RSICC CODE PACKAGE CCC-

RSICC CODE PACKAGE CCC-719

1. NAME AND TITLE

MCB-1C: Monte-Carlo Continuous Energy Burnup Code System.

Installation of MCB-1C requires MCNP4C source files and associated PRPR build system, which are not included in this package .The MCNP4C package has been obsolete for several years. These codes have not been updated for use on current compilers and will certainly require modifications to build. Neither LANL nor RSICC support MCNP4C in any fashion. MCB-1C is being made available through RSICC at the request of several who people have recently expressed interest in it.

DATA LIBRARIES:

The MCB system consists of following libraries, which are distributed with the code package for the convenience of users. Abstracts from the NEADB website are appended to this MCB-1C abstract.

- 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, /02 and /03 packages)

- Transport libraries DLC200 (DLC-200/04 package)

- Dosimetry cross sections library EAF99 (NEA-1668/01 package)

2. CONTRIBUTORS

Department of Nuclear and Reactor Physics, Royal Institute of Technology, Stockholm, Sweden, through the Nuclear Energy Agency Data Bank, Issy-les-Moulineaux, France, where MCB-1C is distributed with package identifier: NEA-1643/01.

3. CODING LANGUAGE AND COMPUTER

Linux-based PC or Unix workstations; Fortran 77 and ANSI C (C00719MNYCPWS).

4. NATURE OF PROBLEM SOLVED

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 MCB-1C patch file and data packages as distributed by the NEADB are very well organized and are being made available through RSICC as received. The RSICC package includes the MCB-1C patch and MCB data libraries. Installation of MCB requires MCNP4C source code and utility programs, which are not included in this MCB distribution. They were provided with the now obsolete CCC-700/MCNP-4C package.

5. METHOD OF SOLUTION

The code integrates the code MCNP4C, which is used for neutron transport calculations, 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.

6. RESTRICTIONS OR LIMITATIONS

Installation of MCB-1C requires MCNP4C source files and associated PRPR build system. The MCNP4C package has been obsolete for several years. These codes have not been updated for use on current compilers and will certainly require modifications to build. Neither LANL nor RSICC support MCNP4C in any fashion. MCB-1C is being made available through RSICC at the request of several who people have recently expressed interest in it.

7. TYPICAL RUNNING TIME

In 2002, 2.5 hours were required for complete installation of the package (including compilation/link of the MCB-1C code and the ASCII to binary conversion of all the nuclear data libraries). The 4 test cases ran in approx. 10 minutes on a PC DELL PowerEdge 1550 Bi 1GHz.

8. COMPUTER HARDWARE REQUIREMENTS

MCB-1C was developed for Linux-based PCs and SGI Unix workstations. 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.

9. COMPUTER SOFTWARE REQUIREMENTS

The MCB install procedure provides options for Linux PC using either g77 or Absoft compilers and for SGI f77 and cc compilers. The MCB system is built on MCNP4C foundation and requires the following executable files and source code from the MCNP4c code pacakge:

PRPR: Pre-processor for Extracting Various Hardware Versions of MCNP.

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

The MCNP4C source code and utility programs are not included in the MCB distribution. These items were provided with the now obsolete CCC-700/MCNP-4C package.

10. REFERENCES

10.a included in package:

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.”

10.b background information

J. Cetnar, W. Gudowski and J. Wallenius, "Monte Carlo Continuous Energy Burnup (MCB 1C) - 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/research/MCB/

11. CONTENTS OF CODE PACKAGE

MCB-1C is transmitted on DVD which includes the references listed in 10.a above and compressed Unix tar files which contain the MCB 1C patch, an install script and data libraries. Note that the MCNP4C source code and utility programs are not included in this distribution.

12. DATE OF ABSTRACT

August 2007.

KEYWORDS: BURNUP; CRITICALITY CALCULATIONS; MONTE CARLO; NEUTRON

This abstract was copied from the NEADB website to serve as documentation for ZZ-MCB-ENDF6.8 data as distributed with MCB-1C:

http://www.nea.fr/abs/html/nea-1669.html

NEA-1669 ZZ-MCB-ENDF/B-6.8. (Abstract last modified 20-DEC-2004)

ZZ MCB-ENDF/B6.8, MCB Continuous-Energy Neutron Cross Section Libraries for Temperatures from 300 to 1800 K.

