Fission source sampling in coupled Monte Carlo simulations
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B. Olsen
und
J. Dufek
Abstract
We study fission source sampling methods suitable for the iterative way of solving coupled Monte Carlo neutronics problems. Specifically, we address the question as to how the initial Monte Carlo fission source should be optimally sampled at the beginning of each iteration step. We compare numerically two approaches of sampling the initial fission source; the tested techniques are derived from well-known methods for iterating the neutron flux in coupled simulations. The first technique samples the initial fission source using the source from the previous iteration step, while the other technique uses a combination of all previous steps for this purpose. We observe that the previous-step approach performs the best.
Abstract
Wir untersuchen das Sampeln der Neutronenspaltquelle und entsprechende Sampling-Methoden, welche für das iterative Lösen gekoppelter Monte-Carlo-Neutronikprobleme geeignet sind. Insbesondere widmen wir uns der Frage, wie die anfängliche Monte-Carlo-Spaltquelle zu Beginn eines jeden neuen Iterationsschrittes optimal gesampelt werden sollte. Numerisch vergleichen wir zwei Vorgehensweisen, die Anfangsspaltquelle zu sampeln; die getesteten Verfahren sind abgeleitet von zwei allgemein bekannten Methoden zur Iteration des Neutronenflusses in gekoppelten Simulationen. Das erste Verfahren sampelt die initiale Spaltquelle auf der Grundlage des vorigen Iterationsschrittes, wohingegen in dem zweiten Verfahren eine Kombination aus allen vorausgegangenen Iterationsschritten gebildet wird. Wir beobachten, dass der Ansatz, die Spaltquelle aus dem vorherigen Iterationsschritt abzuleiten, am besten funktioniert.
Acknowledgements
The simulations were performed on resources provided by the Swedish National Infrastructure for Computing (SNIC) at PDC Center for High Performance Computing at KTH, Royal Institute of Technology.
References
1 Dufek, J.; Gudowski, W.: Stochastic Approximation for Monte Carlo Calculation of Steady-State Conditions in Thermal Reactors. Nuclear Science and Engineering 52 (2006) 274–283, DOI:10.13182/NSE06-210.13182/NSE06-2Suche in Google Scholar
2 Isotalo, A. E.; Leppänen, J.; Dufek, J.: Preventing xenon oscillations in Monte Carlo burnup calculations by enforcing equilibrium xenon distribution. Annals of Nuclear Energy 60 (2013) 75–85, DOI:10.1016/j.anucene.2013.04.03110.1016/j.anucene.2013.04.031Suche in Google Scholar
3 Dufek, J.; Anglart, H.: Derivation of a stable coupling scheme for Monte Carlo burnup calculations with the thermal–hydraulic feedback. Annals of Nuclear Energy 62 (2013) 260–263, DOI:10.1016/j.anucene.2013.06.02510.1016/j.anucene.2013.06.025Suche in Google Scholar
4 Dufek, J.; Valtavirta, V.: Time step length versus efficiency of Monte Carlo burnup calculations. Annals of Nuclear Energy 72 (2014) 409–412, DOI:10.1016/j.anucene.2014.06.01910.1016/j.anucene.2014.06.019Suche in Google Scholar
5 Moore, R. L.; Schnitzler, B. G.; Wemple, C. A.: MOCUP: MCNP-ORIGEN2 coupled utility program. USA: Idaho National Engineering Laboratory; 1995, Report no. INEL-95/052310.2172/130667Suche in Google Scholar
6 Leppänen, J.: Serpent – a Continuous-energy Monte Carlo Reactor Physics Burnup Calculation Code. Finland: VTT Technical Research Centre of Finland; 2012Suche in Google Scholar
7 Fensin, M. L.; Hendricks, J. S.; Anghaie S.: The enhancements and testing for the MCNPX 2.6.0 depletion capability. Nuclear Technology 170 (1995) 68–7910.13182/NT10-2Suche in Google Scholar
8 Griesheimer, D. P.; Gill, D. F.; Nease, B. R.; Sutton, T. M.; Stedry, M. H.; Dobreff, P. S.; Carpenter, D. C.; Trumbull, T. H.; Caro, E.; Joo, H.; Millman, D. L.: MC21 v.6.0 – A continuous-energy Monte Carlo particle transport code with integrated reactor feedback capabilities. Annals of Nuclear Energy 82 (2015) 29–40, DOI:10.1016/j.anucene.2014.08.02010.1016/j.anucene.2014.08.020Suche in Google Scholar
9 Sjenitzer, L. B.; Hoogenboom, E. J.: A Monte Carlo method for calculation on the dynamic behaviour of nuclear reactors. Progress in Nuclear Science and Technology 2 (2011) 716–721, DOI:10.15669/pnst.2.71610.15669/pnst.2.716Suche in Google Scholar
10 Dufek, J.; Hoogenboom, J. E.: Description of a stable scheme for steady-state coupled Monte Carlo-thermal-hydraulic calculations. Annals of Nuclear Energy 68 (2014) 1–3, DOI:10.1016/j.anucene.2013.12.01710.1016/j.anucene.2013.12.017Suche in Google Scholar
11 DOE Fundamentals Handbook – Nuclear Physics and Reactor Theory, Volume 2 of 2. Washington, D.C.: U.S. Department of Energy; 1993, p. 34 –42Suche in Google Scholar
12 Etherington, H.: Nuclear engineering handbook. New York: McGraw-Hill handbooks; 1958, p. 8-27–8-29Suche in Google Scholar
© 2017 Carl Hanser Verlag GmbH & Co. KG
Artikel in diesem Heft
- Frontmatter
- Best estimate approach for the evaluation of critical heat flux phenomenon in the boiling water reactors
- Coupled neutronic core and subchannel analysis of nanofluids in VVER-1000 type reactor
- PCTRAN enhancement for large break loss of coolant accident concurrent with loss of offsite power in VVER-1000 simulation
- Experimental study of natural circulation flow instability in rectangular channels
- Exerimental method and preliminary studies of the passive containment water film evaporation mass transfer
- Automated generation of burnup chain for reactor analysis applications
- Fission source sampling in coupled Monte Carlo simulations
- Hysteresis phenomenon in nuclear reactor dynamics
- Investigation of neutronic and safety parameters variation in 5 MW research reactor due to U3O8Al fuel conversion to ThO2 + U3O8Al
- Implementation of meso-scale radioactive dispersion model for GPU
- Solution of the multilayer multigroup neutron diffusion equation in cartesian geometry by fictitious borders power method
- Half-space albedo problem with modified FN method for linear and quadratic anisotropic scattering
Artikel in diesem Heft
- Frontmatter
- Best estimate approach for the evaluation of critical heat flux phenomenon in the boiling water reactors
- Coupled neutronic core and subchannel analysis of nanofluids in VVER-1000 type reactor
- PCTRAN enhancement for large break loss of coolant accident concurrent with loss of offsite power in VVER-1000 simulation
- Experimental study of natural circulation flow instability in rectangular channels
- Exerimental method and preliminary studies of the passive containment water film evaporation mass transfer
- Automated generation of burnup chain for reactor analysis applications
- Fission source sampling in coupled Monte Carlo simulations
- Hysteresis phenomenon in nuclear reactor dynamics
- Investigation of neutronic and safety parameters variation in 5 MW research reactor due to U3O8Al fuel conversion to ThO2 + U3O8Al
- Implementation of meso-scale radioactive dispersion model for GPU
- Solution of the multilayer multigroup neutron diffusion equation in cartesian geometry by fictitious borders power method
- Half-space albedo problem with modified FN method for linear and quadratic anisotropic scattering
