Abstract
Studsvik has recently extended the CASMO5 advanced lattice physics code for the analysis of VVER 1000 and 1200 reactors. These extensions form the basis of CASMO5-VVER, which is primarily intended to compute homogenized nodal data for SIMULATE5-VVER. CASMO5-VVER leverages the latest nuclear data and numerical methods to VVER analyses. The current CASMO5 data library, based on the ENDF/B-VII.1 nuclear data evaluation, features a 586 energy group structure and nuclear data for hundreds of unique nuclides. Resonance self-shielding, based on the Equivalence Theory and an Optimal Two-Term Rational (OTTR) method, has been extended to support hexagonal geometry. The solution to the two-dimensional transport equation is based on the new Linear Source (LS) approximation for the Method of Characteristics (MOC). The acceleration of the MOC solution is attained through the implementation of a Coarse-Mesh Nonlinear Diffusion (CMND) acceleration. Results from preliminary validation against various critical experiments are presented in this work.
Kurzfassung
Studsvik hat kürzlich den fortgeschrittenen Gittercode CASMO5 für die Analyse von VVER 1000 und 1200 Reaktoren erweitert. Diese Erweiterungen bilden die Grundlage von CASMO5-VVER, das in erster Linie zur Berechnung homogenisierter nodaler Daten für SIMULATE5-VVER gedacht ist. CASMO5-VVER nutzt die neuesten nuklearen Daten und numerischen Methoden für die VVER-Analyse. Die aktuelle CASMO5-Datenbibliothek, die auf den ENDF/B-VII.1 Kerndaten basiert, verfügt über eine 586 Energiegruppenstruktur und Kerndaten für Hunderte von einzelnen Nukliden. Die Resonanz-Selbstabschirmung, basierend auf der Äquivalenztheorie und einer optimalen Zwei-Term-Rational-Methode (OTTR), wurde erweitert, um die hexagonale Geometrie zu unterstützen. Die Lösung der zweidimensionalen Transportgleichung basiert auf einer neuen linearen Quellenapproximation für die Charakteristikenmethode (MOC). Die Beschleunigung der MOC-Lösung wird durch die Implementierung einer nichtlinearen Grobgitter Diffusionsbeschleunigung erreicht. Ergebnisse der vorläufigen Validierung anhand verschiedener kritischer Experimente werden in dieser Arbeit vorgestellt.
References
1 Rhodes, J. D.; Smith, K. S.; Lee, D.: CASMO-5 Development and Applications. Proc. Int. Conf. PHYSOR 2006, Vancouver, BC, Canada, September 10–14, 2006 Search in Google Scholar
2 Bahadir, T.: SIMULATE5-HEX Extension for VVER Analyses. Kerntechnik83 (2018) 268–27410.3139/124.110908 Search in Google Scholar
3 Knott, D.; Edenius, M.; PeltonenJ.; Anttila, M.: Results of Modelling Hexagonal and Circular Cluster Fuel Assembly Designs Using CASMO-4. Proc. Int. Conf. Advances in Nuclear Fuel Management II, Myrtle Beach, SC, March 23–26, 1997 Search in Google Scholar
4 Lindhal, S.-O.: SIMULATE-HEX – The Multi-Group Diffusion Equation in Hexagonal-Z Geometry. Proc. Int. Conf. M&C 2013, Sun Valley, ID, May 5–9, 2013 Search in Google Scholar
5 Ferrer, R. M.; Rhodes, J. D.; Li, L.: Implementation of Coarse-Mesh Nonlinear Diffusion Acceleration for Hexagonal Geometry in CASMO5. Proc. Int. Conf. PHYSOR 2016, Sun Valley, ID, May 1–5, 2016 Search in Google Scholar
