BUG2; two-dimensional multigroup diffusion and burnup in xy, rz geometries. [UNIVAC1108; FORTRAN IV and Assembly language]

Technical Report ·

OSTI ID:6365050

Fowler, T B; Tobias, M L; Fox, J N; Lawler, B E; Koppel, J U; Triplett, J R; Lynn, L L; Waldman, L A; Goldberg, I; Greebler, P; Kelley, M D; Davis, R A; Keck, C E; Redfield, J A; Murphy,; Todt, F

The two-dimensional, multigroup, neutron diffusion equations for x-y or r-z geometry are solved to obtain the multiplication factor and the spatial flux and power distributions. Complete reactor life histories with partial refueling at a number of reload points can be calculated. The depletion scheme of all burnable nuclides is specified by the user at execution time. A regionwide depletion scheme is used. Concentration dependent self-shielding factors may be applied to any nuclide.UNIVAC1108; FORTRAN IV and Assembly language; EXEC2, GAX23; 65,536 words of core storage, 2 million words of drum storage on one data channel, 3 tape units, and a peripheral printer.

• Research Organization: General Atomic Co., San Diego, CA (USA)
• OSTI ID: 6365050
• Report Number(s): ANL/NESC-438; ON: DE83048438
• Resource Relation: Other Information: Maxima of - 10 groups 20,000 space mesh points 255 regions 300 subregions 40 total nuclides 25 heavy nuclides 25 fission products and other burnable nuclides
• Country of Publication: United States
• Language: English

Related Subjects

73 NUCLEAR PHYSICS AND RADIATION PHYSICS
COMPUTER CODES
B CODES
NEUTRON DIFFUSION EQUATION
TWO-DIMENSIONAL CALCULATIONS
DIFFUSION
MULTIGROUP THEORY
NEUTRON FLUX
POWER DISTRIBUTION
DIFFERENTIAL EQUATIONS
EQUATIONS
NEUTRON TRANSPORT THEORY
RADIATION FLUX
TRANSPORT THEORY
654003* - Radiation & Shielding Physics- Neutron Interactions with Matter