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NEA-0655 VSOP.

VSOP, Neutron Spectra, 2-D Flux Synthesis, Fuel Management, Thermohydraulics Calculation

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1. NAME OR DESIGNATION OF PROGRAM:  VSOP program system.
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2. COMPUTERS
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Program name Package id Status Status date
VSOP94 NEA-0655/04 Tested 25-JUL-1995

Machines used:

Package ID Orig. computer Test computer
NEA-0655/04 Many Computers DEC VAX 6000
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3. DESCRIPTION OF PROBLEM OR FUNCTION

VSOP (Very Superior Old Programs) is a system of codes linked together for the simulation of reactor life histories. It comprises neutron cross section libraries and processing routines, repeated neutron spectrum evaluation, 2-D and 3-D diffusion calculation, depletion and shut-down features, in- core and out-of-pile fuel management, fuel cycle cost analysis, and  thermal hydraulics (steady state and transient). Various techniques  have been employed to accelerate the iterative processes and to optimize the internal data transfer.
The code system has been used extensively for comparison studies of thermal reactors, their fuel cycles, thermal transients, and safety assessment. Besides its use in research and development work  for the Gas Cooled High Temperature Reactor, the system has been applied successfully to Light Water and Heavy Water Reactors, MAGNOX, and RBMK.
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4. METHOD OF SOLUTION

The nuclear data for 184 isotopes are contained in two libraries. Fast and epithermal data in a 68 group GAM-I structure have been prepared mainly from ENDF/B-V and JEF-1. Resonance cross section data are given as input. Thermal data in a 30 group THERMOS structure have been collapsed from a 96 group THERMALIZATION (GATHER) library by a relevant neutron energy spectrum generated by the THERMALIZATION code. Graphite scattering matrices are based on the Young phonon spectrum in graphite.

The neutron spectrum is calculated by a combination of the GAM and THERMOS codes. They can simultaneously be employed for many core regions differing in temperature, burnup, and fuel element lay-out.  The thermal cell code THERMOS has been extended to treat the grain structure of the coated particles inside the fuel elements, and the  epithermal GAM code uses modified cross sections for the resonance absorbers prepared from double heterogeneous ZUT-DGL calculations.

The diffusion module of the code is CITATION with 2 - 8 energy groups. It provides the neutron flux for 1515 compositions in 2-D cases, r-z (9999 compositions in 3-D cases, x-y-z). The burnup scheme has been developed from the FEVER code. The build-up history  of up to 49 fission product nuclides in the compositions is followed explicitly. The diffusion part of the program system can be repeated at many short burnup time steps, and the spectrum module can be repeated at larger time steps, when some significant change in the spectrum is expected.

The fuel management and cost module performs the fuel shuffling and general evaluations of the reactor and fuel element life history. The fuel management simulates the currently known shuffling and out of pile routes for various reactors. It has further been extended to include the typical features of the pebble bed reactor such as burnup dependent optional reloading of elements, separated treatment of different fuel streams, and recycling in new fuel element types according to a consistent mass balance and timing.

Optionally, several different types of data files can be set up with characteristic data of the reactor life. These are used for more detailed investigations and display programs. The restart option allows the study of special phases of the reactor life, e.g.  changes of the fueling scheme, of the burnup, of the power output, of the coolant temperature, and of the corresponding reactivity effects. The fuel cycle cost data set is made for the present worth  KPD code. Two-dimensional thermal hydraulics studies for operating and emergency conditions can be performed with the THERMIX code. The averaged temperatures of the different spectrum zones in the core are returned from the thermal hydraulics to the subsequent step of the reactor history.
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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM

In epithermal energy range the cell spectrum calculation is missing. If needed, it must be simulated by disadvantage factors being obtained in other codes.  Further, dynamic common must be defined for the commons VARDIM, COCI, and for VARPRI (in the auxiliary program PRIOR). This is not needed under the workstation operating systems.
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6. TYPICAL RUNNING TIME

The simulation of a 10 year reactor life history with 12 spectrum runs and 20 diffusion calculations requires around 9 minutes. of CPU time on the IBM E/9000-620.
NEA-0655/04
The 12 test case included in this package have been executed by NEA-DB on a DEC VAX 6000-510 computer in about three hours of CPU time.
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7. UNUSUAL FEATURES OF THE PROGRAM

No unusual features, since the program system is based on well-established and proven codes.
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8. RELATED AND AUXILIARY PROGRAMS

VSOP contains the codes: DATA-2, BIRGIT, TRIGIT, ZUT-DGL, GAM-1, THERMOS, FEVER, KPD, THERMIX, LIFE,  PRIOR and ATALS. The code system uses thermal data collapsed from a  THERMALIZATION/GATHER library using a spectrum generated by the THERMALIZATION code.