1. NAME OR DESIGNATION OF PROGRAM - ZZ-MCB-ENDF/B6.8.

2. COMPUTER FOR WHICH PROGRAM IS DESIGNED AND OTHER MACHINE VERSION PACKAGES AVAILABLE -

Program-name Package-ID Status

ZZ-MCB-ENDF/B-6.8 NEA-1669/01 Tested

ZZ-MCB-ENDF/B-6.8 NEA-1669/02 Tested

ZZ-MCB-ENDF/B-6.8 NEA-1669/03 Tested

Machines used:

Package-ID Orig.Computer Test Computer

NEA-1669/01 Linux-based PC,UNIX gen. W.S. Linux-based PC

NEA-1669/02 Linux-based PC,UNIX gen. W.S. Linux-based PC

NEA-1669/03 Linux-based PC,UNIX gen. W.S. Linux-based PC

3. DESCRIPTION OF PROGRAM OR FUNCTION -
MCB-ENDF/B6.8 is a continuous-energy cross section libraries in ACE format suitable for the MCB-1C and MCNP codes. Libraries for various materials were generated at six different temperatures, and cover
the energy range up to 150 MeV.
FORMAT: ACE
NUMBER OF GROUPS: Continuous energy
NUCLIDES: Library includes elements with the following atomic numbers:
Z= 1-9, 11-17, 19-29, 31-68, 71-75, 77, 79, 82-83, 90-99, (i.e.: H, He, Li, Be, B, C, N, O, F, Na, Mg, Al, Si, P, S, Cl, K, Ca, Sc, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Ga, Ge, As, Se, Br, Kr, Rb, Sr, Y, Zr, Nb, Mo, Tc, Ru, Rh, Pd, Pd, Ag, Cd, In, Sn, Sb, Te, I, Xe, Cs, Ba, La, Ce, Pr, Nd, Pm, Sm, Eu, Gd, Tb, Dy, Ho, Er, Lu, Hf, Ta, W, Re, Ir, Au, Pb, Bi, Th, Pa, U, Np, Pu, Am, Cm, Bk, Cf, Es).
TEMPERATURES: 300 K, 600 K, 900 K, 1200 K, 1500 K, and 1800 K.
ORIGIN: ENDF/B-6.8
WEIGHTING SPECTRUM: --

NEA-1669/03:

DVD version, contains NEA-1669/01 and NEA-1669/02

4. METHODS - The library was generated using the NJOY processing code. An example of the input data is provided.

5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM -

6. TYPICAL RUNNING TIME - The ASCII-to-binary conversion can generally be accomplished within a few seconds, even for the large data files.

7. UNUSUAL FEATURES -

8. RELATED OR AUXILIARY PROGRAMS - MAKSF, which is distributed with CCC-700/MCNP4C, can be used to convert MCNP Type 1 data to Type 2 (binary). For use with the code MCB-1C the MAKXSF installation must be slightly modified.

9. STATUS
NEA-1669/01: 30-AUG-2002 Tested at NEADB
NEA-1669/02: 30-AUG-2002 Tested at NEADB
NEA-1669/03: 20-DEC-2004 Tested at NEADB

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.
- J. F. Briesmeister, Ed.
"MCNP - A General Monte Carlo N-Particle Transport Code, Version 4C," LA-13709-M (April 2000).
NEA-1669/01:
NEA-1669/02:
NEA-1669/03:

11. HARDWARE REQUIREMENTS - ASCII card images; Unix workstation. The library distribution was prepared on a LINUX system. On a Windows-based PC, one may use utilities such as WinZip(R) to unzip and extract the library files.

12. PROGRAMMING LANGUAGE -NEA-1669/01:
NEA-1669/02:
NEA-1669/03:

13. SOFTWARE REQUIREMENTS -

14. OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS -

15. NAME AND ESTABLISHMENT OF AUTHORS -
       Department of Nuclear and Reactor Physics
       Royal Institute of Technology
       100 44 Stockholm
       Sweden

16. MATERIAL AVAILABLE -NEA-1669/01:
InstEndf   Installs ENDF/B-6.8 cross section library
endf.mcbxl.tar.gz    ENDF/B-6.8 xsdir
endfb6.8.A1.tar.gz    Transport cross section library based on ENDF/B-6.8
Readme.Endfb6.8  ENDF/B-6.8 cross section library installation dedicated for MCB
NEA-1669/02:
endfb6.8.A2.tar.gz    Transport cross section library based on ENDF/B-6.8
NEA-1669/03:
Readme.Endfb6.8    Original information file
InstEndf-l  Installation procedure for independent library installation
(outside MCB system)
ENDFB6.8 cross section data library (two files per temperature)
XSDIR files XSDIR file for LINUX and UNIX operating systems

17. CATEGORIES -

- Z. Data.

