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| Program name | Package id | Status | Status date |
|---|---|---|---|
| MCB1C | NEA-1643/01 | Tested | 30-AUG-2002 |
Machines used:
| Package ID | Orig. computer | Test computer |
|---|---|---|
| NEA-1643/01 | Linux-based PC,UNIX W.S. | Linux-based PC |
MCB is a Monte Carlo Continuous Energy Burnup Code for a general-purpose use to calculate a nuclide density time evolution with burnup or decay. It includes eigenvalue calculations of critical and subcritical systems as well as neutron transport calculations in fixed source mode or k-code mode to obtain reaction rates and energy deposition that are necessary for burnup calculations.
The code integrates the code MCNP4C, which is used for neutron transport calculation, and a novel Transmutation Trajectory Analysis code (TTA), which serves for density evolution calculation, including formation and analysis of the transmutation chain. MCB is compatible with MCNP and preserves the structure of it. Complete burnup calculations can be done in a one single run and it requires preparation of only one input file by a modest modification of an MCNP input file. The code was extensively tested in benchmark calculations and reactor core designing. The general conclusion from practical application shows that MCB1C produces valuable results that are physically inherent and the correctness of physical model applied has been proved. MCB1C has been also equipped with new features among them the simulation of material processing including continuous feeding of materials is the most important. Development of the code was addressed towards improving calculation effectiveness and system diagnostic and towards improving physical model for rigid treatment but also providing simplified model option for quick design studies or benchmarks.
The MCB systems is build on MCNP4C foundation and requires the following executable files from MCNP processing systems:
PRPR: Pre-processor for Extracting the Various Hardware Versions of MCNP and other codes.
MAKXSF: Preparer of MCNP Cross-Section Libraries.
FSPLIT: Splits the fortran source into subroutines
mcnp4c.id: the main MCNP4C codef file (not patched by patchf files)
mcnpc.id: the c subroutines for MCNP4C codef file
These files are not included in this distribution and are provided with the MCNP-4C package (should be requested separately).
RELATED DATA LIBRARY
The MCB system consists of the following libraries:
The burnup library - BPLIB (included in this package)
JEF2.2 cross section library (request at NEA-1667/01)
JENDL3.2 cross section library (request at NEA-1670/04)
Transport libraries DLC200 (request at RSICC https://rsicc.ornl.gov/Default.aspx)
Dosimetry cross sections library EAF99 (request at NEA-1668/01)
Hard disk space of 15 GB temporarily available during installation of all cross section libraries makes the installation smooth and easy. Decompressed ASCII files of cross section libraries occupy most of the HD space. The binary libraries after installations take HD space as follows: ENDFB6.8 - 905 MB, JEF2.2 - 547 MB, JENDL3.2 - 1.3 GB, DLC200 - 173 MB, EAF99 - 222 MB, and all together about 3.2 GB. For most of the applications, the installation of one transport cross section library will be sufficient, but for particular needs the user can install even all of them.
J. Cetnar, W. Gudowski and J. Wallenius, "MONTE-CARLO CONTINUOUS ENERGY BURNUP (MCB1C) - THE CODE DESCRIPTION, METHODS AND BENCHMARKS" in preparation for NSE.
J. Cetnar, W. Gudowski and J. Wallenius, "MCB: A continuous energy Monte Carlo Burnup simulation code", In "Actinide and Fission Product Partitioning and Transmutation", EUR 18898 EN, OECD/NEA (1999) 523.
| Package ID | Computer language |
|---|---|
| NEA-1643/01 | FORTRAN-77, C-LANGUAGE |
Keywords: Monte Carlo method, burnup, criticality, flux distribution, neutronics, reaction rates, shielding.