The precursor of VSOP was the MAFIA-II code.
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9. STATUS
Package ID Status date Status
NEA-0655/04 25-JUL-1995 Tested at NEADB
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10. REFERENCES

- L. Massimo
"MAFIA-II, A One-Dimensional Burnup Code"
Argonne National Laboratory, ANL-7050 (1966).
- G. D. Joanou, et al.:
"GATHER-II, an IBM-7090 Fortran II Program for the Computation of
Thermal Spectra and Associated Multigroup Cross Sections"
General Atomic GA-4132 (1963).
- J. Darvas
"DATA-2, Head Programm zum VSOP Zyklus fuer
Hochtemperaturreaktoren"
KFA Internal Report, KFA-IRE-70-4 (1970).
- G.D. Joanou, J.S. Dudek:
"GAM - A Consistent P1 Multigroup Code for the Calculation of Fast
Neutron Spectra and Multigroup Constants"
General Atomic GA-1850 (1961).
- H C. Honeck:
"THERMOS - A Thermalization Transport Theory Code for Reactor
Lattice Calculation"
Brookhaven National Laboratory BNL-5826 (1961).
- F. Todt:
"FEVER - A One-Dimensional Few Group Depletion Program for Reactor
Analysis"
General Dynamics - General Atomic, GA-2749 (1962).
- U. Hansen:
"The VSOP System Present Worth Fuel Cycle Calculation Methods and
Codes, KPD"
Atomic Energy Establishment Winfrith, England, Dragon Project
Report 915 (1975).
- K. Petersen, A. Verfondern:
"THERMIX, ein Computerprogramm zur Berechnung des instationaeren
Temperaturverhaltens gasdurchstroemter Kugelschuettungen"
to be published.
- E. Teuchert, K.A. Haas:
"ZUT-DGL-V.S.O.P. Programmzyklus fuer die Resonanzabsorption in
heterogenen Anordnungen"
Internal Report, KFA-IRE-70-1 (1970).
NEA-0655/04, included references:
- E. Teuchert et al:
  V.S.O.P.('94) - Computer Code System for Reactor Physics and Fuel
  Cycle Simulation - Input Manual and Comments
  (April 1994).
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11. MACHINE REQUIREMENTS

Main storage requirement in IBM E/9000-620 is 17 Mbytes for the test case. In the subroutine NACHW the code applies the IBM system routine "MASKE" for suppression of underflows. The subroutines FRIST, ZEITN, ZEIT represent machine dependent time routines. DATUM represents the date, and JOBNAM the name of the job. For the transfer to other machines, they are added  as dummy routines which yield zeroes.
NEA 0655/04: NEA-DB installed the program on a DEC VAX 6000-510 computer.
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12. PROGRAMMING LANGUAGE(S) USED
Package ID Computer language
NEA-0655/04 FORTRAN-77
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13. OPERATING SYSTEM UNDER WHICH PROGRAM IS EXECUTED

MVS at the IBM E/9000-620. AIX V3.2.5 at the IBM RISC System/6000.
NEA-0655/04
The program was installed by NEA-DB on a VAX 6000 running under VMS 6.1. The source code was compiled with the VAX/VMS FORTRAN compiler Version 6.2.
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14. OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS

NEA 0655/04: In this version of the program, output format statements have been translated from the original German into English.
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15. NAME AND ESTABLISHMENT OF AUTHOR