Keywords: CROSS SECTIONS, DATA LIBRARY, FISSION PRODUCTS, MONTE CARLO METHOD, NEUTRONS, REACTION

This abstract was copied from the NEADB website to serve as documentation for ZZ-MCB-JEF2.2 as distributed with MCB-1C:

http://www.nea.fr/abs/html/nea-1667.html

NEA-1667 ZZ-MCB-JEF2.2. (Abstract last modified 30-AUG-2002)

ZZ MCB-JEF2.2, MCB Continuous-Energy Neutron Cross Section Libraries for Temperatures from 300 to 1800 K.

1. NAME OR DESIGNATION OF PROGRAM - ZZ-MCB-JEF2.2.

2. COMPUTER FOR WHICH PROGRAM IS DESIGNED AND OTHER MACHINE VERSION PACKAGES AVAILABLE -

Program-name Package-ID Status

ZZ-MCB-JEF2.2 NEA-1667/01 Tested

Machines used:

Package-ID Orig.Computer Test Computer

NEA-1667/01 Linux-based PC,UNIX gen. W.S. Linux-based PC

3. DESCRIPTION OF PROGRAM OR FUNCTION -
MCB-JEF2.2 is a continuous-energy cross section libraries in ACE format suitable for the MCB-1C and MCNP codes. Libraries for various materials were generated at six different temperatures, and cover the energy range up to 20 MeV.
FORMAT: ACE
NUMBER OF GROUPS: Continuous energy
NUCLIDES: H1, H2, H3, He3, He4, Li6, Li7, Be9, B10, B11, C-nat., N14, N15, O16, O17, Na23, F19, Mg-nat., Al27, Si-nat., P31, S32, S33, S34, S36, Cl-nat, K-nat, Ca-nat., Ti-nat, V-nat, Cr50, Cr52, Cr53, Cr54, Mn55, Fe54, Fe56, Fe57, Fe58, Co59, Ni58, Ni58, Ni60, Ni61, Ni62, Ni64, Cu-nat, Ga-nat, Ge72, Ge73, Ge74, Ge76, As75, Se74, Se76, Se77, Se78, Se80, Se82, Br79, Br81, Kr78, Kr80, Kr82, Kr83, Kr84, Kr85, Kr86, Rb85, Rb86, Rb87, Sr84, Sr86, Sr87, Sr88, Sr89, Sr90, Y89, Y90, Y91, Zr-nat, Zr90, Zr91, Zr92, Zr93, Zr94, Zr95, Zr96, Nb93, Nb94, Nb95, Mo-nat, Mo92, Mo94, Mo95, Mo96, Mo97, Mo98, Mo99, Mo100, Tc99, Ru96, Ru98, Ru99, Ru100, Ru101, Ru102, Ru103, Ru104, Ru105, Ru106, Rh103, Rh105, Pd102, Pd104, Pd105, Pd106, Pd107, Pd108, Pd110, Ag107, Ag109, Ag111, Cd-nat., Cd106, Cd110, Cd111, Cd112, Cd113, Cd114, Cd115, Cd116, In113, In115, Sn114, Sn115, Sn116, Sn117, Sn118, Sn119, Sn120, Sn122, Sn123, Sn124, Sn125, Sn126, Sb121, Sb123, Sb124, Sb125, Sb126, Te120, Te122, Te123, Te124, Te125, Te126, Te127, Te128, Te129, Te130, Te132, I127, I129, I130, I131, I135, Xe124, Xe126, Xe128, Xe129, Xe130, Xe131, Xe132, Xe133, Xe134, Xe135, Xe136, Cs133, Cs134, Cs135, Cs136, Cs137, Ba134, Ba135, Ba136, Ba137, Ba138, Ba140, La139, La140, Ce140, Ce141, Ce142, Ce143, Ce144, Pr141, Pr142, Pr143, Nd142, Nd143, Nd144, Nd145, Nd146, Nd147, Nd148, Nd150, Pm147, Pm148, Pm149, Pm151, Sm144, Sm147, Sm148, Sm149, Sm150, Sm151, Sm152, Sm153, Sm154, Eu151, Eu152, Eu153, Eu154, Eu155, Eu156, Eu157, Gd154, Gd155, Gd156, Gd157, Gd158, Gd160, Tb159, Tb160, Dy160, Dy161, Dy162, Dy163, Dy164, Ho165, Er166, Er167, Lu175, Lu176, Hf174, Hf176, Hf177, Hf178, Hf179, Hf180, Ta181, Ta182, W182, W183, W184, W186, Re185, Re187, Au197, Pb-nat., Bi209, Th230, Th232, Pa231, Pa233, U232, U233, U234, U235, U236, U237, U238, Np237, Np238, Np239, Pu236, Pu237, Pu238, Pu239, Pu240, Pu241, Pu242, Pu243, Pu244, Am241, Am242, Am242m, Am243, Cm241, Cm242, Cm243, Cm244, Cm245, Cm246, Cm247, Cm248, Bk249, Cf249, Cf250,
Cf251, Cf252, Cf253, Es253
TEMPERATURES: 300 K, 600 K, 900 K, 1200 K, 1500 K, and 1800 K.
ORIGIN: JEF-2.2
WEIGHTING SPECTRUM: --

