Skip to main content

Prediction of Fluid Behaviour during Reactor Transient Analysis using Coupled 1D and 3D Models

  • Conference paper
Nuclear Simulation
  • 142 Accesses

Abstract

The detailed understanding of the behaviour of a pressurised water reactor during transient operation is crucial to the design of the reactor.

Hitherto mathematical simulation effort has largely concentrated on one-dimensional loop analysis of the reactor system: this gives a reasonable understanding of the consequences of a transient, such as the blowdown. but does not yield adequate information on features such as pressure effects in the reactor vessel. For greater understanding of such effects, a three-dimensional reactor model is required.

Two- and three-dimensional transient simulations of reactor vessels and of individual components of the circuit have, in the past, been performed in order to provide pressure and temperature fields for input to structural analysis programs. Such decoupled calculations have however tended to underestimate the real situation, since boundary conditions have been specified from stand-alone loop calculations; furthermore, there has been no simple way of incorporating fluid-structure interaction effects.

This paper describes the application of the PHOENICS program to predict the fluid behaviour in a PWR during a hypothetical blowdown. The analysis features two technical novelties. The first is the use of a three-dimensional model for the reactor vessel, directly linked with one-dimensional loop models for the primary water circuits; the second is the dynamic coupling of a fluids code (PHOENICS) with a stress code (ABAQUS) to provide a first step towards the modelling of fluid-structure interaction effects.

This is a preview of subscription content, log in via an institution to check access.

Access this chapter

Chapter
USD 29.95
Price excludes VAT (USA)
  • Available as PDF
  • Read on any device
  • Instant download
  • Own it forever
eBook
USD 39.99
Price excludes VAT (USA)
  • Available as PDF
  • Read on any device
  • Instant download
  • Own it forever
Softcover Book
USD 54.99
Price excludes VAT (USA)
  • Compact, lightweight edition
  • Dispatched in 3 to 5 business days
  • Free shipping worldwide - see info

Tax calculation will be finalised at checkout

Purchases are for personal use only

Institutional subscriptions

Preview

Unable to display preview. Download preview PDF.

Unable to display preview. Download preview PDF.

References

  1. REEDER D. L. ’LOFT System and test Description (5.5 ft Nuclear Core 1 LOCEs)’ NUREG/CR-0247. TREE-1208. July 1978.

    Google Scholar 

  2. PATTON M. L. ’Semiscale MOD-3 Test Program and System Descriptions’. TREE-NUREG-1212. July 19778.

    Google Scholar 

  3. RELAP4/MOD6. ’A Computer Program for Transient Thermal-Hydraulic Analysis of Nuclear Reactors and Related Systems’. EG. and G.. Idaho. Inc., January 1978.

    Google Scholar 

  4. ’TRAC-P1: An Advanced Best-Estimate Computer Program for PWR LOCA Analysis. Vol I: Methods. Models. User Information, and Programming Details. Los Alamos Scientific Laboratory Report LA-7279-MS. Vol I (NUREG/CR0063). June 1978.

    Google Scholar 

  5. MARKATOS N. C.. RAWNSLEY S. M.. and TATCHELL D. G. ’Analysis of a Small-Break Loss-of-Coolant Accident in a Pressurized Water Reactor’. I Mech E Conference Publications 1983-4. Heat and Fluid Flow in Nuclear and Process Plant Safety. Paper C103/83. pp. 121-134 (1983).

    Google Scholar 

  6. MARKATOS N. C.. RAWNSLEY S. M.. and SPALDING D. B002E ’Heat Transfer During a Small-Break Loss-of-Coolant Accident in a Pressurized Water Reactor-A Parametric Study for a 4 in. Lower-Plenum Break.’ Int. J. Heat Mass Transfer Vol. 27. No. 8. pp. 1379–1394. (1984).

    Article  Google Scholar 

  7. KIRKCALDY D.. PHELPS P. J.. and VAN ESSEN D. ’PHOENICS Code Thermal Hydraulic Analysis of the SNR-300 IHX’ ASME Winter annual meeting 1985. HTD-Vol. 51 pp. 9-16.

    Google Scholar 

  8. MES H.. VAN ESSEN D.. KIRKCALDY D.. and PHELPS P. J. ’PHOENICS Code Thermal Hydraulic Analysis of a Prototype LMFBR Straight Tube Steam Generator’. Presented at the ASME Winter Annual Meeting in Anaheim California. Dec. 1986.

    Google Scholar 

  9. HIBBITT H. D. ’A General Purpose Finite Element Code With Emphasis on Nonlinear Applications’. Nuclear Engineering Design 77 pp. 271-297 (1984). North-Holland. Amsterdam.

    Google Scholar 

  10. IDEL’CHIK I. E. ’Handbook of Hydraulic Resistances.’ AEC Translations. 6630. USAEC 1960.

    Google Scholar 

  11. KAYS W M and LONDON A L ’Compact Heat Exhangers’. Mc Graw Hill. New York 1958 (2nd. ed. 1964).

    Google Scholar 

  12. AGEE L. ’Functional Fits of Steam Table Data’. Private Communication 1976.

    Google Scholar 

Download references

Author information

Authors and Affiliations

Authors

Editor information

Editors and Affiliations

Rights and permissions

Reprints and permissions

Copyright information

© 1987 Springer-Verlag, Berlin, Heidelberg

About this paper

Cite this paper

Kirkcaldy, D., Phelps, P.J., Rhodes, N. (1987). Prediction of Fluid Behaviour during Reactor Transient Analysis using Coupled 1D and 3D Models. In: Heller, M.R. (eds) Nuclear Simulation. Springer, Berlin, Heidelberg. https://doi.org/10.1007/978-3-642-83221-5_8

Download citation

  • DOI: https://doi.org/10.1007/978-3-642-83221-5_8

  • Publisher Name: Springer, Berlin, Heidelberg

  • Print ISBN: 978-3-642-83223-9

  • Online ISBN: 978-3-642-83221-5

  • eBook Packages: Springer Book Archive

Publish with us

Policies and ethics