Actions for Sensitivity study of neutron transport through standard and rebar concrete [electronic resource].
Sensitivity study of neutron transport through standard and rebar concrete [electronic resource].
- Published
- Oak Ridge, Tenn. : Oak Ridge National Laboratory, 1982.
Oak Ridge, Tenn. : Distributed by the Office of Scientific and Technical Information, U.S. Dept. of Energy. - Physical Description
- Pages: 5 : digital, PDF file
- Additional Creators
- Oak Ridge National Laboratory and United States. Department of Energy. Office of Scientific and Technical Information
Access Online
- Restrictions on Access
- Free-to-read Unrestricted online access
- Summary
- An investigation is under way at ORNL to (1) develop a data base pertinent to the transport of neutrons through thick concrete shields, (2) use the data base in an energy group boundary selection and collapsing scheme, and (3) develop a simple methodology to access the data base to provide rapid solutions to practical shielding problems. This paper describes work carried out to fulfill objective (1), the work consisting of calculations of the transport of fission neutrons through 1- and 2-m-thick slabs of standard concrete and rebar (steel-reinforced) concrete, together with calculations of the sensitivities of the results to total, absorption, and elastic cross sections. The transport calculations were performed with the one-dimensional discrete ordinates code ANISN in both forward and adjoint modes. The DLC-41C/VITAMIN-C cross-section library (171 neutron, 36 gamma groups) was employed, with a P/sub 3/ cross-section expansion and an S/sub 16/ angular quadrature. In all cases the fission source was assumed to be distributed within the first 1-cm thickness of the slab and the detector was assumed to occupy the last 1-cm thickness of the slab. For the rebar concrete the slab constituents were homogenized, with the horizontal and vertical No. 11 reinforcing steel rods comprising 7.6 vol. % of the slab. The quantity calculated was the absorbed dose rate, and care was taken in the mesh interval selection and source description to ensure agreement between the forward and adjoint results to within 0.02%.
- Report Numbers
- E 1.99:conf-820609-43
conf-820609-43 - Subject(s)
- Other Subject(s)
- Concretes
- Neutron Transport
- Fission Neutrons
- Transmission
- Reinforced Concrete
- Absorption
- Computer Calculations
- Cross Sections
- Discrete Ordinate Method
- Dose Rates
- Elastic Scattering
- Sensitivity Analysis
- Shielding
- Total Cross Sections
- Baryons
- Building Materials
- Elementary Particles
- Fermions
- Hadrons
- Materials
- Neutral-Particle Transport
- Neutrons
- Nucleons
- Radiation Transport
- Reinforced Materials
- Scattering
- Note
- Published through SciTech Connect.
01/01/1982.
"conf-820609-43"
"DE82017416"
American Nuclear Society annual meeting, Los Angeles, CA, USA, 6 Jun 1982.
Roussin, R.W.; Bhuiyan, S.I.; Lucius, J.L. - Funding Information
- W-7405-ENG-26
View MARC record | catkey: 14671993