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Title: COBRA-IV: the model and the method

Abstract

The objective of this report is to present the mathematical basis of the COBRA-IV computer program (Wheeler et al., 1976) being developed by Battelle, Pacific Northwest Laboratory. The COBRA-IV code is an extended version of the COBRA-IIIC subchannel analysis code that computes the flow and enthalpy distributions in nuclear fuel rod bundles and cores for both steady state and transient conditions (Rowe, 1973).

Authors:
; ; ; ; ;
Publication Date:
Research Org.:
Pacific Northwest Lab., Richland, WA (USA)
OSTI Identifier:
5358588
Report Number(s):
BNWL-2214
ON: TI86015346
DOE Contract Number:
AC06-76RL01830
Resource Type:
Technical Report
Country of Publication:
United States
Language:
English
Subject:
22 GENERAL STUDIES OF NUCLEAR REACTORS; 42 ENGINEERING; HEAT TRANSFER; C CODES; HYDRAULICS; NUCLEAR POWER PLANTS; ROD BUNDLES; FUEL RODS; MATHEMATICAL MODELS; STEADY-STATE CONDITIONS; TRANSIENTS; COMPUTER CODES; ENERGY TRANSFER; FLUID MECHANICS; FUEL ELEMENTS; MECHANICS; NUCLEAR FACILITIES; POWER PLANTS; REACTOR COMPONENTS; THERMAL POWER PLANTS; 220200* - Nuclear Reactor Technology- Components & Accessories; 420400 - Engineering- Heat Transfer & Fluid Flow

Citation Formats

Stewart, C.W., Wheeler, C.L., Cena, R.J., McMonagle, C.A., Cuta, J.M., and Trent, D.S. COBRA-IV: the model and the method. United States: N. p., 1977. Web. doi:10.2172/5358588.
Stewart, C.W., Wheeler, C.L., Cena, R.J., McMonagle, C.A., Cuta, J.M., & Trent, D.S. COBRA-IV: the model and the method. United States. doi:10.2172/5358588.
Stewart, C.W., Wheeler, C.L., Cena, R.J., McMonagle, C.A., Cuta, J.M., and Trent, D.S. Fri . "COBRA-IV: the model and the method". United States. doi:10.2172/5358588. https://www.osti.gov/servlets/purl/5358588.
@article{osti_5358588,
title = {COBRA-IV: the model and the method},
author = {Stewart, C.W. and Wheeler, C.L. and Cena, R.J. and McMonagle, C.A. and Cuta, J.M. and Trent, D.S.},
abstractNote = {The objective of this report is to present the mathematical basis of the COBRA-IV computer program (Wheeler et al., 1976) being developed by Battelle, Pacific Northwest Laboratory. The COBRA-IV code is an extended version of the COBRA-IIIC subchannel analysis code that computes the flow and enthalpy distributions in nuclear fuel rod bundles and cores for both steady state and transient conditions (Rowe, 1973).},
doi = {10.2172/5358588},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Fri Jul 01 00:00:00 EDT 1977},
month = {Fri Jul 01 00:00:00 EDT 1977}
}

Technical Report:

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  • The objective of the Core Thermal Model Development program currently in progress at the Battelle Pacific Northwest Laboratory is to develop numerical simulation methods for evaluating thermal hydraulic performance of water-cooled nuclear reactor cores and other components during postulated accident conditions. The primary thrust of this program to date has been continued development of the COBRA computer code as a tool for multidimensional transient simulation applicable to either PWR or BWR cores. COBRA-IV incorporates numerous numerical features which significantly extend computational capabilities beyond that of the widely used COBRA-III-C. One unique feature of COBRA-IV is that it maintains the subchannelmore » connection logic used by COBRA-III-C but applies this logic in a manner similar to the finite-difference methods common in numerical hydrodynamics.« less
  • COBRA-IV PC is a modified version of COBRA-IV-I, adapted for use with most IBM PC and PC-compatible desktop computers. Like COBRA-IV-I, COBRA-IV PC uses the subchannel analysis approach to determine the enthalpy and flow distribution in rod bundles for both steady-state and transient conditions. The steady-state and transient solution schemes used in COBRA-IIIC are still available in COBRA-IV PC as the implicit solution scheme option. An explicit solution scheme is also available, allowing the calculation of severe transients involving flow reversals, recirculations, expulsions, and reentry flows, with a pressure or flow boundary condition specified. In addition, several modifications have beenmore » incorporated into COBRA-IV PC to allow the code to run on the PC. These include a reduction in the array dimensions, the removal of the dump and restart options, and the inclusion of several code modifications by Oregon State University, most notably, a critical heat flux correlation for boiling water reactor fuel and a new solution scheme for cross-flow distribution calculations. 7 refs., 8 figs., 1 tab.« less
  • This report summarizes the result of studies concerning the range of applicability of two subchannel codes for a variety of thermal-hydraulic analyses. The subchannel codes used include COBRA IIIC/MIT and the newly developed code, COBRA IV-I, which is considered the benchmark code for the purpose of this report. Hence, through the comparisons of the two codes, the applicability of COBRA IIIC/MIT is assessed with respect to COBRA IV-I. A variety of LWR thermal-hydraulic analyses are examined. Results of both codes for steady-state and transient analyses are compared. The types of analysis include BWR bundle-wide analysis, a simulated rod ejection andmore » loss of flow transients for a PWR. The system parameters were changed drastically to reach extreme coolant conditions, thereby establishing upper limits. In addition to these cases, both codes are compared to experimental data including measured coolant exit temperatures in a core, interbundle mixing for inlet flow upset cases and two-subchannel flow blockage measurements. The comparisons showed that, overall, COBRA IIIC/MIT predicts most thermal-hydraulic parameters quite satisfactorily. However, the clad temperature predictions differ from those calculated by COBRA IV-I and appear to be in error. These incorrect predictions are caused by the discontinuity in the heat transfer coefficient at the start of boiling. Hence, if the heat transfer package is corrected, then COBRA IIIC/MIT should be just as applicable as the implicit option of COBRA IV-I.« less
  • The results of thermal-hydraulic simulations of a Combustion Engineering System 80 U-tube steam generator are presented. Both steady-state and transient conditions are considered to compare the results of the Battelle COBRA-IV homogeneous-equilibrium code and a newly developed two-fluid (UVUT) two-phase flow model called COBRA-TF (EPRI). The details of formulating the code input to simulate the primary and secondary flows are discussed. All results presented here must be considered as only a demonstration and not as actual steam generator behavior.
  • The COBRA-IV-I computer code uses the subchannel analysis approach to determine the enthalpy and flow distribution in rod bundles for both steady-state and transient conditions. The steady-state and transient solution schemes used in COBRA-IIIC are still available in COBRA-IV-I as the implicit solution scheme option. In addition to these techniques, a new explicit solution scheme is now available which allows the calculation of severe transients involving flow reversals, recirculations, expulsion and reentry flows, with a pressure or flow boundary condition specified. Significant storage compaction and reduced running times have been achieved to allow the calculation of problems involving hundreds ofmore » subchannels.« less