6 Rhodes, J. D.; Gheorghiu, N.; Ferrer, R. M.: CASMO5 JENDL-4.0 and ENDF/B-VII.1beta4 Libraries. Proc. Int. Conf. PHYSOR 2012, Knoxville, TN, April 15–20, 2012 Search in Google Scholar
7 Ferrer, R. M.; Rhodes, J. D.: A Linear Source Approximation Scheme for the Method of Characteristics. Nuclear Science and Engineering182 (2016) 151–16510.13182/NSE15-6 Search in Google Scholar
8 Hykes, J. M.; Ferrer, R. M.: Solving the Bateman Equations in CASMO5 Using Implicit ODE Numerical Methods for Stiff Systems. Proc. Int. Conf. M&C 2013, Sun Valley, ID, May 5–9, 2013 Search in Google Scholar
9 Haugh, B.: Generic Application of the Studsvik Scandpower Core Management System to Pressurized Water Reactors. Studsvik Scandpower, Inc. technical report SSP-14-P01/028-TR-NP-A (Revision 0) (2017) Search in Google Scholar
10 Chadwick, M. B.; et al.: ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Yields and Decay Data. Nuclear Data Sheets112 (2011) 2878–299610.1016/j.nds.2011.11.002 Search in Google Scholar
11 The JEFF team; JEFF-3.2: Evaluated nuclear data library. URL: http://www.oecdnea.org/dbdata/jeff (2014) Search in Google Scholar
12 Koning, A. J.; Rochman, D.: Modern Nuclear Data Evaluation with the TALYS Code System. Nuclear Data Sheets113 (2012) 2841–293410.1016/j.nds.2012.11.002 Search in Google Scholar
13 MacFarlane, R. E.: The NJOY Nuclear Data Processing System. Version 2012. Los Alamos National Laboratory technical report LA-UR-12-27079 (2012) Search in Google Scholar
14 Hfaiedh, N.; Santamarina, A.: Determination of the Optimized SHEM Mesh for Neutron Transport Calculations. Proc. Int. Conf. M&C 2005, Palais des Papes, Avignon, France, September 12–15, 2005 Search in Google Scholar
15 Hébert, A.; Santamarina, A.: Refinement of the Santamarina-Hfaiedh Energy Mesh between 22.5 eV and 11.4 keV. Proc. Int. Conf. PHYSOR 2008, Interlaken, Switzerland, September 14–19, 2008 Search in Google Scholar
16 Stamm'ler, R. J. J.; Abbate, M. J.: Methods of Steady-state Reactor Physics in Nuclear Design., Academic Press, London (1983) Search in Google Scholar
17 Knott, D.; Yamamoto, A.: Lattice Physics Computations. D. G.Cacuci (Ed.); Handbook of Nuclear Engineering. Springer US, Boston, MA (2010) 10.1007/978-0-387-98149-9_9 Search in Google Scholar
18 Bell, G. I.; Glasstone, S.: Nuclear Reactor Theory. Van Nostrand Reinhold Company, New York (1970) Search in Google Scholar
19 Lee, D.; Smith, K.; Rhodes, J.: The Impact of 238U Resonance Elastic Scattering Approximations on Thermal Reactor Doppler Reactivity. Annals of Nuclear Energy36 (2009) 274–28010.1016/j.anucene.2008.11.026 Search in Google Scholar
20 Choi, S.; Lee, H.; Hong, S.G.; Lee, D.: Resonance Self-Shielding Methodology of New Neutron Transport Code STREAM. J. Nucl. Sci. Technol.52 (2015) 1133–115010.1080/00223131.2014.993738 Search in Google Scholar
21 Goldstein, R.; Cohen, E. R.: Theory of Resonance Absorption of Neutrons. Nuclear Science and Engineering13 (1962) 132–14010.13182/NSE62-1 Search in Google Scholar
