BRT-I: Battelle-Revised-Thermos
- Battelle Pacific Northwest Labs., Richland, WA (United States)
Nature of Physical Problem Solved: The code computes the space dependent thermal neutron density, flux and current spectra over the energy range 0 to 0.683 eV in either slab or cylindrical geometry. Method of Solution: The neutron density is computed from the collision probability form of the integral transport theory matrix equation using either a combination of power iteration, overrelaxation and extrapolation or straight power iteration. The neutron currents are computed from either the gradient of the scaler flux or the uncollided flux matrix. The flux and current spectra is used to weight point thermal cross sections over an arbitrary thermal energy range for use in multigroup transport or diffusion theory codes.
- Research Organization:
- Battelle Pacific Northwest Labs., Richland, WA (United States)
- Sponsoring Organization:
- USDOE National Nuclear Security Administration (NNSA), Nuclear Criticality Safety Program (NCSP); US Atomic Energy Commission (AEC)
- DOE Contract Number:
- AT(45-1)-1830
- NSA Number:
- NSA-24-052354
- OSTI ID:
- 4103405
- Report Number(s):
- BNWL-1434
- Country of Publication:
- United States
- Language:
- English
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