The reactor dynamics code DYN3D and its trigonal-geometry nodal diffusion model

References

1 Selected safety aspects of WWER-440 model 213 nuclear power plants. Final report of the IAEA project on evaluation of safety aspects for WWER-440 model 213 nuclear power plants. Int. Atomic Energy Agency, Vienna, 1996, ISBN 92-0-101196-2Suche in Google Scholar

2 Krepel, J; Rohde, U.; Grundmann, U; Weiss, F.-P.: DYN3D-MSR spatial dynamics code for molten salt reactors. Annals of Nuclear Energy34 (2007) 44910.1016/j.anucene.2006.12.011Suche in Google Scholar

3 Rohde, U.; Baier, S.; Duerigen, S.; Fridman, E.; Kliem, S.; Merk, B.: Development and verification of the coupled 3D neutron kinetics/thermal-hydraulics code DYN3D-HTR for the simulation of transients in block-type HTGR. Nuclear Engineering and Design251 (2012) 41210.1016/j.nucengdes.2011.09.051Suche in Google Scholar

4 Fridman, E.; Shwageraus, E.: Modeling of SFR cores with Serpent-DYN3D codes sequence. Annals of Nuclear Energy53 (2013) 35410.1016/j.anucene.2012.08.006Suche in Google Scholar

5 Rachamin, R.; Wemple, C.; Fridman, E.: Neutronic analysis of SFR core with HELIOS-2, SERPENT, and DYN3D codes. Annals of Nuclear Energy55 (2013) 19410.1016/j.anucene.2012.11.030Suche in Google Scholar

6 Duerigen, S: Neutron Transport in Hexagonal Reactor Cores Modeled by Trigonal-Geometry Diffusion and Simplified P3 Nodal Methods. PhD thesis, Karlsruhe Institute of Technology, 2013. http://digbib.ubka.uni-karlsruhe.de/volltexte/1000034870.Suche in Google Scholar

7 Grundmann, U.; Rohde, U.; Mittag, S.; Kliem, S.: DYN3D version 3.2 – code for calculation of transients light water reactors (LWR) with hexagonal or quadratic fuel elements – description of models and methods. Tech. Report FZR-434, Forschungszentrum Rossendorf (FZR) August 2005Suche in Google Scholar

8 Rohde, U.; Mittag, S.; Grundmann, U.; Petkov, P.; Hadek, J.: Application of a stepwise verification and validation procedure to the 3D neutron kinetics code DYN3D within the European NURESIM project. Proc. 17th Int. Conf. on Nuclear Engineering (ICONE17) Brussels, Belgium, July 12–16 200910.1115/ICONE17-75446Suche in Google Scholar

9 Beckert, C.; Grundmann, U.: Development and verification of a nodal approach for solving the multigroup SP₃ equations. Annals of Nuclear Energy35 (2008) 7510.1016/j.anucene.2007.05.014Suche in Google Scholar

10 Duerigen S.; Fridman, E.: The simplified P₃ approach on a trigonal geometry of the nodal reactor code DYN3D. Kerntechnik77 (2012) 226Suche in Google Scholar

11 Grundmann, U.: HEXNEM – a nodal method for the solution of the neutron diffusion equation in hexagonal geometry. Proc. Int. Conf. on Mathematics and Computation. Reactor Physics and Environmental Analysis in Nuclear Applications (M&C '99) Madrid, Spain, September 27–30 1999Suche in Google Scholar

12 Grundmann, U.; Hollstein, F.: A two-dimensional intranodal flux expansion method for hexagonal geometry. Nuclear Science and Engineering133 (1999) 201Suche in Google Scholar

13 Manera, A.; Rohde, U.; Prasser, H.-M.; van der Hagen, T. H. J. J.: Modelling of flashing-induced instabilities in the start-up phase of natural-circulation BWRs using the code FLOCAL. Nuclear Engineering and Design235 (2005) 1517Suche in Google Scholar

14 Rohde, U.; Lucas, D.: Solution of the 4th AER dynamic benchmark by use of the code DYN3D with a particle-in-cell method for the description of boron transport. 7th AER Symposium on VVER Reactor Physics and Reactor Safety, Hoernitz near Zittau, Germany, September 23–26 1997, p 387Suche in Google Scholar

15 Rohde, U.: The modeling of fuel rod behaviour under RIA conditions in the code DYN3D. Annals of Nuclear Energy28 (2001) 134310.1016/S0306-4549(00)00128-6Suche in Google Scholar

16 Mittag, S.; Kliem, S.; Weiss, F.-P.; Kyrki-Rajamaki, R.; Hamalainen, H.; Langenbuch, S.; Danilin, S.; Hadek, J.; Hegyi, G.; Kuchin, A.; Panayotov, D.: Validation of coupled neutron kinetic/thermal-hydraulic codes Part 1: Analysis of a VVER-1000 transient (Balakovo-4). Annals of Nuclear Energy28 (2001) 85710.1016/S0306-4549(00)00095-5Suche in Google Scholar

17 Hadek, J.; Grundmann, U.: Neutron flux reconstruction in a hexagonal cassette – theory and implementation into the code DYN3D/H1.1. Nucleon3 (1997) 8Suche in Google Scholar

18 Bilodid, I.; Mittag, S.: Use of the local Pu-239 concentration as an indicator of burnup spectral history in DYN3D. Annals of Nuclear Energy37 (2010) 120810.1016/j.anucene.2010.04.019Suche in Google Scholar

19 Fridman, E.; Duerigen, S.; Bilodid, Y.; Kotlyar, D.; Shwageraus, E.: Axial discontinuity factors for the nodal diffusion analysis of high conversion BWR cores. Annals of Nuclear Energy62 (2013) 129Suche in Google Scholar