22 Yamamoto, A.; et al: Derivation of Optimum Polar Angle Quadrature Set for the Method of Characteristics Based on Approximation Error for the Bickley Function. Journal of Nuclear Science and Technology44 (2007) 129–13610.1080/18811248.2007.9711266 Search in Google Scholar
23 Cho, J.; et al: Whole Core Transport Calculation Employing Hexagonal Modular Ray Tracing and CMFD Formulation. Journal of Nuclear Science and Technology45 (2008) 10.1080/18811248.2008.9711475 Search in Google Scholar
24 Ferrer, R. M.; Rhodes, J. D.: The Linear Source Approximation and Particle Conservation in the Method of Characteristics for Isotropic and Anisotropic Sources. Annals of Nuclear Energy115 (2018) 209–219, 10.1016/j.anucene.2018.01.023 Search in Google Scholar
25 Choi, S.; Smith, K.; Lee, H. C.; Lee, D.: Impact of Inflow Transport Approximation on Light Water Reactor Analysis. Journal of Computational Physics299 (2015) 352–37310.1016/j.jcp.2015.07.005 Search in Google Scholar
26 Smith, K. S.; RhodesIII, J. D.: Full-Core, 2-D, LWR Core Calculations with CASMO-4E. Proc. Int. Conf. PHYSOR 2002, Seoul, Korea, October 7–10, 2002 Search in Google Scholar
27 Krýsl, V.; Mikoláš, P.; Sprinzl, D.; Švarný, J.: Proposal of “FullCore” VVER-1000 Calculation Benchmark. Proc. of the 26th Symposium of AER, Helsinki, Finland, October 10–14, 2016 Search in Google Scholar
28 Yamaji, K.; et al.: Simple and Efficient Parallelization Method for MOC Calculation. Journal of Nuclear Science Technology47 (2010) 90–10210.1080/18811248.2010.9711931 Search in Google Scholar
29 Lewis, E. E.; Miller, W. F.: Computational Methods of Neutron Transport. American Nuclear Society, La Grange Park, IL, USA (1993) Search in Google Scholar
30 Pusa, M.: Rational Approximations to the Matrix Exponential in Burnup Calculations. Nuclear Science and Engineering, 169 (2011) 155–16710.13182/NSE10-81 Search in Google Scholar
31 Hykes, J.; Ferrer, R.: A Quadratic Depletion Coupling Scheme with Adaptive Stepsize Control in CASMO5. Proc. Int. Conf. M&C 2017, Jeju, Korea, April 16–20, 2017 Search in Google Scholar
32 The VVER Experiments: Regular and Perturbed Hexagonal Lattice of Low-Enriched UO2 Fuel Rods in Light Water. NEA/NSC/DOC/(2006)1, ZR6-VVER-EXP-001, Revision 0, March 31, 2007; LEU-COMP-THERM-015, Revision 3, September 30, 2005; LEU-COMP-THERM-036, Revision 1, September 30, 2005 Search in Google Scholar
33 VVER Physics Experiments: Hexagonal (1.27-cm Pitch) Lattices of U(4.4 wt.% 235U)O2 Fuel Rods in Light Water, Perturbed by Boron, Hafnium, or Dysprosium Absorber Rods, or By Water Gap With/Without Empty Aluminium Tubes. NEA/NSC/DOC/(95)03/IV, LEU-COMP-THERM-061, Revision 0, September 30, 2002 Search in Google Scholar
34 Henry, A.: Nuclear-Reactor Analysis. Cambridge, MA, MIT Press (1975) Search in Google Scholar
35 VVER Physics Experiments: Hexagonal (1.27-cm Pitch) Lattices of U(4.4 wt.% 235U)O2 Fuel Rods in Light Water, Perturbed by Boron, Hafnium, or Dysprosium Absorber Rods, or By Water Gap With/Without Empty Aluminium Tubes. NEA/NSC/DOC/(2006)1, PFacility-VVER-EXP-001 CRIT-RRATE, Revision 0, March 31, 2006 Search in Google Scholar
© 2019, Carl Hanser Verlag, München