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NEA-0418 BRUCH-D-06.

BRUCH-D-06, LOCA of PWR Primary System with 23 Control Volume and 9 Rupture Points

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1. NAME OR DESIGNATION OF PROGRAM:  BRUCH-D-06.
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2. COMPUTERS
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Program name Package id Status Status date
BRUCH-D-06 NEA-0418/01 Tested 01-APR-1979
BRUCH-D-06 NEA-0418/05 Tested 15-APR-1983

Machines used:

Package ID Orig. computer Test computer
NEA-0418/01 IBM 360 series IBM 360 series
NEA-0418/05 CDC CYBER 174 CDC CYBER 174
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3. NATURE OF PHYSICAL PROBLEM SOLVED

BRUCH-D-06 has been developed for the study of a PWR  during a loss-of-coolant accident.
Simulation of a rupture of a recirculation line at nine positions with a variable break area is possible. Emergency coolant can be injected into six nodes. Simultaneous calculation of the average fuel rod temperature distribution and the heat transfer coefficient  is included as well as the calculation of some hot rods based on the fluid- and thermodynamic state of the coolant within the average channel. The overall heat transfer coefficients of the steam generator tubes can be varied.
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4. METHOD OF SOLUTION

BRUCH-D-06 is based on a multi-point model with  the assumption of uniform conditions and thermal equilibrium within  nodes. The model utilizes the conservation equations for mass, energy, and momentum together with the equation of state for water.  This leads to a system based on 23 nodes of 96 first order differential equations, most of which are non-linear. This system is integrated numerically by a special method (4).
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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM

The approximations for calculating the fluid properties are valid within a pressure range from 5 to 160 bar and an enthalpy range from 600 kJ/kg to 4000 kJ/kg.
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6. TYPICAL RUNNING TIME

A typical problem (double ended break) of 35 sec transient time requires about 30 min CPU time on the IBM 360/91.
NEA-0418/05
NEA-DB executed the sample case on CDC CYBER 174 in 50 minutes.
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7. UNUSUAL FEATURES OF THE PROGRAM

BRUCH-D-06 differs from BRUCH-D-05 in the following respects:
- The parallel recirculation lines of the reactor do not have to be  identical any longer. This means that one type of pump can be treated in the broken line and another type can be simulated in the intact line.
- The pump model has been changed.
- BRUCH-D-05 can handle a break in the hot or the cold leg between pump and pressure vessel.
BRUCH-D-06 can, in addition, handle a break between steam generator and pump.
- The water level in the pressure vessel can be determined using a simple model.
- It is also possible to simulate a system which consists of only the pressure vessel and an open ended pipe.
- A model has been added to account for the latent heat of the internal parts of the system.
- Electrically heated 'fuel rods' can also be simulated.
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8. RELATED AND AUXILIARY PROGRAMS:  BRUCH-D-03, -04, -05.
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9. STATUS
Package ID Status date Status
NEA-0418/01 01-APR-1979 Tested at NEADB
NEA-0418/05 15-APR-1983 Tested at NEADB
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10. REFERENCES

- H. Karwat and K. Wolfert:
  'BRUCH-D, A Digital Program for PWR-Blowdown Investigations'
  Nucl. Eng. Des. 11(1970) 241-254.
- K.J. Liesch, F. Steinhoff, I. Vojtek, and K. Wolfert:
  'BRUCH-D-02, A Computer Program for the Analysis of Fluid and
   Thermodynamic Processes in the Primary Circuit of Pressurized-Water Reactors in the Case of Loss-of-Coolant Accidents'
   Program Description
  MRR-P3 (July 1973), (in German).
- I. Vojtek and K.J. Liesch:
  'BRUCH-D03, Annex to the Program Description of BRUCH-D-02'
  MRR-P-O (1. Ergaenzung) (January 1974), (in German).
- E. Hofer and W. Werner:
'IMEX, A New Method for the Efficient Numerical Integration of the System of Ordinary Differential Equations Used in the Blow-Down
   Code BRUCH'
  MRR-128 (July 1973), (in German).
- K.J. Liesch and G. Raemhild:
  'BRUCH-D-04. A Computer Program for the Analysis of Fluid-and
   Thermodynamic Transients in the Primary Circuit of Pressurized Water Reactors in the Case of Loss-of-Coolant Accidents'
   (4th Extended Version).
   MRR-P-15 (July 1975), (in German).
NEA-0418/01, included references:
- K. Hofmann:
  BRUCH-D-06 - A Computer Program for the Analysis of Fluid-and
  Thermodynamic Transients in the Primary Circuit of Pressurized
  Water Reactors in the Case of Loss-of-Coolant Accidents
  Program Description
  MRR-P-25 (December 1976), (in German).
NEA-0418/05, included references:
- K. Hofmann:
  BRUCH-D-06 - A Computer Program for the Analysis of Fluid-and
  Thermodynamic Transients in the Primary Circuit of Pressurized
  Water Reactors in the Case of Loss-of-Coolant Accidents
  Program Description
  MRR-P-25 (December 1976), (in German).
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11. MACHINE REQUIREMENTS

To execute the sample case on CDC CYBER 174, 150,000 octal words of main storage are required.
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12. PROGRAMMING LANGUAGE(S) USED
Package ID Computer language
NEA-0418/01 FORTRAN-IV
NEA-0418/05 FORTRAN-IV
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13. OPERATING SYSTEM OR MONITOR UNDER WHICH PROGRAM IS EXECUTED:  NOS 1.4.531 (CDC CYBER 174).
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14. OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS

ANY OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS
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15. NAME AND ESTABLISHMENT OF AUTHOR

K.J. Liesch
Laboratorium fuer Reaktorregelung und Anlagensicherung
Garching
Technische Universitaet
Munich, Germany.F.R.
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16. MATERIAL AVAILABLE
NEA-0418/01
File name File description Records
NEA0418_01.001 INFORMATION 31
NEA0418_01.002 SOURCE PROGRAM (F4,EBCDIC) 7887
NEA0418_01.003 SAMPLE CASE 1 -- JCL 33
NEA0418_01.004 SAMPLE CASE 1 -- INPUT DATA 170
NEA0418_01.005 SAMPLE CASE 1 -- PRINTED OUTPUT 4321
NEA0418_01.006 SAMPLE CASE 2 -- JCL 29
NEA0418_01.007 SAMPLE CASE 2 -- INPUT DATA 170
NEA0418_01.008 SAMPLE CASE 2 -- PRINTED OUTPUT 1577
NEA0418_01.009 SAMPLE CASE 3 -- INPUT DATA 29
NEA0418_01.010 SAMPLE CASE 3 -- PRINTED OUTPUT 300
NEA-0418/05
File name File description Records
NEA0418_05.003 BRUCH-D-06 INFORMATION FILE 22
NEA0418_05.004 BRUCH-D-06 JCL 11
NEA0418_05.005 BRUCH-D-06 SOURCE UPDATE (FORTRAN-4) 7764
NEA0418_05.006 BRUCH-D-06 SOURCE CARD IMAGES (FORTRAN-4) 7738
NEA0418_05.007 BRUCH-D-06 INPUT DATA FOR TEST CASE 170
NEA0418_05.008 BRUCH-D-06 OUTPUT OF TEST CASE 4322
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17. CATEGORIES
  • G. Radiological Safety, Hazard and Accident Analysis

Keywords: blowdown, heat transfer, loss-of-coolant accident, pwr reactors, reactor safety, steam generators, thermodynamics.