20 Grundmann, U.; Mittag, S.: Super-homogenisation factors in pinwise calculations by the reactor dynamics code DYN3D. Annals of Nuclear Energy38 (2011) 211110.1016/j.anucene.2011.06.030Suche in Google Scholar

21 Austregesilo, H.; Bals, C.; Hora, A.; Lerchl, G.; Romstedt, P.: ATHLET, Mod 2.1 Cycle A, Models and Methods. GRS – P – 1/Vol. 4, Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbH, Germany, 2006Suche in Google Scholar

22 NUREG/CR-5535/Rev 1, RELAP5/MOD3.3 code manual. Volume I: Code Structure, System Models, and Solution Methods. December 2001, Idaho Falls, Idaho, USASuche in Google Scholar

23 Grundmann, U.; Lucas, D.; Rohde, U.: Coupling of the thermohydraulic code ATHLET with the neutron kinetic core model DYN3D. Proc. Int. Conf. on Mathematics and Computations, Reactor Physics and Environmental Analyses, Portland, Oregon, USA, April 30-May 4 1995Suche in Google Scholar

24 Kozmenkov, Y.; Orekhov, Y.; Grundmann, U.; Kliem, S.; Rohde, U.; Seidel, A.: Development and benchmarking of the DYN3D/RELAP code system. Annual Meeting of the German Nuclear Society, Dresden, Germany, May 15–17 2001Suche in Google Scholar

25 Kliem, S.; Mittag, S.; Rohde, U.; Weiß, F.-P.: ATWS analysis for PWR using the coupled code system DYN3D/ATHLET. Annals of Nuclear Energy36 (2009) 123010.1016/j.anucene.2009.04.001Suche in Google Scholar

26 Kliem, S.; Rohde, U.; Weiß, F.-P.: Core response of a PWR to a slug of under-borated water. Nuclear Engineering and Design230 (2004) 121Suche in Google Scholar

27 Sánchez, V.; Imke, U.; Gomez, R.: SUBCHANFLOW: A thermal-hydraulic sub-channel program to analyse fuel rod bundles and reactor cores. 17th Pacific Basin Nuclear Conference, Cancún, Q. R., México, October 24–30 2010Suche in Google Scholar

28 Gomez-Torres, A. M.; Sanchez-Espinoza, V. H.; Ivanov, K.; Macian-Juan, R.: DYNSUB: A high fidelity coupled code system for the evaluation of local safety parameters – Part I: Development, implementation and verification. Annals of Nuclear Energy48 (2012) 108Suche in Google Scholar

29 Kliem, S.; Gommlich, A.; Grahn, A.; Rohde, U.; Schütze, J.; Frank, T.; Gomez, A.; Sanchez, V.: Development of multi-physics code systems based on the reactor dynamics code DYN3D. Kerntechnik76 (2011) 160Suche in Google Scholar

30 Lassmann, K.: TRANSURANUS: a fuel rod analysis code ready for use. Journal of Nuclear Materials188 (1992) 295Suche in Google Scholar

31 Finnemann, H.; Bennewitz, F.; Wagner, M. R.: Interface current techniques for multidimensional reactor calculations. Atomkernenergie30 (1977) 123Suche in Google Scholar

32 Koebke, K.; Hetzelt, L; Wagner, M. R.; Winter, H.-J.: Principles and application of advanced nodal reactor analysis methods. Proc. Topical Meeting on Reactor Physics and Shielding, Chicago, Illinois, USA, September 17–19 1984Suche in Google Scholar

33 Cho, N. Z.; Kim, Y. H.; Park, K. W.: Extension of analytic function expansion nodal method to multigroup problems in hexagonal-z geometry. Nuclear Science and Engineering, 126 (1997) 35Suche in Google Scholar

34 Noh, J. M.; Cho, N. Z.: A new approach of analytic basis function expansion to neutron diffusion nodal calculation. Nuclear Science and Engineering116 (1994) 165Suche in Google Scholar

35 Downar, T. J.; Lee, D.; Xu, Y.; Kozlowski, T.; Staudenmier, J.: PARCS v2.6 U. S. NRC Core Neutronics Simulator. Theory Manual (Draft) 2004Suche in Google Scholar

36 Kochunas, B.; Xu, Y.; Downar, T. J.: Development of the triangular nodal method TRIPEN for prismatic HTR analysis. Proc. PHYSOR 2010 – Advances in Reactor Physics to Power the Nuclear Renaissance, Pittsburgh, Pennsylvania, USA, May 9–14 2010Suche in Google Scholar

37 Chao, Y. A.; Shatilla, Y. A.: Conformal mapping and hexagonal nodal methods – II: Implementation in the ANC-H code. Nuclear Science and Engineering121 (1995) 210Suche in Google Scholar

38 Derstine, K. L.: DIF3D: A code to solve one-, two-, and three-dimensional finite-difference diffusion theory problems. Tech. Report ANL-82-64, Argonne National Laboratory, 198410.2172/7157044Suche in Google Scholar

39 Loetsch, T.; Khalimonchuk, V.; Kuchin, A.: Proposal of a benchmark for core burnup calculations for a VVER-1000 reactor core. Proc. 19th Symposium of AER on VVER Reactor Physics and Reactor Safety, Varna, Bulgaria, September 21–25 2009Suche in Google Scholar

40 Casal, J. J.; Stamm'ler, R. J. J.; Villarino, E. A.; Ferri, A. A.: HELIOS: geometric capabilities of a new fuel assembly program. Proc. Int. Topical Meeting on Advances in Mathematics, Computations and Reactor Physics, Pittsburgh, Pennsylvania, USA, April 28 – May 2 1991Suche in Google Scholar