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NEA-0807 TRAWA.

TRAWA, LWR Dynamic by Coupled Neutron Diffusion and Thermohydraulics Calculation

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1. NAME OR DESIGNATION OF PROGRAM:  TRAWA.
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2. COMPUTERS
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Program name Package id Status Status date
TRAWA NEA-0807/01 Tested 08-AUG-1983

Machines used:

Package ID Orig. computer Test computer
NEA-0807/01 UNIVAC 1110 UNIVAC 1110
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3. DESCRIPTION OF PROBLEM OR FUNCTION

The purpose of the program is to study reactor dynamics in thermal water-cooled reactors. It treats the core as one or a few axially one-dimensional subregions.  The two group neutron diffusion equations are solved simultaneously  with the heat conduction equations and the two-phase hydraulic equations for one or more channels. Neither thermal nor hydraulic mixing appear between channels. Doppler, coolant density, coolant temperature, and soluble poison density feedbacks due to the thermo- hydraulics of the channels are described by using polynomial expansions for the group constants. The hydraulic circuit outside the reactor core consists of by-pass channels and risers with two- phase flow and of pump lines with incompressible flow.

Various transients can be calculated by applying external disturbances. They can affect e.g. on movements of control rods, core inlet hydraulic conditions, system pressure or coefficients of  neutronic shape function expansion between subregions.
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4. METHOD OF SOLUTION

Nontrivial implicit methods are employed in the discretization of the equations to allow for sparse spatial mesh and flexible choice of time steps. The same spatial and temporal discretization is used for neutronics and thermohydraulics.
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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM

The dimensions of the program variable tables can easily be extended. Now the main dimensions are:
   52 axial mesh points in core
    3 subregions
   10 axial regions with different fuel compositions
    7 radial mesh points in fuel rod
    6 delayed neutron groups
    6 coupled legs in pressure balance calculation

No flow reversals are allowed.
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6. TYPICAL RUNNING TIME

The running time on UNIVAC 1108 is 0.7 ... 2.4 seconds per time-step depending on the number of mesh points and subregions. Time-steps extending from some milliseconds to ten seconds have been used in calculating different transients.
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7. UNUSUAL FEATURES OF THE PROGRAM

New input data can be given during the transient. The program calculates the DNB ratio with different correlations in any subregion.

The thermal properties of the fuel, the cladding and the gas gap between them are temperature dependent and the properties of the coolant are functions of pressure and enthalpy. Different constitutive equations can be used in the hydraulics.
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8. RELATED AND AUXILIARY PROGRAMS:
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9. STATUS
Package ID Status date Status
NEA-0807/01 08-AUG-1983 Tested at NEADB
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10. REFERENCES:
NEA-0807/01, included references:
-  Rajamaki, Markku:
    TRAWA: A Transient Analysis Code for Water Reactors. Espoo 1980.
    Technical Research Centre of Finland, Nuclear Engineering
    Laboratory, Report 24. 149 p. + app. 31 p.
-  Raiko, Riitta & Rajamaki, Markku:
    TRAWA: A Transient Analysis Code for Water Reactors. Supplem.
    Part 1. Helsinki 1978. Technical Research Centre of Finland,
    Nuclear Engineering Laboratory, Report 33. 54 p.
-  Kyrki, Riitta:
    NRF Benchmark Problem 1980, Specification and Results of
    Finland. Helsinki 1980. Technical Research Centre of Finland,
    Nuclear Engineering Laboratory, Technical Report REP-14/80.
    8 + 28 p.
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11. MACHINE REQUIREMENTS:  74 Kwords of 36 bits.
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12. PROGRAMMING LANGUAGE(S) USED
Package ID Computer language
NEA-0807/01 FORTRAN-V (UNIVAC)
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13. OPERATING SYSTEM UNDER WHICH PROGRAM IS EXECUTED

UNIVAC 1108 EXEC  8. UNIVAC 1108 MATH PACK library routine GJR is used to solve a group of simultaneous linear equations.
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14. OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS:
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15. NAME AND ESTABLISHMENT OF AUTHOR

          M. Rajamaki
          Technical Research Centre of Finland
          Nuclear Engineering Laboratory
          P.O. Box 169
          00181 Helsinki 18
          Finland
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16. MATERIAL AVAILABLE
NEA-0807/01
File name File description Records
NEA0807_01.003 TRAWA INFORMATION FILE 69
NEA0807_01.004 TRAWA SOURCE PROGRAM (FORTRAN-66) 9400
NEA0807_01.005 TRAWA INPUT DATA FOR TEST CASE 240
NEA0807_01.006 TRAWA OUTPUT OF TEST CASE 3302
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17. CATEGORIES
  • F. Space - Time Kinetics, Coupled Neutronics - Hydrodynamics - Thermodynamics
  • H. Heat Transfer and Fluid Flow

Keywords: heat transfer, neutron diffusion equation, nuclear power plants, reactor dynamics, thermal conductivity, thermal reactors, thermodynamics, transients, water cooled reactors.