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NEA-0649 CARMEN-SYSTEM.

CARMEN-SYSTEM, Programs System for Thermal Neutron Diffusion and Burnup with Feedback

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1. NAME OR DESIGNATION OF PROGRAM:  CARMEN-SYSTEM.
A code system for neutronics PWR calculation by diffusion theory with space-dependent feedback effects.
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2. COMPUTERS
To submit a request, click below on the link of the version you wish to order. Only liaison officers are authorised to submit online requests. Rules for requesters are available here.
Program name Package id Status Status date
CARMEN-SYSTEM NEA-0649/02 Tested 18-OCT-1983
CARMEN-SYSTEM NEA-0649/03 Tested 09-NOV-1983

Machines used:

Package ID Orig. computer Test computer
NEA-0649/02 UNIVAC 1110 UNIVAC 1110
NEA-0649/03 CDC CYBER 740 CDC CYBER 740
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3. DESCRIPTION OF PROBLEM OR FUNCTION

CARMEN is a system of programs  developed for the neutronic calculation of PWR cycles. It includes the whole chain of analysis from cell calculations to core calcula-  tions with burnup. The core calculations are based on diffusion theory with cross sections depending on the relevant space-dependent feedback effects which are present at each moment along the cycles.

The diffusion calculations are in one, two or three dimensions and in two energy groups. The feedback effects which are treated locally are: burnup, water density, power density and fission products. In order to study in detail these parameters the core should be divided into as many zones as different cross section sets are expected to be required in order to reproduce reality correctly. A relevant difference in any feedback parameter between zones produces diffe- rent cross section sets for the corresponding zones.

CARMEN is also capable to perform the following calculations:
-  Multiplication factor by burnup step with fixed boron concentra-
   tion
-  Buckling and control rod insertion
-  Buckling search by burnup step
-  Boron search by burnup step
-  Control rod insertion search by burnup step.
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4. METHOD OF SOLUTION

The cell code (LEOPARD-TRACA) generates the fuel assembly cross sections versus burnup. This is the basic libra- ry to be used in the CARMEN code proper.

With a planar distribution guess for power density, water density and fluxes, the macroscopic cross sections by zone are calculated by CARMEN, and then a diffusion calculation is done in the whole geometry. With the distribution of power density, heat accumulated in the coolant and the thermal and fast fluxes determined in the diffusion calculation, CARMEN calculates the values of the most re-  levant parameters that influence the macroscopic cross sections by zone: burnup, water density, effective fuel temperature and fission  product concentrations.

If these parameters by zone are different from the reference values  included in the basic library, the cross sections by zone are cor- rected by these feedback effects and a new diffusion calculation is  made with them, and the process continues until power convergence is reached between two successive diffusion calculations. Then the next burnup step starts. In one execution of CARMEN the whole burnup cycle may be calculated step by step.
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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM

The present version  of CARMEN has the following restrictions:
Number of rows, columns and planes.....................210
Number of groups.........................................2
Number of zones.........................................99
Number of nuclides in a cross section set................2
Number of coarse meshes in each direction..............150
Number of cross section sets............................99
Number of burnup steps to calculate.....................15
Number of burnup steps in the basic library.............20

When CARMEN is executed without burnup option, the following re- strictions apply:
Number of rows..........................................210
Number of groups.......................................1000
Number of zones.........................................200
Number of nuclides in a cross section set...............200
Number of coarse meshes in each direction...............150
Number of cross section sets.............................99
Largest nuclide number..................................200

These limitations may be changed in the source program.
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6. TYPICAL RUNNING TIME

Each diffusion calculation needs about 0.15 minutes of CPU time. Each burnup step requires 3 or 4 itera- tions. Therefore each burnup step requires about 0.6 minutes. Criti- cality search calculations require more iterations.
NEA-0649/02
The test case was executed by NEA-DB on UNIVAC 1110 in  120 CPU seconds.
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7. UNUSUAL FEATURES OF THE PROGRAM:
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8. RELATED AND AUXILIARY PROGRAMS

LEOPARD-TRACA and LIBRA, included in the CARMEN system, generate the basic library of cross sections and the feedback parameters to be used directly in CARMEN proper.

CITATION is the diffusion calculation module in CARMEN. It may op- tionally be executed as stand-alone code.

MARIA is an alternative system to provide a cross section library for CARMEN.
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9. STATUS
Package ID Status date Status
NEA-0649/02 18-OCT-1983 Tested at NEADB
NEA-0649/03 09-NOV-1983 Tested at NEADB
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10. REFERENCES

- C. Ahnert, J.M. Martinez-Val:
  PENELOPE-CITATION: Code System for Nuclear Reactor Neutronics
  Calculations by Diffusion Theory with Cross Section Feedback
  JEN-300 (1975) (In Spanish)
- R.F. Barry:
  LEOPARD, A Spectrum Dependent Non-Spatial Depletion Code for the
  IBM-7094
  WCAP-3269 (1963)
- J.M. Aragones, C. Ahnert:
  MARIA System, A Code Block for PWR Fuel Assembly Calculations
  JEN, to be published in 1983
NEA-0649/02, included references:
- C. Ahnert and J.M. Aragones:
  CARMEN-SYSTEM: A Code Block for Neutronic PWR Calculation by
  Diffusion Theory with Spacedependent Feedback Effects.
  J.E.N.515  (1982)
- T.B. Fowler, D.R. Vondy and G.W. Cunningham:
  Nuclear Reactor Core Analysis Code CITATION.
  ORNL-TM-2496, Revision 2  (July 1971)
- In-Core Fuel Management Programs for Nuclear Power Reactors
  IAEA-TECDOC-314  (October 1984)
NEA-0649/03, included references:
- T.B. Fowler, D.R. Vondy and G.W. Cunningham:
  Nuclear Reactor Core Analysis Code CITATION.
  ORNL-TM-2496, Revision 2  (July 1971)
- In-Core Fuel Management Programs for Nuclear Power Reactors
  IAEA-TECDOC-314  (October 1984)
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11. MACHINE REQUIREMENTS

