Mercatali, L.

Sanchez, V.H.

Venturini, A.

English

Sanchez, V. H.

KITopen

Institute for Neutron Physics and Reactor Technology (INR) 1 2010.05.11. KIT – University of the State of Baden-Wuerttemberg and National Research Center of the Helmholtz Association Institute for Neutron Physics and Reactor Technology www.kit.edu                  Monte Carlo neutronics investigations of  VVER-1000 fuel assemblies     L. Mercatali1, A. Venturini2, V.H. Sanchez1  1KIT- Institute for Neutron Physics and Reactor Technology (Germany) 2University of Pisa (Italy)  Annual Meeting on Nuclear Technology (Berlin, May 5-7 2015)    luigi.mercatali@kit.edu  Institute for Neutron Physics and Reactor Technology (INR) 2 L. Mercatali – Annual Meeting on Nuclear Technology (Berlin, May 6, 2015) Outline Introduction & motivation The “OECD VVER-1000 LEU and MOX Assemblies Burnup Computational Benchmark” 1. Specifications 2. Modeling of the benchmark exercises 3. Results  Monte Carlo & deterministic static solutions  Burn-up  kinf  Nuclide concentrations  Pin-by pin fission rates Summary and conclusions Institute for Neutron Physics and Reactor Technology (INR) 3 Introduction & motivation The loading of Mixed Uranium Plutonium OXide (MOX) and Low Enriched Uranium (LEU) fuels in commercial nuclear reactors requires well validated computational methods and codes capable to provide reliable predictions of the neutronics characteristics of such fuels in terms of reactivity conditions (kinf), nuclide inventory and pin power generation over the entire fuel cycle length  Within the framework of Joint United States/Russian Fissile Materials Disposition Program an important task is to verify and validate neutronics codes for the use of MOX fuel in VVER-1000 reactors  In this work new solutions for the (UO2+Gd) and (UO2+PuO2+Gd) fuel assemblies proposed within the “OECD VVER-1000 LEU and MOX Assemblies Burnup Computational Benchmark” have been produced  The SCALE and SERPENT codes have been used with ENDF/B-VII and JEFF3.1 nuclear data libraries (NDLs)  L. Mercatali – Annual Meeting on Nuclear Technology (Berlin, May 6, 2015) Institute for Neutron Physics and Reactor Technology (INR) 4 OECD VVER-1000 burnup computational benchmark – LEU/MOX assemblies The LEU assembly consists of low enriched fuel rods (3.7 wt %) and 12 rods poisoned with 4.0 wt % of gadolinium (3.6 wt % 235U)  The MOX assembly contains fuel rods with three different plutonium loadings: the central region with 4.2 wt % of fissile plutonium (93 wt % 239Pu), two outer rings a 3.0 wt % and the outermost ring a 2.0 wt %   LEU MOX L. Mercatali – Annual Meeting on Nuclear Technology (Berlin, May 6, 2015) Institute for Neutron Physics and Reactor Technology (INR) 5 OECD VVER-1000 benchmark – specifications State Fuel temp.  K Non-fuel temp. K 135Xe,149Sm Boron [ppm] Moderator density [g/cm3] S1 1027 575 Eq. 600 0.7235 S2 1027 575 0.0 600 0.7235 S3 575 575 0.0 600 0.7235 S4 575 575 0.0 0.0 0.7235 S5 300 300 0.0 0.0 1.0033 Constant power density: 108 MW/m3 Requested responses: kinf for state S1 at each burn-up step   kinf for states S2-S5 at 0, 20, 40 MWd/kgHM   Pin-by-pin fission rate distribution   Nuclide assembly average concentrations + nuclide average concentrations in cell 1       and  cell 24 for 235U, 236U, 238U, 239Pu, 240Pu, 241Pu, 242Pu,135Xe,149Sm, 155Gd and 157Gd    S1 - S2:  Poisons (Xe + Sm) effects S2 - S3:  Doppler coefficient S3 - S4:  Boron effect S4 - S5:  Moderator temperature coefficient     Operational states L. Mercatali – Annual Meeting on Nuclear Technology (Berlin, May 6, 2015) Institute for Neutron Physics and Reactor Technology (INR) 6 Modeling assumptions – SCALE 6.1 Transport solutions (SN6) Boundary condition: white Convergence criterion (spatial/eigenvalue): ε=1E-6 Predictor-corrector algorithms Cross-section library: ENDF/B-VII.0   LEU   MOX   L. Mercatali – Annual Meeting on Nuclear Technology (Berlin, May 6, 2015) Institute for Neutron Physics and Reactor Technology (INR) 7 LEU – sensitivity studies on burn-up steps ~100 pcm ~50 pcm ~10 pcm Negligible effect of gadolinium mesh refinement (<20 pcm) L. Mercatali – Annual Meeting on Nuclear Technology (Berlin, May 6, 2015) Institute for Neutron Physics and Reactor Technology (INR) 8 Modeling assumptions – SERPENT 2.1 Solution of Depletion Equations: CRAM Predictor-corrector method (PC) vs Stochastic Implicit Euler (SIE) Pin-by-pin model Boundary condition: reflective Gadolinium refinement 10000 neutrons, 4000 cycles, 100 inactive cycles (Shannon enrtopy criterion) Tmp card (Doppler broadening) vs. Temperature weighted compositions (~200 pcm) Cross-section library: ENDF/B-VII.0, JEFF3.1, JEFF2.2  L. Mercatali – Annual Meeting on Nuclear Technology (Berlin, May 6, 2015) Institute for Neutron Physics and Reactor Technology (INR) 9    SERPENT and MCNP4B results for kinf at zero burnup with JEFF2.2 data  Monte Carlo solutions - kinf Objective: assess the validity of our model through comparison with the                 reference MCNP4B solutions  Overall good agreement  Small deviations due to different data (ENDF/B.V) for 16O, 152Gd, natZr and 1H in the MCNP4B results and different statistics LEU SERPENT (S) MCNP4B (M) Δk  (M-S)·105 S2 1.17987 ± 8.9E-05 1.1800 ± 6E-05 13 S3 1.19365 ± 8.6E-05 1.1925 ± 6E-05 -115 S4 1.25387 ± 8.5E-05 1.2531 ± 7E-05 -77 S5 1.32394 ± 7.6E-05 1.3235 ± 6E-05 -44 MOX SERPENT (S) MCNP4B (M) Δk  (M-S)·105 S2 1.19258 ± 8.6E-05 1.1922 ± 7E-05 -38 S3 1.20919 ± 8.4E-05 1.2091 ± 6E-05 -9 S4 1.24408 ± 8.5E-05 1.2430 ± 6E-05 -108 S5 1.32487 ± 7.6E-05 1.3256 ± 7E-05 73 L. Mercatali – Annual Meeting on Nuclear Technology (Berlin, May 6, 2015) Institute for Neutron Physics and Reactor Technology (INR) 10 Monte-carlo and deterministic solutions  - kinf  LEU SERPENT (SJ) [JEFF3.1] SERPENT (SE) [ENDF/B-VII] SCALE (SCE) [ENDF/B-VII] Δk (SE-SJ)·105 Δk  (SE-SCE)·105 S2 1.17399 ± 8.9E-05 1.17587 ± 8.8E-05 1.17068 188 519 S3 1.18787 ± 8.7E-05 1.18996 ± 8.6E-05 1.18557 209 439 S4 1.24808 ± 8.5E-05 1.24993 ± 8.7E-05 1.24538 185 455 S5 1.32088 ± 7.6E-05 1.32305 ± 7.7E-05 1.31770 217 535 MOX SERPENT (SJ) [JEFF3.1] SERPENT (SE) [ENDF/B-VII] SCALE (SCE) [ENDF/B-VII] Δk (SE-SJ)·105 Δk  (SE-SCE)·105 S2 1.19596 ± 8.5E-05 1.19762 ± 8.6E-05 1.19178 166 584 S3 1.21259 ± 8.4E-05 1.21419 ± 8.4E-05 1.21005 160 414 S4 1.24765 ± 8.4E-05 1.24923 ± 8.4E-05 1.24491 158 432 S5 1.32908 ± 7.6E-05 1.33013 ± 7.6E-05 1.32652 105 361 Systematic under-prediction of JEFF3.1 with respect to JEFF2.2 (LEU) Systematic over-prediction of JEFF3.1 with respect to JEFF2.2 (MOX) ENDF/B-VII-kinf larger with respect to JEFF3.1-kinf by ~100-200 pcm SCALE results are systematically higher than SERPENT ones due to the multi-group and spatial approximations     SERPENT and SCALE results for kinf at zero burnup with different NDLs L. Mercatali – Annual Meeting on Nuclear Technology (Berlin, May 6, 2015) Institute for Neutron Physics and Reactor Technology (INR) 11 Reactivity effects Due to the harder neutron spectrum of the MOX assembly:  Xenon effect lower in MOX, the difference becoming less important with the increasing of burnup  Boron worth reduced in MOX. Difference: ~71% (BOC) - ~24% (EOC) Doppler coefficient slightly more negative for the MOX at BOC and almost equals at EOC Moderator temperature coefficient more negative in the case of MOX   