MURE v2 - SMURE: Serpent - MCNP Utility for Reactor Evolution.
AUXILIARY PROGRAMS:
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 ENDF fission product yield file
Not included in the distribution package:
COBRA-EN (P00507IBMPC00)
MCNP5/MCNPX (C00740MNYCP02)
NJOY99 (P00480MNYCP00)
SERPENT2 (Available at http://montecarlo.vtt.fi/)
Laboratoire de Physique Subatomique et de Cosmologie de Grenoble, France
Institut de Physique Nucleaire d'Orsay, France
Through the OECD Nuclear Energy Agency Data Bank, Issy-Les Moulineaux, France.
C++ on a LINUX or UNIX Operating System
RSICC ID: C00764MNYWS01; NEADB identifier is NEA-1845/002
The main aim of the MURE package is to perform nuclear reactor time-evolution using successive calls to the widely used particle transport code Monte Carlo N-Particle (MCNP) or Serpent. (S)MURE is an object-oriented package; 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, and other results. An interface to NJOY to generate cross sections 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 MCNP/MCNPX or Serpent2, and depletion is calculated using numerical integration via the Runge-Kutta algorithm. Successive MCNP 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, and others. Users can easily implement their own evolution controls owing to the object-oriented programming and inheritance mechanism. Standard evolutions evaluate one-group constant reaction rates between two MCNP runs for solving the Bateman equations at each step. However, predictor-corrector methods can also be used, as well as quasi-multi group flux, in which reaction rates are calculated outside of MCNP from flux tallies for each cell, with a highly discretized energy binning. Reaction rates in this method are calculated after each MCNP 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 MCNP (by 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, and protons have not been transported, and fuel evolution involving reactions induced by these particles is not performed.
On a 2.4 GHz Pentium 4, the compilation of MURE takes about 2 minutes. Run times depend mainly on MCNP run times; for the evolving examples provided, and using the quasi-multi group approach, the typical run time is about 10 minutes.
MURE runs on the LINUX operating system and most likely will run on any UNIX computer. Expanding and compiling the code system requires 120 MB of hard disk space.
· A C++ compiler (such as g++ of GCC)
· MCNP or MCNPX (C00740MNYCP02) or SERPENT2 (available at http://montecarlo.vtt.fi/)
· If coupling with thermics and thermal-hydraulics is required: COBRA-EN (P00507IBMPC00)
· For graphical user interface (GUI): ROOT (http://root.cern.ch)
· For radiotoxicity post processing calculations: LAPACK library (available for any LINUX distribution or at http://www.netlib.org/lapack)
· If ENDF2ACE is needed: NJOY is required (NJOY99, P00480MNYCP00)
·
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.
Included in documentation
· MURE 2: SMURE, Serpent, MCNP Utility for Reactor Evolution, User Guide Version 1.0, LPSC report no. LPSC-17002 (February 2017).
· FAQ on MURE package.
· L. Perot (IPNOrsay), O. Meplan (LPSC Grenoble): ENDF2ACEUserGuide, February 2017 (PDF).
· MURE Project Documentation, February 2017 (HTML).
The package is transmitted digitally as a download link in a .tar format, which includes reference material, documentation, source files, and sample problems. No executables are included with the package.
June 2021
KEYWORDS: COST ANALYSIS, DEPLETION, FISSION PRODUCTS, FUEL MANAGEMENT, INVENTORIES, MONTE CARLO METHOD, NEUTRON.