NEA-1643 MCB1C.
MCB1C, Monte-Carlo Continuous Energy Burnup Code
NAME OR DESIGNATION OF PROGRAM, COMPUTER, DESCRIPTION OF PROGRAM OR FUNCTION, METHODS, RESTRICTIONS, TYPICAL RUNNING TIME, FEATURES, RELATED OR AUXILIARY PROGRAMS, STATUS, REFERENCES, HARDWARE REQUIREMENTS, LANGUAGE, SOFTWARE REQUIREMENTS, OTHER RESTRICTIONS, NAME AND ESTABLISHMENT OF AUTHORS, MATERIAL, CATEGORIES
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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 |
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3. DESCRIPTION OF PROGRAM OR FUNCTION
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.
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.
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4. METHODS
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 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.
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8. RELATED OR AUXILIARY PROGRAMS
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 following libraries, which are distributed with the code package for the convenience of users:
- The burnup library - BPLIB (included in NEA-1643/01 package)
- ENDFB6.8 cross section library (NEA-1669/01 and NEA-1669/02 packages)
- JEF2.2 cross section library (NEA-1667/01 package)
- JENDL3.2 cross section library (NEA-1670/01, NEA-1670/02 and NEA-1670/03 packages)
- Transport libraries DLC200 (DLC-200/04 package)
- Dosimetry cross sections library EAF99 (NEA-1668/01 package)
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.
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 following libraries, which are distributed with the code package for the convenience of users:
- The burnup library - BPLIB (included in NEA-1643/01 package)
- ENDFB6.8 cross section library (NEA-1669/01 and NEA-1669/02 packages)
- JEF2.2 cross section library (NEA-1667/01 package)
- JENDL3.2 cross section library (NEA-1670/01, NEA-1670/02 and NEA-1670/03 packages)
- Transport libraries DLC200 (DLC-200/04 package)
- Dosimetry cross sections library EAF99 (NEA-1668/01 package)
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.
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10. REFERENCES
- 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.
Users are also kindly requested to check and update the references on MCB1C web-page http://www.neutron.kth.se/mcb
- 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.
Users are also kindly requested to check and update the references on MCB1C web-page http://www.neutron.kth.se/mcb
NEA-1643/01, included references:
- J. Cetnar, W. Gudowski, J. Wallenius and K. Tucek:Simulation of Nuclide Transmutations with Monte-Carlo Continuous
Energy Burnup Code (MCB1C)
- W. Gudowski et al.:
IAEA Benchmark on Accelerator-Driven Systems
- J. Cetnar, W. Gudowski, J. Wallenius:
User Manual for Monte-Carlo Continuous Energy Burnup (MCB) Code
Version 1C
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| Package ID | Computer language |
|---|---|
| NEA-1643/01 | FORTRAN-77, C-LANGUAGE |
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NEA-1643/01
mcb.tar.gz The MCB code (compressed tar file)bplib.tar.gz The burnup library - BPLIB (compressed tar file)
eaf.mcbxl.tar.gz xsdir for EAF (compressed tar file)
eaf99.A.tar.gz Dosimetry cross sections library EAF99
Install Installation script for MCB
InstEaf Installation script for EAF99 cross sections library
MakePatch Script for making MCB compilation script
Manual_MCB.pdf User manual (PDF format)
MCB1C_um.doc Usernal manual (Word format)
Readme Readme for MCB1C installation
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- C. Static Design Studies
- D. Depletion, Fuel Management, Cost Analysis, and Power Plant Economics
- J. Gamma Heating and Shield Design
Keywords: Monte Carlo method, burnup, criticality, flux distribution, neutronics, reaction rates, shielding.