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1. NAME AND TITLE**

RETRAC: Code System for the Analysis of Materials Test Reactor (MTR) Cores.

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**2. CONTRIBUTOR**

Laboratoire des Analyses de Surete, Algiers, Algeria.

**3. CODING LANGUAGE AND COMPUTER**

FORTRAN 77; VAX (C00635D0VAX00).

**4. NATURE OF PROBLEM SOLVED**

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

**5. 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 Range-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.

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 a net superiority over the one used by the PARET code, particularly in steady-state calculations.

**6. RESTRICTIONS OR LIMITATIONS**

The RETRAC code uses steady-state thermal-hydraulic correlations. Their use is not always justified, but this seems to be quite useful in quasi-steady-state cases such as loss-of-flow transients.

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7. TYPICAL RUNNING TIME**

The running time depends essentially on the time step selected and the accuracy desired by the code user. RSIC tested the sample problems in about 50 minutes.

**8. COMPUTER HARDWARE REQUIREMENTS**

The code was developed on a VAX-4000 working station. Minimum space required is 650 kbytes.

**9. COMPUTER SOFTWARE REQUIREMENTS**

RETRAC is written in FORTRAN 77 and runs under Virtual Memory System.

**10. REFERENCES**

**a: Included in documentation:**

B. Baggoura, T. Hamidouche, and A. Bousbia-Salah, "RETRAC, A Program for the Analysis of M.T.R. Research Reactor Cores," Internal Technical Report (September 1993).

**b: Background information:**

E. R. Choen, "Some Topics in Reactor Kinetics," Proc. 2nd Int. Conf. Peaceful Uses of Atomic Energy, Geneva, Switzerland, September 1-13, 1958, Vol. 11, United National Publications.

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 tot he Use of Low Enriched Uranium Fuel Guidebook," IAEA-TECDOC-233, International Atomic Energy Agency (1980).

W. L. Woodruff, "A Kinetics and Thermal-Hydraulics Capability for the Analysis of Research Reactors," Nucl. Technol., 64, 196 (1984).

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11. CONTENTS OF CODE PACKAGE**

The referenced document in 10.a. and 1 DS/HD (1.44 MB) diskette are included. The diskette, written in DOS, contains source file, two sample problems, and their related output files.

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12. DATE OF ABSTRACT**

June 1995.

** KEYWORDS:** REACTOR ACCIDENT; KINETICS