|Program name||Package id||Status||Status date|
|MURE V2 - SMURE||NEA-1845/02||Arrived||17-FEB-2017|
|Package ID||Orig. computer||Test computer|
The main aim of the (S)MURE package is to perform nuclear reactor time-evolution using successive calls to the widely-used particle transport codes MCNP or Serpent.
(S)MURE is an Object-Oriented package and therefore users are free to interact with it in their own way or to use the evolution controls already developed.
MURE also provides coupling of the neutronics (with or without fuel burn-up) and thermal-hydraulics using a sub-channel 3D code, COBRA-EN.
A graphical interface is provided to visualize and post-treat the results, including radiotoxicity calculations, waste heats, etc.
An interface to NJOY to generate cross-section in the MCNP ACE format (ENDF2ACE) is also provided in the MURE package.
See http://lpsc.in2p3.fr/MURE/html/MURE/MURE.html for more details.
(S)MURE provides an interface to MCNP or Serpent to build complex geometries using Object-Oriented programming and/or the ability to calculate nuclear fuel depletion. Moreover, it is very easy to modify a MURE input to switch from MCNP to Serpent or vice-versa.
Neutron transport is performed by Monte-Carlo(MC) transport code (MCNP/MCNPX or Serpent2) and depletion is calculated using numerical integration via the Runge-Kutta algorithm. Successive MC runs and Bateman equation resolutions are performed until the end of the evolution time. Interactions during the evolution calculation allow the user to impose conditions such as power levels, constant keff, flux, control rods position, etc. ... It is easy for the user to implement their own evolution control owing to the Object-Oriented programming and inheritance mechanism.
Standard evolutions evaluate one-group constant reaction rates between 2 MC runs for solving the Bateman equations at each step. However, Predictor-Corrector methods can also be used, as well as "quasi-multi group" flux where reaction rates are calculated outside of MC runs from flux tallies for each cell with a highly discretized energy binning. Reaction rates in this method are calculated after each MC run using the same ACE cross-section files that were used in the neutron transport ; the advantage of this method is a large CPU time gain in MC runs(at least a factor 30).
For high energy physics (above 20 MeV), very little testing has been performed. To date, MURE has only been used with neutron transport (electrons, photons, protons have not been transported and fuel evolution involving reactions induced by these particles is not performed).
Auxiliary programs included in the distribution:
MureGui: GUI to visualize and post-treat MURE evolution results
ENDF2ACE: Interface to NJOY to prepare ACE format files for MCNP from ENDF cross-section library.
ExtractTree/ExtractXsdir: build the BaseSummary.dat file, needed for MURE evolution
GenerateFPYield: generate binary fission product yield file from an ENDF fission product yield file
Not included in the distribution package:
MCNP or MCNPX (CCC-0740)
|Package ID||Status date||Status|
Méplan O., Nuttin A., Laulan O., David S., Michel-Sendis F. et al.:
MURE : MCNP Utility for Reactor Evolution - Description of the methods, first applications and results, Proceedings of the ENC 2005 (CD-Rom) - ENC 2005 - European Nuclear Conference. Nuclear Power for the XXIst Century : From basic research to high-tech industry, France
Michel-Sendis F., Méplan O., David S., Nuttin A., Bidaud A. et al.:
Plutonium incineration and uranium 233 production in thorium fueled light water reactors, GLOBAL 2005 Proceedings (CD-Rom) - GLOBAL 2005: International Conference on Nuclear Energy Systems for Future Generation and Global Sustainability, Japan
|Package ID||Computer language|
a C++ compiler (such as g++ of GCC)
At least one of the 2 following MC codes:
- MCNP or MCNPX (CCC-0740)
- SERPENT (NEA-1840) ; the required version is SERPENT2 available at http://montecarlo.vtt.fi/
If coupling with thermics and thermal-hydraulics is required: COBRA-EN (NEA-1614)
For Graphical User Interface (GUI): ROOT (http://root.cern.ch)
If the user needs to use ENDF2ACE: NJOY99 is required (PSR-0480)
Keywords: depletion, fission products, fuel management, inventories, monte carlo method, neutron.