4. METHODS - The library was generated using the NJOY processing code. An example of the input data is provided.

5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM -

6. TYPICAL RUNNING TIME - The ASCII-to-binary conversion can generally be accomplished within a few seconds, even for the large data files.

7. UNUSUAL FEATURES -

8. RELATED OR AUXILIARY PROGRAMS - MAKSF, which is distributed with
CCC-700/MCNP4C, can be used to convert MCNP Type 1 data to Type 2 (binary). For use with the code MCB-1C the MAKXSF installation must be slightly modified.

9. STATUS
NEA-1667/01: 30-AUG-2002 Tested at NEADB

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.
- J. F. Briesmeister, Ed. "MCNP - A General Monte Carlo N-Particle Transport Code, Version 4C," LA-13709-M (April 2000).
NEA-1667/01:

11. HARDWARE REQUIREMENTS - ASCII card images; Unix workstation.
The library distribution was prepared on a LINUX system. On a Windows-based PC, one may use utilities such as WinZip(R) to unzip and extract the library files.

12. PROGRAMMING LANGUAGE -NEA-1667/01:

13. SOFTWARE REQUIREMENTS -

14. OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS -

15. NAME AND ESTABLISHMENT OF AUTHORS -
   Department of Nuclear and Reactor Physics
   Royal Institute of Technology
   100 44 Stockholm
   Sweden

16. MATERIAL AVAILABLE -NEA-1667/01:
InstJef   Installs JEF2.2 cross section library
jef.mcbxl.tar.gz    JEF2.2 xsdir
jef2.2.A.tar.gz    Transport cross section library based on JEF2.2
Readme.Jef2.2   JEF2.2 cross section library installation dedicated for MCB

17. CATEGORIES -

- Z. Data.

Keywords: CROSS SECTIONS, DATA LIBRARY, FISSION PRODUCTS, MONTE CARLO METHOD, NEUTRONS, REACTION

This abstract was copied from the NEADB website to serve as documentation for ZZ-MCB-JENDL3.2 data as distributed with MCB-1C:

http://www.nea.fr/abs/html/nea-1670.html

NEA-1670 ZZ-MCB-JENDL-3.2. (Abstract last modified 20-DEC-2004)

ZZ MCB-JENDL-3.2, MCB Continuous-Energy Neutron Cross Section Libraries for Temperatures from 300 to 1800 K.

1. NAME OR DESIGNATION OF PROGRAM - ZZ-MCB-JENDL-3.2.

2. COMPUTER FOR WHICH PROGRAM IS DESIGNED AND OTHER MACHINE VERSION PACKAGES AVAILABLE -

Program-name Package-ID Status

ZZ-MCB-JENDL-3.2 NEA-1670/01 Tested

ZZ-MCB-JENDL-3.2 NEA-1670/02 Tested

ZZ-MCB-JENDL-3.2 NEA-1670/03 Tested

ZZ-MCB-JENDL-3.2 NEA-1670/04 Tested

Machines used:

