last modified: 16-MAY-2011 | catalog | categories | new | search |

NEA-1080 FEMAXI-6.

FEMAXI-6, Thermal and Mechanical Behaviour of LWR Fuel Rods

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1. NAME OR DESIGNATION OF PROGRAM:  FEMAXI-6 Ver.1(U).
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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
FEMAXI-6 VER.1(U) NEA-1080/10 Tested 16-MAY-2011

Machines used:

Package ID Orig. computer Test computer
NEA-1080/10 PC Windows PC Windows
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3. DESCRIPTION OF PROGRAM OR FUNCTION

FEMAXI-6(Updated) predicts the thermal and mechanical behaviour of a light water reactor fuel rod during normal and transient (not accident) conditions. It can analyse the integral behaviour of a whole fuel rod throughout its life as well as the localised behaviour of a small part of fuel rod. Temperature distribution, radial and axial deformations, fission gas release, and inner gas pressure are calculated as a function of irradiation time and axial position. Stresses and strains in the pellet and cladding are calculated and PCMI analysis is performed. Also, thermal conductivity degradation of pellet and cladding waterside oxidation are modeled. Its analytical capabilities also cover the boiling transient anticipated in BWR.
  
RODBURN calculates the power generation density profile in the radial and axial directions and fast neutron flux, and concentrations of fission product isotopes and fissile materials of a single rod irradiated in PWR, BWR and Halden BWR. RODBURN gives an output file which can be read by FEMAXI-6.
NEA-1080/10
This version differs from the previous one in the following:
- a few formulae were updated in the manual and the source code
- the input options were expanded in the following points:
  . Thermal expansion modelling
  . Pellet swelling option
  . Pellet plasticity model
  . Cladding surface heat transfer model

All changes are marked in red in the reference report.
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4. METHOD OF SOLUTION

FEMAXI-6: Elasto-plasticity, creep, thermal expansion, pellet cracking and crack healing, relocation, densification, swelling, hot pressing, heat generation distribution, fission gas release, pellet-cladding mechanical interaction, cladding creep and oxidation are modelled by the code. Efforts have been made to improve the numerical accuracy and stability of transient analysis.
  
RODBURN uses ORIGEN1 library, RABBLE library and other libraries. First, resonance integral is calculated. Then, three group constants are selected and simplified neutronics analysis is performed along the designated power history and rod geometries.
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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM:  None
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6. TYPICAL RUNNING TIME

Less than 1 minute in the sample test case on a Windows PC.
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7. UNUSUAL FEATURES OF THE PROGRAM

RODBURN can provide to the FEMAXI-6 a result file so that FEMAXI-6 can read it as one calculation condition for determining the power generation density and the fast flux in the axial and radial direction of the rod.
FEMAXI-6 can read the result files of RODBURN and PLUTON, and use them as one of calculation conditions to determine the power generation density in the radial direction of rod.
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8. RELATED AND AUXILIARY PROGRAMS

RODBURN for Power Profiles and Isotopics in PWR, BWR Fuel Rods (included)
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9. STATUS
Package ID Status date Status
NEA-1080/10 16-MAY-2011 Tested at NEADB
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10. REFERENCES

- Masaaki Uchida and Hiroaki Saito:
RODBURN: A Code for Calculating Power Distribution in Fuel Rods, JAERI-M 93-108 (1993) (in Japanese)
NEA-1080/10, included references:
- Motoe SUZUKI and Hiroaki SAITOU:
Light Water Reactor Fuel Analysis Code FEMAXI-6 (Ver.1) -Detailed Structure and
User's Manual-, JAEA-Data/Code 2005-003 (2006)
- RODBURN input manual
- RODBURN internal structure schematics
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11. MACHINE REQUIREMENTS:  Windows PC.
NEA-1080/10
At the NEA Databank, the code was tested on:
- COMPUTER : Dell INTEL Duo E6550,  2.33 GHz (for Windows)        
- OPERATING SYSTEM : DOS, under Windows XP Pro SP 3
- COMPILER: lahey/ Fujitsu F 95  v7.1  (Windows)
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12. PROGRAMMING LANGUAGE(S) USED
Package ID Computer language
NEA-1080/10 FORTRAN-77, FORTRAN-90
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13. OPERATING SYSTEM UNDER WHICH PROGRAM IS EXECUTED:  Windows XP, Acrobat Reader including Japanese fonts.
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14. OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS

See the installation guide document telling the hint to install and compile.
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15. NAME AND ESTABLISHMENT OF AUTHORS

Motoe SUZUKI
Reactor Safety Research Group
Nuclear Safety Research Center
Japan Atomic Energy Agency
Tokai-mura, Naka-gun, Ibaraki-ken 319-1195
JAPAN
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16. MATERIAL AVAILABLE
NEA-1080/10
source program
name-list parameter format file
libraries
sample
input/output files for test run
graphic output source program
sample data file for graphic program control
description documents for models
materials properties and empirical equations
input manual
RODBURN source program, libraries, sample input data for test run,
sample output, input manual, and installation guide document.
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17. CATEGORIES
  • D. Depletion, Fuel Management, Cost Analysis, and Power Plant Economics
  • I. Deformation and Stress Distributions, Structural Analysis and Engineering Design Studies

Keywords: boiling transient, burnup profile, fast neutron flux, finite element method, fission gas release, fission products, fuel rods, fuel-cladding interaction, heat generation density profile, light-water reactors, mechanical properties, pellet clad mechanical interaction, pellet thermal conductivity degradation, radial profile of heat generation, stress analysis, thermal stresses, transient analysis.