VITAMIN-E: an ENDF/B-V multigroup cross-section library for LMFBR core and shield, LWR shield, dosimetry and fusion blanket technology
The Department of Energy (DOE) Office of Fusion Energy (OFE) and the Division of Reactor Research and Technology (DRRT) jointly sponsored the development of a coupled fine-group cross section library. This 171-neutron, 36-gamma-ray group library was based upon ENDF/B-IV and was intended to be applicable to fusion reactor neutronics and LMFBR core and shield analysis. Versions of the library are available from the Radiation Shielding Information Center (RSIC) at the Oak Ridge National Laborary in both AMPX and CCCC formats. Computer codes for energy group collapsing, interpolation on Bondarenko factors for resonance self-shielding and temperture corrections, and various other useful data manipulations are also available. The experience gained in the generation, validation and utilization of this library along with its broad range of applicability has led to the request for updating this data set using ENDF/B-V. Additional support in this regard has been provided by the Defense Nuclear Agency (DNA) and by the Electric Power Research Institute (EPRI) in support of weapons analyses and light water reactor shielding and dosimetry problems, respectively. The purpose of the report is to provide detailed specifications and rationale for the proposed ENDF/B-V update (designated VITAMIN-E) to the VITAMIN-C library.
- Research Organization:
- Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
- DOE Contract Number:
- W-7405-ENG-26
- OSTI ID:
- 6180002
- Report Number(s):
- ORNL-5505; ENDF-274; TRN: 79-010468
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
REACTOR TECHNOLOGY
NUCLEAR DATA COLLECTIONS
CROSS SECTIONS
DATA COMPILATION
GROUP CONSTANTS
LMFBR TYPE REACTORS
MULTIGROUP THEORY
NEUTRON DOSIMETRY
SHIELDING
WATER COOLED REACTORS
BREEDER REACTORS
DATA
DOSIMETRY
EPITHERMAL REACTORS
FAST REACTORS
FBR TYPE REACTORS
INFORMATION
LIQUID METAL COOLED REACTORS
NEUTRON TRANSPORT THEORY
NUMERICAL DATA
REACTORS
TRANSPORT THEORY
220100* - Nuclear Reactor Technology- Theory & Calculation