Package-ID Orig.Computer Test Computer

NEA-1670/01 Linux-based PC,UNIX gen. W.S. Linux-based PC

NEA-1670/02 Linux-based PC,UNIX gen. W.S. Linux-based PC

NEA-1670/03 Linux-based PC,UNIX gen. W.S. Linux-based PC

NEA-1670/04 Linux-based PC,UNIX gen. W.S. Linux-based PC

3. DESCRIPTION OF PROGRAM OR FUNCTION -
MCB-JENDL-3.2 is a continuous-energy cross section libraries in ACE format suitable for the MCB-1C and MCNP codes. Libraries for various materials were generated at six different temperatures, and cover the energy range up to 20 MeV.
FORMAT: ACE
NUMBER OF GROUPS: Continuous energy
NUCLIDES: Library includes elements with the following atomic numbers:
Z= 1-9, 11-29, 31-65, 72-74, 82-83, 88-100 (i.e.: H, He, Li, Be, B, C, N, O, F, Na, Mg, Al, Si, P, S, Cl, Ar, K, Ca, Sc, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Ga, Ge, As, Se, Br, Kr, Rb, Sr, Y, Zr, Nb, Mo, Tc, Ru, Rh, Pd, Pd, Ag, Cd, In, Sn, Sb, Te, I, Xe, Cs, Ba, La, Ce, Pr, Nd, Pm, Sm, Eu, Gd, Tb, Hf, Ta, W, Pb, Bi, Ra, Ac, Th, Pa, U, Np, Pu, Am, Cm, Bk, Cf, Es, Fm).
TEMPERATURES: 300 K, 600 K, 900 K, 1200 K, 1500 K, and 1800 K.
ORIGIN: JENDL-3.2
WEIGHTING SPECTRUM: --

NEA-1670/04:

Dec 2004: DVD version, contains NEA-1670/01, NEA-1670/02 and NEA-1670/03

4. METHODS - The library was generated using the NJOY processing code. An example of the input data is provided.

5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM -

6. TYPICAL RUNNING TIME - The ASCII-to-binary conversion can generally be accomplished within a few seconds, even for the large data files.

7. UNUSUAL FEATURES -

9. STATUS
NEA-1670/01: 30-AUG-2002 Tested at NEADB
NEA-1670/02: 30-AUG-2002 Tested at NEADB
NEA-1670/03: 30-AUG-2002 Tested at NEADB
NEA-1670/04: 20-DEC-2004 Tested at NEADB

11. HARDWARE REQUIREMENTS - ASCII card images; Unix workstation.
The library distribution was prepared on a LINUX system. On a windows-based PC, one may use utilities such as WinZip(R) to unzip and extract the library files.

12. PROGRAMMING LANGUAGE -NEA-1670/01:
NEA-1670/02:
NEA-1670/03:
NEA-1670/04:

13. SOFTWARE REQUIREMENTS -

14. OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS -

15. NAME AND ESTABLISHMENT OF AUTHORS -
  Department of Nuclear and Reactor Physics
  Royal Institute of Technology
  100 44 Stockholm
  Sweden

16. MATERIAL AVAILABLE -NEA-1670/01:
Inst.Jendl   Installs JENDL-3.2 cross section library
jendl.mcbxl.tar.gz    JENDL-3.2 xsdir
jendl3.2.A1.tar.gz    Transport cross section library based on JENDL3.2
Readme.Jendl3.2   JENDL3.2 cross section library installation dedicated for MCB
NEA-1670/02:
jendl3.2.A2.tar.gz    Transport cross section library based on JENDL3.2
NEA-1670/03:
jendl3.2.A3.tar.gz    Transport cross section library based on JENDL3.2
NEA-1670/04:
Readme.Jendl3.2     Original information file
InstJendl-l    Installation procedure for independent library installation
(outside MCB system)
JENDL3.2 cross section data library, two files per temperature
XSDIR files for LINUX and operating systems

17. CATEGORIES -

- Z. Data.

Keywords: CROSS SECTIONS, DATA LIBRARY, FISSION PRODUCTS, MONTE CARLO METHOD, NEUTRONS, REACTION

This abstract was copied from the NEADB website to serve as documentation for ZZ-MCB-DLC200 as distributed with MCB-1C:

http://www.nea.fr/abs/html/dlc-0200.html

DLC-0200 ZZ-MCNPDATA. (Abstract last modified 30-AUG-2002)

ZZ MCNPDATA, Standard Neutron, Photon and Electron Data Libraries for MCNP-4C and MCB1C

1. NAME - ZZ-MCNPDATA.

DLC-0200/04:

NAME - ZZ-MCB-DLC200.

2. COMPUTER FOR WHICH PROGRAM IS DESIGNED AND OTHER MACHINE VERSION PACKAGES AVAILABLE -

Program-name Package-ID Status

ZZ-MCNPDATA DLC-0200/01 Obsolete

ZZ-MCNPDATA DLC-0200/02 Obsolete

ZZ-MCNPDATA DLC-0200/03 Arrived

ZZ-MCB-DLC200 DLC-0200/04 Tested

Machines used:

