THE CALCULATION OF THERMAL NEUTRON SPECTRA

Technical Report · Sat Oct 31 04:00:00 EST 1959

OSTI ID:4264060

Schofield, P; Hassitt, A

Results are reported of calculations of thermal neutron spectra, with particular reference to graphite moderated systems. Cross sections for inelastic scattering of neutrons by polycrystalline graphite were simplified phonon spectrum for the crystal lattice. At high temperatures and high neutron energies, multiphonon scattering is important and has been taken into account approximately. These cross sections are used first to calculate the thermal neutron spectrum in an infinite homogeneous capturing medium. The neutron balance equation is solved by dividing the energy range into thirty groups, and solving the resulting set of are compared with those obtained on the assumption that the moderator is a monatomic gas. The infinite medium solutions are taken as a starting point for the calculation of spectra in inhomogeneous systems. Multigroup diffusion theory is used to calculate the neutron spectrum over two adjacent half-spaces at different temperatures. The variation of the thermal spectrum over a simple lattice is calculated. The relative reaction rates of various materials in neutron fluxes with the calculated energy spectra are computed, and compared, where possible, with experimental results. (auth)

• Research Organization: Atomic Energy Research Establishment, Harwell, Berks, Eng.
• NSA Number: NSA-13-007078
• OSTI ID: 4264060
• Report Number(s): A/CONF.15/P/18
• Country of Publication: Country unknown/Code not available
• Language: English

Related Subjects

CAPTURE
CONFIGURATION
CROSS SECTIONS
CRYSTALS
DISTRIBUTION
EQUATIONS
ERRORS
GASES
GRAPHITE MODERATOR
GROUP THEORY
HIGH TEMPERATURE
HOMOGENEOUS REACTORS
INELASTIC SCATTERING
INTERACTIONS
LATTICES
MATERIALS TESTING
MODERATORS
NUMERICALS
PHOTONS
REACTOR CORE
REACTORS
SCATTERING
SPECTRA
STANDARDS
TEMPERATURE
THERMAL NEUTRONS
TRANSPORT THEORY
VARIATIONS
ZONES