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Licensed Unlicensed Requires Authentication Published by De Gruyter May 6, 2013

Simulation of the neutronic-thermal hydraulic stability behaviour of boiling water reactors

Simulation des nuklear-thermohydraulischen Stabilitätsverhaltens von Siedewasserreaktionen
  • A. H. Gorzel , M. W. Henschel , E. M. Mrosk and W. E. Scheerer
From the journal Kerntechnik


Neutronic-thermal hydraulic oscillations are a well known phenomenon in the operation of boiling water reactors. Despite the progresses that have been made concerning their simulation, the computational determination of the stability threshold in the power-flow map still suffers from inaccuracies. In this work, the decay ratio of neutronic-thermal hydraulic oscillations and hence the stability threshold was calculated with the SIMULATE-3K code developed by Studsvik Scandpower. It is known from experience that there is a higher probability for a less stable reactor operation in the case of one or more pumps tripped while the remaining pumps run at minimum speed. For the analysis of this state, the characteristics of the pump system were modelled carefully. It is shown that the stability behaviour is determined by the axial and radial power profile due to control rod manoeuvring and fuel burn-up and by the number of pumps running. Furthermore the calculations reveal the impacts of the insertion of shallow rods and of the control-rod group insertion on the decay ratio. A comparison of the calculations with experimental results of stability measurements shows that by a thorough analysis and proper modelling of the plant characteristics the decay ratios can be calculated with the SIMULATE-3K code.


Nuklear-thermohydraulische Schwingungen sind ein bekanntes Phänomen beim Betrieb von Siedewasserreaktionen. Trotz der Fortschritte, die es hinsichtlich ihrer Simulation gab, ist die theoretische Bestimmung der Stabilitätsgrenze im Betriebskennfeld mit Unsicherheiten verknüpft. In dieser Arbeit wurde das Abklingverhältnis nuklear-thermohydraulischer Schwingungen und damit auch die Stabilitätsgrenze mithilfe des Programms SIMULATE-3K der Firma Studsvik Scanpower ermittelt. Aus früheren Messungen und analytischen Untersuchungen ist bekannt, dass sich zwischen den Betriebspunkten Naturumlauf und Mindestdrehzahl aller Zwangsumwälzpumpen ein minimaler Abstand zwischen der Stabilitätsgrenze und der Umwälzregelkennlinie einstellt. Für die theoretische Beschreibung dieses Zustands wurde das Pumpenkennfeld als mittleres Pumpenkennfeld – entsprechend den Anforderungen des Programms – sorgfälltig modelliert. Es werden die Auswirkungen sowohl von Änderungen des axialen unnd radialen Leistungsprofils aufgrund von Steuerstabbewegungen und fortschreitendem Abbrand als auch von der Anzahl der laufenden Pumpen auf das Stabilitätsverhalten dargestellt. Darüber hinaus demonstrieren die Berechnungen die Einflüsse von „Shallow Rods“ und des Pulkeinfahrens von Steuerstäben auf das Abklingverhältnis. Ein Vergleich der Berechnungen mit Stabilitätsmessungen zeigt – bei hinreichend genauer Modellierung der Anlage – die Anwendbarkeit des Programms SIMULATE-3K für Stabilitätsberechnungen.



1 Wulff, W.; Cheng, H. S. and Mallen, A. N.: Causes of instability of La Salle and Consequences from postulated SCRAM failure. International Workshop on BWR stability, CSNI Report 178, New York, 1990.Search in Google Scholar

2 Wahba, A.-B. and Langenbuch, S.: Stability of boiling water reactors - phenomena, occurences and monitoring. Kerntechnik62 (1997) 160.Search in Google Scholar

3 Boure, J. A.; Bergles, A. E. and Tong, L. S.: Review of two Phase Flow Instabilitiy. Nuclear Engineering and Design25 (1973) 165.10.1016/0029-5493(73)90043-5Search in Google Scholar

4 March-Leuba, J.: Dynamic Behaviour of Boiling Water Reactors, PhD Thesis, The University of Tennessee, Knoxville, 1984.Search in Google Scholar

5 Pfefferlen, H.; Rausch, R. and Watford, G.: BWR Core Thermal-Hydraulic Stability: Experience and Safety Significance. International Workshop on Boiling Water Reactor Stability, Proceedings, pp. 45–58, Holtsville, New York, CSNI Report 178, OECD Nuclear Energy Agency, Paris, 1990.Search in Google Scholar

6 March-Leuba, J.; Cacuci, D. G. and Perez, R. B.: Non-Linear Dynamics and Stability of Boiling Water Reactors: Part 1 - Qualitative Analysis and Part 2 - Quantitative Analysis. Nuclear Science and Engineering93 (1986) pp. 111–123 and pp. 124136.Search in Google Scholar

7 Andersson, S. and Stepniewski, M.: Assessment of RAMONA-3B Calculations of Core-Wide and Regional Power/Flow Oscillations - Comparison with Oskarsham 3 Natural Circulation Test Data. International Workshop on Boiling Water Reactor Stability, Proceedings, pp. 397411, Holtsville, New York, CSNI Report 178, OECD Nuclear Energy Agency, Paris, 1990.Search in Google Scholar