Package-ID Orig.Computer Test Computer

DLC-0200/03 Many Computers

DLC-0200/04 Linux-based PC,UNIX gen. W.S. Linux-based PC

3. DESCRIPTION - These cross-section libraries are released by the Diagnostics Applications Group, X-5, at Los Alamos National Laboratory for use with the MCNP Monte Carlo code package. This release includes all of the X-5 distributed neutron data libraries, the photon libraries MCPLIB1 and MCPLIB02, the electron libraries EL1 and EL03, an updated XSDIR file, and information files Readme.txt and Readme_endf60.txt. This release is intended to completely replace previous RSICC releases DLC-105, DLC-181, and DLC-189 as well as the cross sections previously included with CCC-200/MCNP4A, and will be updated as new libraries become available.
The README file provides information regarding each data library of this release. Additional documentation for some of the individual libraries and example SPECS files for use with MAKXSF are also provided. The XSDIR file is specific to this release and may not work with previous packages. Currently the neutron data library ENDF60 (based on ENDF/B-VI, up through and including release 2) is the default library for continuous-energy neutron transport. Additionally, the libraries MCPLIB02 and EL03 are the default libraries for photon and electron transport respectively. More information on the data libraries contained in this release is available in Appendix G of the MCNP4C manual.

DLC-0200/04:

DESCRIPTION OF PROGRAM OR FUNCTION - ZZ-MCB-DLC200 contains the same cross section tables as the DLC-0200/03 package for the MCNP-4C code, except that the installation procedures are adapted to the MCB1C code system (NEA 1643/01).

4. APPLICATION OF THE DATA - DLC-200/MCNPDATA is for use with Version 4C and later of the MCNP transport code. This data library provides a comprehensive set of cross sections for a wide range of radiation transport applications using the Monte Carlo code package CCC-700/MCNP4C. See Appendix G of the MCNP report LA-13709-M for information on the libraries and how to select specific nuclides for use in MCNP.
SOURCE AND SCOPE OF DATA - A wide variety of continuous-energy, discrete, multigroup, thermal and dosimetry neutron data libraries are available in this release. The continuous-energy neutron data libraries available include: ENDF60, RMCCS, RMCCSA, ENDF5U, ENDF5P, NEWXS, ENDF5MT*, MISC5XS**, ENDL85, KIDMAN, 100XS, URES***, ENDF6DN***, ENDF62MT***, and ENDL92***. The discrete neutron data libraries include: NEWXSD, DRMCCS, and DRE5. The multigroup neutron data library is MGXSNP, and the thermal S(alpha, beta) libraries are TMCCS and THERXS. The neutron dosimetry libraries are 531DOS, 532DOS, and LLLDOS. The photon transport libraries are MCPLIB and MCPLIB02, and the electron libraries are EL and EL03***. The photon and electron data libraries contain data for elements having Z<95. The data libraries, as distributed, are in ASCII, or type 1, format.
* The data library ENDF5MT contains data previously available in the library EPRIXS, along with the U600K data library.
** The data library MISC5XS contains corrected data for ENDF/B-V based Zr as described below, and the libraries previously known as IRNAT, MISCXS, ARKRC, TM169, GDT2GP, and T2DDC. The ENDF/B-V Zr data has been corrected for five ZAID's from the libraries RMCCS, DRMCCS, ENDF5P, DRE5, and EPRIXS. Below is a summary of the changes that have been made for Zr:
       (Previous)  (Corrected)
RMCCS   40000.51c  40000.57c MISC5XS 300K
DRMCCS  40000.51d  40000.57d MISC5XS 300K
ENDF5P  40000.50c  40000.56c MISC5XS 300K
DRE5    40000.50d  40000.56d MISC5XS 300K
EPRIXS  40000.53c  40000.58c MISC5XS 600K
*** These five data libraries (URES, ENDF6DN, ENDF62MT, ENDL92, and EL03) are newly released in DLC-200. In the February 2001 update, URES was replaced with URESA. Please consult the README file and the more detailed documentation provided for descriptions of these libraries.

5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM -

6. TYPICAL RUNNING TIME - Run times vary among computers.

7. UNUSUAL FEATURES -

8. NAME AND TITLE OF DATA RETRIEVAL PROGRAMS -
The two files, PRPR and MAKXSF, are to be used only if the user wants to translate the data into binary form. They are distributed with the MCNP4C code package. This translation is recommended for saving disk space and cutting down on the running time of MCNP. Instructions for using PRPR and the source file for MAKXSF to produce the MAKXSF executable are included with the MCNP4C package. Input to the executable MAKXSF includes the SPECS file, the XSDIR file (nuclide data file directory) and the data libraries. The user should study the SPECS and XSDIR files and become familiar with the instructions for MAKXSF in the MCNP4C manual. Some platforms have a restriction on the length of filenames, and the SPECS file may need to be appropriately edited.

