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IAEA1286 RETRAC. (Abstract last modified 19-MAR-2001)


RETRAC, Reactor Core Accident Simulation

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Abstract Table of Contents:

NAME | COMPUTER | PROBLEM | SOLUTION | RESTRICTIONS | CPU | FEATURES | AUXILIARIES | STATUS | REFERENCES | REQUIREMENTS | LANGUAGE | OPERATING SYSTEM | OTHER RESTRICTIONS | AUTHOR | MATERIAL | CATEGORIES |

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1. NAME OR DESIGNATION OF PROGRAM - RETRAC : REactor TRansient Analysis Code. 


2. COMPUTER FOR WHICH PROGRAM IS DESIGNED AND OTHER MACHINE VERSION PACKAGES AVAILABLE -

To request or retrieve programs click on the one of the active versions below. A password and special authorization is required.

Explanation of the status codes.

Program-name         Package-ID  Status
RETRAC               IAEA1286/01 Arrived

Machines used:
Package-ID     Orig.Computer                       Test Computer
IAEA1286/01    VAX under VMS                       

3. DESCRIPTION OF PROGRAM OR FUNCTION - The RETRAC code uses a set of coupled neutron point-kinetics equations and thermal-hydraulic conservation laws to simulate nuclear reactor core behaviour under transient or accident conditions. The reactor core is represented by single equivalent unit cells composed of three regions: fuel, clad, and moderator (coolant).


4. METHOD OF SOLUTION - At each time step, core thermal power is calculated by solving a set of six delayed neutron group kinetics equations with adjusted reactivity feedbacks. The numerical resolution is performed by using the Runge-Kutta-Gill method. The externally inserted reactivity is specified in the input data file, whereas Doppler, fuel, clad, and water temperature reactivity feedbacks are calculated by the code itself. Core cooling is treated as a homogeneous one-dimensional fluid flow through a representative unit cell composed of three successive regions: fuel, clad, and coolant. Several flow regime models are considered for both single- and two-phase states of the coolant. The conservation laws are solved by the method of characteristics coupled with an implicit finite difference scheme to ensure stability and convergence of the numerical algorithm.
Validation tests of the RETRAC code were performed by using the International Atomic Energy Agency 10-MW benchmark cores, for protected transients. Further assessment studies are in progress using experimental data.


5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM - The RETRAC code uses steady-state thermal-hydraulic correlations. Their use is not always justified, but it seems to be quite useful in quasi-steady cases such as as loss-of-flow transients.


6. TYPICAL RUNNING TIME - The running time depends essentially on the time step selected and the accuracy desired by the code user.


7. UNUSUAL FEATURES OF THE PROGRAM - The method of characteristics used to solve the set of thermal-hydraulic conservation equations is a
very stable and highly converging numerical scheme, which has shown net superiority over the one used by the PARET code, particularly in steady-state calculations.


8. RELATED AND AUXILIARY PROGRAMS - No additional programs are required. 


9. STATUS
IAEA1286/01: 19-MAR-2001 Masterfiled Arrived


10. REFERENCES  - 
- E. Cohen: 
  "Some Topics in Reactor Kinetics", Proc. 2nd Int. Conf. Peaceful 
  Uses of Atomic Energy. Vol. 11, Geneva, Sept. 1 to 13, UN 
  Publication (1958) 
- S. Nakamura: 
  Computational Methods in Engineering and Science, 
  Wiley-Interscience, New York (1977) 
- International Atomic Energy Agency: 
Safety and Licensing Guidebook of Research Reactor Core Conversion from the Use of High Enriched Uranium to the Use of Low Enriched
  Uranium, Vol. 3, Appendixes G and H, International Atomic Energy 
  Agency, Vienna (1990) 
- Research Reactor Core Conversion from the Use of High Enriched 
  Uranium to the Use of Low Enriched Uranium Fuel Guidebook, 
IAEA-TECDOC-233, International Atomic Energy Agency, Vienna (1980) - W. Woodruff:
  A Kinetics and Thermal-Hydraulic Capability for the Analysis of 
  Research Reactors, Nucl. Technol. 64, 196 (1984) 
IAEA1286/01:
- B. Baggoura et al.:
  RETRAC, A Computer Program for the Analysis of MTR Research
  Reactor Cores
  04/93/LSN/DTDSESN/CRS


11. MACHINE REQUIREMENTS - Minimum space required: 650 Kbytes. 


12. PROGRAMMING LANGUAGE(S) USED -
IAEA1286/01: FORTRAN-77 


13. OPERATING SYSTEM UNDER WHICH PROGRAM IS EXECUTED - VMS. 


14. OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS - The 
code requires at least two logical units for input and output files. 


15. NAME AND ESTABLISHMENT OF AUTHORS - 
B. Baggoura, T. Hamidouche and A. Bousbia-Salah 
Centre de Radioprotection et de Surete 
Laboratoire des Analyses de Surete 
02, Bd Frantz Fanon 
BP 1017 Alger Gare 
16000 Algiers, 
Algeria 


16. MATERIAL AVAILABLE -
IAEA1286/01:
COMRET.COM Command for execution           
RETRAC.FOR Source code                     
RETRAC1.DAT Input for sample problem 1     
RETRAC1.OUT Output for sample problem 1    
RETRAC2.DAT Input for sample problem 2     
RETRAC2.OUT Output for sample problem 2    
report                   04/93/LSN/DTDSESN/CRS                      REPPT


17. CATEGORIES  - 

- G.  Radiological Safety, Hazard and Accident Analysis


Keywords: ACCIDENTS, ADVANCED TEST REACTOR, HYDRAULICS, IRRADIATION, MATERIALS TESTING, REACTOR KINETICS, THERMODYNAMICS, TRANSIENTS

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