Initial state Final state Burnup (MWd/KgHM) LEU MOX SERPENT SCALE Mean SERPENT SCALE Mean S1 S2 0 -3.58 -* -4.03 -2.63 - -3.33 20 -2.91 - -3.96 -2.44 - -3.44 40 -2.37 - -3.29 -2.16 - -3.08 S2 S3 0 -1.41 -1.48 -1.36 -1.66 -1.83 -1.74 20 -1.48 - -1.52 -1.48 - -1.55 40 -1.35 - -1.40 -1.33 - -1.40 S3 S4 0 -6.00 -5.98 -5.99 -3.50 -3.49 -3.46 20 -5.40 - -5.48 -3.76 - -3.83 40 -4.81 - -4.91 -3.86 - -3.98 S4 S5 0 -7.28 -6.98 -6.90 -8.09 -8.16 -7.93 20 -6.87 - -6.72 -6.98 - -7.03 40 -4.35 - -5.33 -5.49 - -5.41   LEU and MOX assemblies: 100 x (kinit. – kfin.) L. Mercatali – Annual Meeting on Nuclear Technology (Berlin, May 6, 2015) Institute for Neutron Physics and Reactor Technology (INR) 12 Burn-up calculations - kinf The decrease of kinf is slower for MOX Different shapes of the burn-up swings due to different spectra SERPENT- kinf systematically higher than SCALE- kinf  The agreement in between SERPENT/SCALE stands within ±0.39% (LEU) and ±0.33% (MOX) L. Mercatali – Annual Meeting on Nuclear Technology (Berlin, May 6, 2015) Institute for Neutron Physics and Reactor Technology (INR) 13 Burn-up calculations – assembly averaged nuclide concentrations (SERPENT/Benchmark mean)     Excellent agreement  for U      Good agreement for Pu   Slight discrepancies for       Sm-149,  Gd-155 and Gd-157        in the MOX case     L. Mercatali – Annual Meeting on Nuclear Technology (Berlin, May 6, 2015) Institute for Neutron Physics and Reactor Technology (INR) 14    Burn-up calculations – Cell 1: nuclide concentrations (SERPENT/SCALE/Benchmark mean)     Excellent/good agreement  for U and Pu isotopes    Discrepancies for Sm-149 at high burn-up levels    L. Mercatali – Annual Meeting on Nuclear Technology (Berlin, May 6, 2015) Institute for Neutron Physics and Reactor Technology (INR) 15 Burn-up calculations – Cell 24: nuclide concentrations (SERPENT/SCALE/Benchmark mean)     Good agreement       for U and Pu isotopes     Discrepancies         for Xe-135, Sm-149, Gd-155       and Gd-157    L. Mercatali – Annual Meeting on Nuclear Technology (Berlin, May 6, 2015) Institute for Neutron Physics and Reactor Technology (INR) 16 Pin-by-pin fission rates Good agreement SERPENT/SCALE  Maximum deviations: LEU: -4.2% (pin 35) MOX: -1.6% (pin 63)   LEU Assembly 0 MWd/kgHM 20 MWd/kgHM 40 MWd/kgHM Pin number 21 55 1 Max δ [%] (SCALE-SERPENT) -0.78 2.14 3.22 Average δ [%] (SCALE-SERPENT) 0.25 0.69 1.45 MOX Assembly 0 MWd/kgHM 20 MWd/kgHM 40 MWd/kgHM Pin number 10 4 46 Max δ [%] (SCALE-SERPENT) -1.26 3.97 5.04 Average δ [%] (SCALE-SERPENT) 0.48 1.08 1.44 L. Mercatali – Annual Meeting on Nuclear Technology (Berlin, May 6, 2015) Institute for Neutron Physics and Reactor Technology (INR) 17 Summary and conclusions SERPENT and SCALE calculation schemes for VVER-1000 reactors have been tested on 2-D problems for fresh and depleted LEU and MOX fuel assemblies representative of the designs expected for the plutonium disposition mission  Small deviations have been generally observed between the SCALE and SERPENT computed kinf, nuclide concentrations and pin-by-pin fission rates.  Very good agreement is found between SERPENT solutions and the corresponding previous MCNP4B results, making of this work also a new Monte Carlo reference solution for the OECD VVER-1000 LEU and MOX Assemblies Burnup Computational Benchmark with modern NDLs   L. Mercatali – Annual Meeting on Nuclear Technology (Berlin, May 6, 2015) 

Monte Carlo neutronics investigations of VVER-1000 fuel assemblies

https://publikationen.bibliothek.kit.edu/240100682/3816797

Monte Carlo neutronics investigations of VVER-1000 fuel assemblies

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