A blank COMMON stores the largest variables.
At present its size has been fixed at 65000 words, giving an overall size of the program of 108 kwords. The COMMON may be adapted to available machine capacity.
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12. PROGRAMMING LANGUAGE(S) USED
Package ID Computer language
NEA-0649/02 FORTRAN-V (UNIVAC)
NEA-0649/03 FORTRAN-IV
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13. OPERATING SYSTEM UNDER WHICH PROGRAM IS EXECUTED:
EXEC-8 (UNIVAC 1100). NOS (CDC CYBER 740).
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14. OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS:
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15. NAME AND ESTABLISHMENT OF AUTHOR

This code system has been included in the Coordinated Research Programme (CRP) on "Codes Adaptable to Small and Medium-Size Com- puters Available in Developing Countries for In-Core Fuel Manage- ment" of the International Atomic Energy Agency.

- Carol Ahnert
  Junta de Energia Nuclear
  Avda Complutense
  Madrid, Spain

- Jose M. Aragones
  Universidad Politecnica de Madrid
  Castellana 80
  Madrid, Spain
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16. MATERIAL AVAILABLE
NEA-0649/02
File name File description Records
NEA0649_02.003 CARMEN-SYSTEM INFORMATION FILE 77
NEA0649_02.004 CARMEN SOURCE CARD IMAGES (FORTRAN-5) 30499
NEA0649_02.005 CARMEN JCL 39
NEA0649_02.006 CARMEN INPUT DATA FOR TEST CASE 130
NEA0649_02.007 CARMEN PRINTED OUTPUT OF TEST CASE 4593
NEA0649_02.008 CARMEN MAP 94
NEA0649_02.009 LIBRA SOURCE (FORTRAN-5) 82
NEA0649_02.010 LIBRA BCD LIBRARY CASE 1 320
NEA0649_02.011 LIBRA BCD LIBRARY CASE 2 320
NEA0649_02.012 LIBRA BCD LIBRARY CASE 3 320
NEA0649_02.013 LIBRA BCD LIBRARY CASE 4 20
NEA0649_02.014 LIBRA INPUT DATA FOR TEST CASE 1 2
NEA0649_02.015 LIBRA INPUT DATA FOR TEST CASE 2 2
NEA0649_02.016 LIBRA INPUT DATA FOR TEST CASE 3 2
NEA0649_02.017 LIBRA INPUT DATA FOR TEST CASE 4 2
NEA0649_02.018 LEOPARD SOURCE (FORTRAN-5) 3540
NEA0649_02.019 LEOPARD JCL 19
NEA0649_02.020 LEOPARD INPUT DATA FOR TEST CASE 20
NEA0649_02.021 LEOPARD OUTPUT OF TEST CASE 839
NEA0649_02.022 SPOTS SOURCE (FORTRAN-5) 823
NEA0649_02.023 SPOTS INPUT DATA FOR TEST CASE 3789
NEA0649_02.024 SPOTS OUTPUT OF TEST CASE 2865
NEA-0649/03
File name File description Records
NEA0649_03.003 CARMEN-SYSTEM INFORMATION FILE 33
NEA0649_03.004 CARMEN-SYSTEM JCL TO RUN TEST CASE 131
NEA0649_03.005 SPOTS SOURCE (FORTRAN-4) 817
NEA0649_03.006 SPOTS INPUT DATA (FORMATTED LIBRARY) 3798
NEA0649_03.007 LEOPARD SOURCE (FORTRAN-4) 3644
NEA0649_03.008 LEOPARD INPUT DATA FOR 4 TEST CASES 97
NEA0649_03.009 LEOPARD PUNCHED OUTPUT 980
NEA0649_03.010 LIBRA SOURCE (FORTRAN-4) 84
NEA0649_03.011 LIBRA INPUT DATA FOR 4 TEST CASES 8
NEA0649_03.012 CARMEN SOURCE SEGMENT 1 1679
NEA0649_03.013 CARMEN SOURCE SEGMENT 2 4356
NEA0649_03.014 CARMEN SOURCE SEGMENT 3 3995
NEA0649_03.015 CARMEN SOURCE SEGMENT 4 4515
NEA0649_03.016 CARMEN SOURCE SEGMENT 5 3023
NEA0649_03.017 CARMEN SOURCE SEGMENT 6 3185
NEA0649_03.018 CARMEN SOURCE SEGMENT 7 4296
NEA0649_03.019 CARMEN SOURCE SEGMENT 8 4100
NEA0649_03.020 CARMEN SOURCE SEGMENT 9 1245
NEA0649_03.021 CARMEN DIRECTIVES 49
NEA0649_03.022 CARMEN INPUT DATA FOR TEST CASE 130
NEA0649_03.023 CARMEN PRINTED OUTPUT OF TEST CASE 4589
NEA0649_03.024 SPOTS PRINTED OUTPUT OF TEST CASE 2861
NEA0649_03.025 LEOPARD PRINTED OUTPUT OF 4 TEST CASES 9873
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17. CATEGORIES
  • C. Static Design Studies

Keywords: Doppler broadening, boron, burnup, cross sections, diffusion, feedback, fission products, interpolation, isotopes, pwr reactors, thermodynamics.