IAEA1286 RETRAC.
RETRAC, Reactor Core Accident Simulation
NAME OR DESIGNATION OF PROGRAM, COMPUTER, DESCRIPTION OF PROGRAM OR FUNCTION, METHOD OF SOLUTION, RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM, TYPICAL RUNNING TIME, UNUSUAL FEATURES OF THE PROGRAM, RELATED AND AUXILIARY PROGRAMS, STATUS, REFERENCES, MACHINE REQUIREMENTS, LANGUAGE, OPERATING SYSTEM UNDER WHICH PROGRAM IS EXECUTED, OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS, NAME AND ESTABLISHMENT OF AUTHORS, MATERIAL, CATEGORIES
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| Program name | Package id | Status | Status date |
|---|---|---|---|
| RETRAC | IAEA1286/01 | Arrived | 19-MAR-2001 |
Machines used:
| Package ID | Orig. computer | Test computer |
|---|---|---|
| IAEA1286/01 | VAX under VMS |
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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).
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).
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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.
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.
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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)
- 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, included references:
- B. Baggoura et al.:RETRAC, A Computer Program for the Analysis of MTR Research
Reactor Cores
04/93/LSN/DTDSESN/CRS
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IAEA1286/01
test-case data mag tapeCOMRET.COM Command for execution DATTPsource program mag tapeRETRAC.FOR Source code SRCTP
test-case data mag tapeRETRAC1.DAT Input for sample problem 1 DATTP
test-case output mag tapeRETRAC1.OUT Output for sample problem 1 OUTTP
test-case data mag tapeRETRAC2.DAT Input for sample problem 2 DATTP
test-case output mag tapeRETRAC2.OUT Output for sample problem 2 OUTTP
report 04/93/LSN/DTDSESN/CRS REPPT
Keywords: accidents, advanced test reactor, hydraulics, irradiation, materials testing, reactor kinetics, thermodynamics, transients.