          E. Teuchert, K.A. Haas
          Forschungszentrum Juelich GmbH. (KFA)
          Institut fuer Sicherheitsforschung ind Reaktortechnik
          D-52425 Juelich
          Germany
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16. MATERIAL AVAILABLE
NEA-0655/04
File name File description Records
NEA0655_04.001 Information file 248
NEA0655_04.002 GAM-library (68 groups) in ASCII 49799
NEA0655_04.003 THERMALIZATION-library (96 groups) in ASCII 119388
NEA0655_04.004 Th-232 Resonance data (formatted) 262
NEA0655_04.005 U-238 Resonance data (formatted) 165
NEA0655_04.006 General organization 1432
NEA0655_04.007 Input 2892
NEA0655_04.008 Organization of spectrum calculation 1 1066
NEA0655_04.009 Organization of spectrum calculation 2 1622
NEA0655_04.010 Neutron spectrum 1 987
NEA0655_04.011 Neutron spectrum 2 2196
NEA0655_04.012 Burnup 1 2973
NEA0655_04.013 Burnup 2 3011
NEA0655_04.014 Fuel management, fuel cycle costs 1 1564
NEA0655_04.015 Fuel management, fuel cycle costs 2 2459
NEA0655_04.016 Fuel management, fuel cycle costs 3 2341
NEA0655_04.017 Thermal hydraulics 1 2459
NEA0655_04.018 Thermal hydraulics 2 2358
NEA0655_04.019 Thermal hydraulics 3 2443
NEA0655_04.020 Thermal hydraulics 4 2567
NEA0655_04.021 Thermal hydraulics 5 2463
NEA0655_04.022 Thermal hydraulics 6 2439
NEA0655_04.023 Decay heat 556
NEA0655_04.024 Diffusion calculation 1 2922
NEA0655_04.025 Diffusion calculation 2 2855
NEA0655_04.026 Diffusion calculation 3 3335
NEA0655_04.027 Diffusion calculation 4 2497
NEA0655_04.028 Diffusion calculation 5 2906
NEA0655_04.029 Coupling between thermal hydrau. & neutronic 926
NEA0655_04.030 Dummy subroutines for VAX 16
NEA0655_04.031 Fuel elements and input materials 1562
NEA0655_04.032 Resonance absorption cross sections 1 2254
NEA0655_04.033 Resonance absorption cross sections 2 2440
NEA0655_04.034 Reactor geometry, 2-dim 2124
NEA0655_04.035 Reactor geometry, 3-dim 359
NEA0655_04.036 Converting program for lib GAM.LIB 55
NEA0655_04.037 Converting program for lib THERMA.LIB 53
NEA0655_04.038 Compiling fuel life for decay heat evaluat. 742
NEA0655_04.039 Compiling fuel life for isotropic burnup 355
NEA0655_04.040 Printout of 3D distribution of power 433
NEA0655_04.041 Command file to install and run tests (IBM) 1873
NEA0655_04.042 Command file for compile and link steps(VAX) 62
NEA0655_04.043 Command file to run test cases (VAX) 236
NEA0655_04.044 Data to prepare UNIT 30 4
NEA0655_04.045 Data to calculate resonance integrals 31
NEA0655_04.046 Data to calculate resonance integrals 18
NEA0655_04.047 Data for fuel element design 40
NEA0655_04.048 Data to prepare a 30 group THERMOS-lib 8
NEA0655_04.049 Data to create the volume matrix 50
NEA0655_04.050 Data for geometric data design 50
NEA0655_04.051 Data to restart OTTO case with 2D. 30
NEA0655_04.052 Data to restart LOCA 30
NEA0655_04.053 Data to prepare 30 group THERMOS-lib 37
NEA0655_04.054 Data for startup core of the 100 MW OTTO 267
NEA0655_04.055 Data to restart OTTO case with 2D 137
NEA0655_04.056 Data to restart calc. of temp. coefficients 128
NEA0655_04.057 Data to restart lib. for LIFE(fuel elements) 23
NEA0655_04.058 Data to restart LOCA 188
NEA0655_04.059 Data to compile fuel life history for decay 4
NEA0655_04.060 Tests output 87140
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17. CATEGORIES
  • B. Spectrum Calculations, Generation of Group Constants and Cell Problems
  • C. Static Design Studies
  • D. Depletion, Fuel Management, Cost Analysis, and Power Plant Economics
  • H. Heat Transfer and Fluid Flow
  • K. Reactor Systems Analysis

Keywords: ENDF/B, HTR reactors, LWR reactors, burnup, cost, cross sections, depletion, diffusion, flux synthesis, fuel cycle, fuel management, graphite moderated reactors, heavy water cooled reactors, joint evaluated data file, neutron spectra, r-z, reactor kinetics, simulation, thermodynamics, two-dimensional, x-y-z.