9. STATUS
DLC-0200/01: 02-OCT-2000 Obsolete
DLC-0200/02: 06-JUL-2001 Obsolete
DLC-0200/03: 25-APR-2002 Masterfiled Arrived
DLC-0200/04: 30-AUG-2002 Tested at NEADB

10. REFERENCES -
A. A. Adams, S. C. Frankle, and R. C. Little, "Revised Criticality Benchmarks for MCNP," Los Alamos National Laboratory memorandum (1996).
J. D. Court, R. C. Brockhoff, and J. S. Hendricks, "Lawrence Livermore Pulsed Sphere Benchmark Analysis of MCNP ENDF/B-VI," Los Alamos National Laboratory report LA-12885 (1994).
J. D. Court, J. S. Hendricks, and S. C. Frankle, "MCNP ENDF/B-VI Validation: Infinite Media Comparisons of ENDF/B-VI and ENDF/B-V," and errata, Los Alamos National Laboratory report LA-12887 (1994).
J. D. Court and J. S. Hendricks, "Benchmark Analysis of MCNP ENDF/B-VI Iron," Los Alamos National Laboratory report LA-12884 (1994).
S. C. Frankle, "Photon Production Assessment for the MCNP ENDF/B-VI Data Library," Los Alamos National Laboratory report LA-13093 (1995).
S. C. Frankle, "Cross-section and Reaction Nomenclature for MCNP Continuous- energy Libraries and DANTSYS Multigroup Libraries," Los Alamos National Laboratory memorandum XTM:SCF-96-313 (1996).
S. C. Frankle, "Experience with ENDF60 and QA of a Continuous-energy Library," Los Alamos National Laboratory memorandum XTM:SCF-96-363 (1996).
DLC-0200/03:
- S.C. Frankle and R.C. Little:
README, Informal overview of the DLC-200 MCNPDATA package (Apr.2000)
- S.C. Frankle:
README_ENDF60, informal overview of the ENDF60 neutron data library
(April 2000)
- J.S. Hendricks, S.C. Frankle and J.D. Court:
ENDF/B-VI data for MCNP, and Errata,
Los Alamos Nat. Lab. report LA-12891 (1994)
- H. G. Hughes:
Information on the MCPLIB02 photon library
Los Alamos Nat. Lab. memorandum X-6:HGH-93-77 (revised 1996)
- R.C. Little:
Summary documentation for the 100XS neutron cross-section library
Los Alamos Nat. Lab. memorandum XTM:95-259 and LA-UR-96-24 (1995)
- R.C. Little and R.E. Seamon:
Dosimetry/Activation cross sections for MCNP
Los Alamos National Laboratory memorandum (March 13, 1984)
- R.C .Little:
Neutron and photon multigroup data tables for MCNP3B
Los Alamos Nat. Lab. memorandum X-6:RCL-87-225 (1987)
- R.C. Little and R.E. MacFarlane:
ENDF/B-VI neutron library for MCNP with probability tables
Los Alamos Nat. Lab. research note XCI-RN(U)98-041,
LA-UR-98-5718 (December 1998)
- S. Frankle :
ENDF62MT, a multi-temperature neutron library for MCNP (Rev.0)
Los Alamos Nat. Lab. memorandum XTM:96-153 (April 1996)
- C. Werner:
New data library for MCNP delayed neutron capability
Los Alamos Nat. Lab. memorandum XCI:CJW-99-25 (April 1999)
- K.J. Adams:
Electron Upgraede for MCNP4B
X-5-RN(U)-00-14 (May 25, 2000
- R.C. Little:
ReadURES.txt (February 2001)
DLC-0200/04:

11. DATA FORMAT AND COMPUTER - All data libraries are distributed in compressed mode. Files in the Unix tar file are in ASCII format and can be used with MCNP4C on all computer platforms supported by the code. Note that the trailing "1" was dropped from file names for Type 1 (ASCII) libraries. On the distribution CD, a self-extracting compressed file for Windows users contains both the MCNP4C code system with executables and cross-sections from MCNPDATA in binary format for PC users.

12. PROGRAMMING LANGUAGE -DLC-0200/03:
DLC-0200/04:

13. SOFTWARE REQUIREMENTS -

14.OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS -

15. NAME AND ESTABLISHMENT OF AUTHORS -
Contributed by:
                Radiation Safety Information Computational Center
                Oak Ridge National Laboratory
                Oak Ridge, Tennessee, U. S. A.
Developed by:   Los Alamos National Laboratory,
                Los Alamos, New Mexico, U.S.A.

