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NESC0479 FREADM1

FREADM-1, Reactor Kinetics Thermohydraulics Calculation for Fast Reactor Accidents

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1. NAME OR DESIGNATION OF PROGRAM:  FREADM1
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2. COMPUTERS
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Program name Package id Status Status date
FREADM-1 NESC0479/01 Tested 01-DEC-1973

Machines used:

Package ID Orig. computer Test computer
NESC0479/01 IBM 370 series IBM 370 series
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3. DESCRIPTION OF PROBLEM OR FUNCTION

FREADM1  is a  fast  reactor,
multichannel, accident  analysis program  designed to  efficiently
simulate a reactor transient from initiation  to the point of core
disassembly.   Models are  included  for  nuclear kinetics  (point
model),  core  thermo-hydraulics,  voiding,  fuel  redistribution,
failure  propagation,  programmed reactivity  insertion,  and  the
dynamics of primary-system coolant flow.  A broad range of assumed
accident initiating  and propagating  activities may  be simulated
using triggering logic included in the code.
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4. METHOD OF SOLUTION

Time  integration of  the  equations for  the
dynamics  of  nuclear  kinetics, heat  transfer  and  primary-loop
coolant flow may be performed in a coupled mode, or independently.
Transient  temperature   conditions  computed  for   single  pins,
representative  of  a number  of  bundle  types,  may be  used  to
initiate   prescribed  accident   initiating  and/or   propagating
activities.   The  user  supplies criteria  for  initiating  these
activities.  Reactivity  feedback from the various  activities may
be computed using models in the code, or input from tables.
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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM

FREADM1         is
restricted  to accidents  which initiate  and propagate  uniformly
within annular or cylindrical coaxial core regions.
   Heat transfer calculations for up to 12 radial core regions may
be performed  simultaneously with up  to 12 separate  bundle types
subdivided into 9  or fewer axial sections.   Radial heat transfer
calculations are  done using up  to 10  radial fuel nodes,  a clad
node and a coolant node for each  axial section.  Loop flow may be
treated using a  maximum of two independent  primary-coolant loops
(loops may be of different sizes to simulate more than two loops).
Point kinetics with up to six  delayed neutron precurser groups is
used to compute the transient power level of the core.
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6. TYPICAL RUNNING TIME

Running  time   for  FREADM1   is  strongly
dependent on the  options utilized and  values supplied  for time-
step control.  200  kinetics time-steps with Doppler  feedback for
12 bundle types and 9 axial  sections require about 0.9 minutes of
GE635 processor time.  87 coupled kinetics and heat transfer time-
steps using 12 bundle types and 9 axial sections require about 4.4
minutes of GE635  processor time.  32 coupled  loop flow, kinetics
and heat  transfer time-steps  using 12 bundle  types and  9 axial
sections  require  about  5  minutes   of  GE635  processor  time.
Computation time per  time-step for heat transfer  calculations is
proportional to NBT *  NAX where NBT = number of  bundle types and
NAX =  number of axial  sections.  Computation time  per time-step
for an independent loop flow and  voiding calculations is about 15
seconds of GE635 processor time.
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7. UNUSUAL FEATURES OF THE PROGRAM

FREADM1 is highly  flexible both
in  the complexity  of the  problem  solved and  the selection  of
solution  methods.    Loop  flow,   kinetics  and   heat  transfer
calculations may be done separately or in a coupled mode.  Trigger
logic  provided  in  FREADM1  permits  the  sequence  of  accident
initiating and  propagating activities to  be determined  based on
computed  local conditions  reaching preset  levels  input by  the
user.  Tables are available which permit a flexible description of
the processes for  parametric studies.   By controlling  the time-
step  size  in  the  kinetics  calculation,  time-steps  for  heat
transfer and loop flow calculations  are performed with time-steps
determined by accuracy  requirements on these processes  only.  An
integral number of  kinetics steps may be taken  per heat transfer
step.  Input to the code is free form with much input optional.
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8. RELATED AND AUXILIARY PROGRAMS

FREADM1 will compute the accident
through  core disassembly.   Integration of  the nuclear  kinetics
equations is as performed in FORE2 (NESC Abstract 174).
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9. STATUS
Package ID Status date Status
NESC0479/01 01-DEC-1973 Tested at NEADB
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10. REFERENCES

- E.Y. Morikawa:
  ERRUNS - Entry in Library Subroutine ERRSYS,
  GESJ Note (November 11, 1968) and FSNOW, Save File Code Function.
NESC0479/01, included references:
- D.D. Freeman, E.G. Leff, D.J. Bender, and W.G. Meinhardt:
  The FREADM-1 Code - A Fast Reactor Excursion and Accident Dynamics
  Model
  GEAP-13608 (September 1970).
   FREADM1 Subroutine Descriptions including
   LINK and LLINK GE System Subroutines,
   ISERVE, GE 635 Computer Service Function,
   F4TRBK, F4TRAC, F4TRID, FORTRAN I/O Error Processing,
   ERRSYS, FORTRAN Compatible Error Subroutine.
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11. MACHINE REQUIREMENTS:  24K memory with 3 peripheral storage units
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12. PROGRAMMING LANGUAGE(S) USED
Package ID Computer language
NESC0479/01 FORTRAN-IV
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13. OPERATING SYSTEM UNDER WHICH PROGRAM IS EXECUTED:   GECOS-III.
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14. OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS:
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15. NAME AND ESTABLISHMENT OF AUTHORS

                 D. D. Freeman, E. G. Leff, D. J. Bender,
                 and W. G. Meinhardt
                 Breeder Reactor Development Operation
                 General Electric Company
                 Sunnyvale, California  94086
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16. MATERIAL AVAILABLE
NESC0479/01
File name File description Records
NESC0479_01.001 SOURCE PROGRAM (FORTRAN) 8779
NESC0479_01.002 ORIGINAL GE635-SYSTEM ROUTINES 591
NESC0479_01.003 SAMPLE PROBLEM 359
NESC0479_01.004 OUTPUT LIST OF SAMPLE PROBLEM 2466
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17. CATEGORIES
  • G. Radiological Safety, Hazard and Accident Analysis

Keywords: accidents, coolants, cylinders, fast reactors, heat transfer, kinetics, transients.