Development of Multigroup Cross Section Generation Code MC 2-3 for Fast Reactor Analysis

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1 Development o Multigroup Cross Section Generation Code MC 2-3 or Fast Reactor Analysis International Conerence on Fast Reactors and Related Fuel Cycles December 7-11, 2009 Kyoto, Japan Changho Lee and Won Sik Yang Nuclear Engineering Division, U.S.A.

2 Background Under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) o U.S. DOE, an integrated, advanced neutronics code system is being developed to allow the high idelity description o a nuclear reactor and simpliy the multi-step design process Development o UNIC with unstructured inite element mesh capabilities on a large scale o parallel computation environment Integration with thermal-hydraulics and structural mechanics calculations As part o this eort, an advanced multigroup cross section generation code named MC 2-3 is being developed The ANL multigroup generation code system, ETOE-2 / MC 2-2 / SDX, has been successully used or ast reactor analysis Recent studies with the ENDF/B-VII.0 data identiied some improvement needs o MC 2-2 Increased importance o resolved resonances in the ENDF/B-VII.0 data due to the extended upper energy cuto and signiicantly increased number o resolved resonances required the use o RABANL or a rigorous treatment o resolved resonances Use o RABANL is limited to the relatively low energy range where the isotropic source approximation is valid 2

3 ETOE-2 / MC 2-2 / SDX ETOE-2 Generate MC 2 libraries by processing ENDF/B data, including ultraine group smooth cross sections (2,082 groups with constant lethargy rom 20 MeV to 0.4 ev) Screen out wide resonances to smooth cross sections Convert the resolved resonances in the Reich-Moore ormalism to those in the multipole ormalism MC 2-2 SDX Sel-shield unresolved and resolved resonances using the generalized resonance integral method based on the narrow resonance (NR) approximation Perorm the consistent P1 or B1 transport spectrum calculations Multigroup method or above resolved resonance energy range Continuous slowing down method or the resolved resonance energy range RABANL option or the hyperine group slowing-down calculation based on isotropic elastic scattering (applicable below ~tens kev) Perorm the 1D integral transport calculation to account or the local heterogeneity eects 3

4 MC 2-2/SDX vs. MC 2-3 4

5 Changes and Improvements in MC 2-3 Numerical integration o resolved resonances with pointwise cross sections based on the NR approximation Reconstruction o pointwise cross sections with Doppler broadening Optionally, use o PENDF iles rom NJOY Multigroup spectrum calculation with the consistent P 1 transport equation or the entire energy range New capability o treating anisotropic inelastic scattering Sel-shielding o resonance-like cross sections above the resonance energy or intermediate-weight nuclides (Fe, Cr, Ni, etc.) 1D transport calculation with ultraine (2082) or user-deined groups (SDX capability) 1D hyperine (> ~100,000) group transport calculation MOC solver with higher-order anisotropic scattering in the LS and CMS (up to ~1 MeV) * i ( u u) i 1 ug u N i g ψ l ( u) σ s ( u) e Pl ( µ s ) i i sl g g = du du n * u n u Pn c ψ + g 1 ug 1 lg (1 αi ) n= 0 σ ( ) (2 1) ( ) ( µ ) Inline cross section generation as a module o UNIC Standalone version or conventional multi-step analyses FORTRAN 90/95 memory structure 5

6 Critical Experiments k in pcm rom Monte Carlo results Z P P R L 1 6 ( R Z ) Z P P R L 2 0 ( R Z ) Z P P R D Z P P R E Z P P R F Z P P R L 1 5 ( R Z ) Z P P R A Z P P R B Z P P R C Z P R MCC-2 MCC-3 k (p cm F l a t t o p F l a t t o p - P u F l a t t o p G o d i v a J e z e b e l - P u J e z e b e l J e z e b e l B i g t e n Z P R A

7 C/E o Fission Reaction Rate Ratios or LANL Assemblies Assembly GODIVA JEZEBEL JEZEBEL -23 C/E FLATTOP -25 C/E FLATTOP -Pu FLATTOP -23 C/E U 235 Data U Np 235 σ / σ U U 235 σ / σ U Pu 235 σ / σ U σ / σ Experiment ± ± ± ±0.025 C/E MCNP a) MC 2-3 b) Experiment ± ± ± ±0.013 C/E MCNP MC Experiment ± ±0.015 MCNP MC Experiment ± ± ± ±0.012 MCNP MC Experiment ± ±0.012 MCNP C/E MC Experiment ± ±0.013 MCNP MC

8 C/E o Fission Reaction Rate Ratios or LANL Assemblies 1.1 MC2-3 MCNP 1.1 U-238 / U-235 Pu-239 / U-235 MC2-3 MCNP C/E 0.9 G o d i v a F l a t t o p J e z e b e l J e z e b e l F l a t t o p - P u F l a t t o p Go d i v a F l a t t o p J e z e b e l 1.1 MC2-3 MCNP 1.0 C/E C/E Np-237 / U MC2-3 MCNP U-233 / U C/E Go d i v a F l a t t o p J e z e b e l J e z e b e l F l a t t o p - P u F l a t t o p Go d i v a F l a t t o p J e z e b e l

9 Hyperine-Group Spectrum Calculation Inner core composition o ZPR-6/6A No rm alized F lu x E+05 Energy (ev) N o rm a l ize d F lu x Hyper FG Ultra FG Ultra FG (NR lux) Hyper FG 1.E E E+03 1.E+04 1.E+05 1.E+06 Energy (ev) 9

10 Ultraine and Hyperine Group Spectrum Calculation with Anisotropic Scattering Sources Normalized Flux UFG Calc. w/ Ansiotropic Scattering HFG Calc. w/ Anisotropic Scattering HFG Calc. w/o Anisotropic Scattering E+03 1.E+04 1.E+05 1.E Energy (ev) Normalized Flux E+05 1.E+06 Energy (ev) 10

11 ZPPR-15A Critical Experiments Relector Blanket Outer Core Inner Core Loading Experiment VIM M C 2-2 M C y (height) z (length) * Uncertainty: Experiment < ±0.0018, VIM < ± x (width) 11

12 ZPR-6 Critical Experiments A ull core heterogeneous reactor calculations with explicit uel plate representation 50,000,000 vertices (~equivalent to 200 million PARTISN inite dierence cells) 200+ angles with P 5 anisotropic scattering 9, 33, 70, and 230 groups No thermal-hydraulics considerations (i.e. clean comparison with MCNP/VIM) Plate by Plate ZPR Geometry 12

13 UNIC Results with MC 2-3 Cross Sections Homogeneous cell cross sections with MC 2-3 without the heterogeneity eect o uel drawers Energy Group K-eective VIM : ± k pcm Power Distribution Cell-averaged cross sections with the 1D slab transport calculation o MC 2-3 to account or the heterogeneity eect o uel drawers Energy Group K-eective k pcm VIM ±

14 Summary New multigroup cross section generation code MC 2-3 has been developed with improved methods Veriication tests with LANL, ZPR-6, ZPPR-15A, ZPPR-21, and BFS critical experiments showed more rigorous and accurate solutions compared to MC 2-2 / SDX 1D hyperine-group transport calculation capability with higher-order anisotropic scattering sources is near completion Initial integration o MC 2-3 into UNIC or inline cross section generation was accomplished Development o eicient algorithms or inline multigroup cross section generation is in progress 14

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