DLC-0200/04:

NAME AND ESTABLISHMENT OF AUTHORS -
Contributed by:
       Department of Nuclear and Reactor Physics
       Royal Institute of Technology
       100 44 Stockholm
       Sweden

16. MATERIAL AVAILABLE -DLC-0200/03:
D200.PDF MCNPDATA manual
Readme.1st General description of CD and installation written by RSICC
Readme4c2.txt Compilation & executation info written by LANL
D200DOS2.exe ASCII MCNPDATA cross sections
D200TAR2.GZ ASCII MCNPDATA
701allcp.00 Content of package
Readme.rsi RSICC readme
DLC-0200/04:
Readme.Dlc200 Information file
InstDlc-l Procedure for independent library installation (outside MCB)
dlc.mcbxl.tar.gz Cross section data libraries
dlc200.A.tar.gz Cross section data libraries

17. CATEGORIES  -

- J.  Gamma Heating and Shield Design
- Z.  Data.

Keywords: GROUP CONSTANTS, LIBRARIES, MULTIGROUP, NEUTRON CROSS SECTIONS, PHOTON TRANSPORT

This abstract was copied from the NEADB website to serve as documentation for ZZ-MCB-EAF99 data as distributed with MCB-1C:

http://www.nea.fr/abs/html/nea-1668.html

NEA-1668 ZZ-MCB-EAF99. (Abstract last modified 30-AUG-2002)

ZZ MCB-EAF99, MCB Continuous-Energy Neutron Cross Section Libraries for Temperatures from 300 to 1800 K.

1. NAME OR DESIGNATION OF PROGRAM - ZZ-MCB-EAF99.

2. COMPUTER FOR WHICH PROGRAM IS DESIGNED AND OTHER MACHINE VERSION PACKAGES AVAILABLE -

Program-name Package-ID Status

ZZ-MCB-EAF99 NEA-1668/01 Tested

Machines used:

Package-ID Orig.Computer Test Computer

NEA-1668/01 Linux-based PC,UNIX gen. W.S. Linux-based PC

3. DESCRIPTION OF PROGRAM OR FUNCTION -
MCB-EAF99 is a continuous-energy activation cross section library in ACE format suitable for the MCB-1C and MCNP codes. Libraries for various materials were generated at six different temperatures, and cover the energy range up to 20 MeV.
FORMAT: ACE
NUMBER OF GROUPS: Continuous energy
NUCLIDES: Library includes elements with atomic numbers Z= 1 - 84, 86, 88 - 100 (i.e.: H, He, Li, Be, B, C, N, O, F, Ne, Na, Mg, Al, Si, P, S, Cl, Ar, K, Ca, Sc, Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zn, Ga, Ge, As, Se, Br, Kr, Rb, Sr, Y, Zr, Nb, Mo, Tc, Ru, Rh, Pd, Pd, Ag, Cd, In, Sn, Sb, Te, I, Xe, Cs, Ba, La, Ce, Pr, Nd, Pm, Sm, Eu, Gd, Tb, Dy, Ho, Er, Tm, Yb, Lu, Hf, Ta, W, Re, Os, Ir, Pt, Au, Hg, Tl, Pb, Bi, Po, Rn, Ra, Th, Pa, U, Np, Pu, Am, Cm, Bk, Cf, Es, Fm).
TEMPERATURES: 300 K, 600 K, 900 K, 1200 K, 1500 K, and 1800 K.
ORIGIN: EAF-99
WEIGHTING SPECTRUM: --

4. METHODS - The library was generated using the NJOY processing code. An example of the input data is provided.

5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM -

6. TYPICAL RUNNING TIME - The ASCII-to-binary conversion can generally be accomplished within a few seconds, even for the large data files.

7. UNUSUAL FEATURES -

9. STATUS
NEA-1668/01: 30-AUG-2002 Tested at NEADB

12. PROGRAMMING LANGUAGE -NEA-1668/01:

13. SOFTWARE REQUIREMENTS -

14. OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS -

15. NAME AND ESTABLISHMENT OF AUTHORS -
       Department of Nuclear and Reactor Physics
       Royal Institute of Technology
       100 44 Stockholm
       Sweden

16. MATERIAL AVAILABLE -NEA-1668/01:
eaf.mcbxl.tar.gz    xsdir for EAF (compressed tar file)
eaf99.A.tar.gz    Dosimetry cross sections library EAF99
InstEaf    Installation script for EAF99 cross sections library

17. CATEGORIES -

- Z. Data.

Keywords: CROSS SECTIONS, DATA LIBRARY, DOSIMETRY, FISSION PRODUCTS, MONTE CARLO METHOD, NEUTRONS, REACTION