of the molten salt fast reactor, J. Nucl. Mat., 441 (2013)

7 [4] Bznuni S., Verification and validation of WWER-440 reactor fuel, control assembly and reflector cross-sections library generated by HELIOS-2 code, BNL-09_01, Nuclear and Radiation Safety Center, Yerevan, Armenia, 2011 [5] Simeonov T., HELIOS2: Benchmarking Against Experiments for Hexagonal and Square Lattices, 19th AER Symposium on VVER Reactor Physics and Reactor Safety, Bulgaria (2009) [6] Conclusions of HELIOS-2 training, Yerevan, Armenia (2010). [7] Ward A., Xu Y., Downar T., GenPMAXS v6.1.3, University of Michigan (2015). [8] Leppänen J., Pusa M., Fridman E., Overview of methodology for spatial homogenization in the Serpent 2 Monte Carlo code, Annals of Nuclear Energy, 96 (2016) [9] Downar T., Xu Y., Seker V., Hudson N., PARCS U.S. NRC Core Neutronics Simulator: Theory manual, Department of Nuclear Engineering and Radiological Sciences University of Michigan, USA (2015) 7