8 March-Leuba, J. and Blackmann, E. D.: A Mechanism for Out-of Phase Power Instabilities in Boiling Water Reactors. Nucl. Sci. Eng.107 (1991) 173.Search in Google Scholar

9 Ceceñas-Falcón, M. and Edwards, R. M.: Out-of-Phase Boiling Water Reactor Stability Monitoring. Nuclear Technology143 (2003) 125.Search in Google Scholar

10 Ginestar, D.; Miró, R.; Verdú, G. and Hennig, D.: A Transient Modal Analysis of a BWR Instability Event. J. of Nuclear Science and Technology39 (2002) 554.10.1080/18811248.2002.9715234Search in Google Scholar

11 Takeuchi, Y.; Takigawa, Y. and Uematsu, H.: A Study of Boiling Water Reactor Regional Stability from the Viewpoint of Higher Harmonics. Nuclear Technology106 (1994) 300.Search in Google Scholar

12 Park, G. C.; Podowski, M.; Becker, M. and Lahey, R. T.: The Development of NUFREQ-N. An Analytical Model for the Stability Analysis of Nuclear Coupled Density-Wave Oscillations in Boiling Water Nuclear Reactors. NUREG/CR-3375, 1983.Search in Google Scholar

13 Kreuter, D. and Wehle, F.: Siemens-KWU Experience on Linear and Nonlinear Analyses of Out-of-Phase BWR Instabilities. TOPFUELÜ95, Vol. II, pp. 135140, 1995.Search in Google Scholar

14 Wulff, W.; Cheng, H. S. and Diamond, D. J.: A Description and Assessment of RAMONA-3B Mod. 0 Cycle 4: A Computer Code with Three-Dimensional Neutron Kinetic for BWR System Transients. NUREG/CR-3664, BNL-NUREG-51764, Brookhaven National Laboritory, 1984.Search in Google Scholar

15 Moberg, L.: Assessment and Application of the RAMONA 3-Dimensional Transient Code to BWR Stability. Proc. Int. Workshop on Boiling Water Stability, Holtsville, New York, CSNI Report 178 (1991) 385.Search in Google Scholar

16 Borkowski, J.; Rhodes III, J.; Esser, P.; Smith, K.: A Three-Dimensional Transient Analysis Capability for SIMULATE-3. Trans. Am., Nuc., Soc.71 (1994) 456.Search in Google Scholar

17 Borkowski, J.; Smith, K.; Hagrman, D.; Kropaczek, D.; Rhodes III, J.; Esser, P.: SIMULATE-3K Simulations of the Ringhals-1 BWR Stability Measurements. Proceedings PHYSOR 1996, Mito, Ibaraki, Japan, 1996.Search in Google Scholar

18 Borkowski, J.; Smith, K.; Hagrman, D.; Rhodes III, J.: Best-estimate Three-Dimensional Transient Analysis Using Design-basis Methodology. International Meeting on Best-Estimate Methods in Nuclear Installation Safety Analysis (BE-2000), Washington, DC, 2000.Search in Google Scholar

19 Grandi, G. M. and Smith, K. S.: BWR Stability Analysis with SIMULATE-3K. Proceedings PHYSOR 2002, Seoul, Korea, October7–10, 2002.Search in Google Scholar

20 Brandes, L. P. and Waschull, W.: Practical Experience with Instability Measurements in German BWR's and Related Consequences. Proc. Int. Workshop Boiling Water Reactors Stability, Holtsville, New York, 1990, p. 104.Search in Google Scholar

21 Dayal, D. and Preusche, G.: Stability Tests at KWU Nuclear Power Plants. ANS Topical Meeting on Anticipated and Abnormal Transients in Nuclear Power Plants, Atlanta, GA, 1987.Search in Google Scholar

22 Kreuter, D. and Wehle, F.: Ergenisse einer Parameterstudie zur SWR-Stabilität, Annual Meeting of the German Nuclear Society, Bonn, 1991.Search in Google Scholar

23 Rhodes, Joel D.: CASMO4 - A Fuel Assembly Burnup Program - User's manual Studsvik Scandpower, 2003.Search in Google Scholar

24 Covington, L. J.: SIMULATE-3 - Advanced Three-Dimensinal Two-Group Reactor Analysis Code - User's Manual Studsvik Scandpower, 2001.Search in Google Scholar

25 Kropaczek, D. J.; Smith, K. S.; Borkowski, J. A.: A Fully Implicit Five Equation Channel Hydraulics Model for SIMULATE-3K. Proceedings Joint Int. Conf. on Mathematical Methods and Supercomputing for Nuclear Applications, Saratoga Springs, 1, 140119.Search in Google Scholar

26 Cronin, J. T.; Smith, K. S. and Umbarger, J. A.: INTERPIN-CS User's Manual. Studsvik Scandpower, 1995.Search in Google Scholar

27 Grandi, G. M.: RAMONA5 User Manual. Studsvik Scandpower, 2001.Search in Google Scholar

28 Grandi, G. M.; Borkowsky, J. A. and Smith, K. S.: SIMULATE-3K Models & Methodology. Rev. 1, Upgrade to Peripheral Systems, 2002.Search in Google Scholar

Received: 2005-2-8
Published Online: 2013-05-06
Published in Print: 2005-08-01

© 2005, Carl Hanser Verlag, München

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