FRAP-T4; FRAP-T; transient analysis of oxide fuel rods. [CDC7600; IBM360; FORTRAN (99%) and Assembly language (1%). The Assembly language is used in a few of the INEL Environmental Routines]

Technical Report ·

OSTI ID:6345870

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,; Bohn, M P

The FRAP-T series of programs predict the thermal, mechanical and internal gas pressure response of a water-cooled, zircaloy-clad, oxide fuel rod during a loss-of-coolant accident (LOCA), power-coolant mismatch (PCM), or flow blockage accident. They consider the coupled effects of these response parameters by iterating to convergence at each time-step in a transient. Transient heat conduction is analyzed. Material and water properties required by FRAP-T are supplied by a material property subcode, MATPRO, and the STH20 subroutine. FRAP-T4 is the fourth in the series with each release incorporating the most recent advancements made in fuel rod response analysis models.CDC7600;IBM360; FORTRAN (99%) and Assembly language (1%). The Assembly language is used in a few of the INEL Environmental Routines.; OS/360 MVT (IBM360), SCOPE 2.1.1 (CDC7600); 147,000 (octal) words of SCM and 63,000 (octal) words of LCM are required for FRAP-T4. Execution of FRAP-T on the IBM360 required 350K bytes of storage and 450K bytes were used for compilation.

• Research Organization: EG and G Idaho, Inc., Idaho Falls (USA)
• OSTI ID: 6345870
• Report Number(s): ANL/NESC-658; ON: DE83048658
• Country of Publication: United States
• Language: English

Related Subjects

22 GENERAL STUDIES OF NUCLEAR REACTORS
220300* -- Nuclear Reactor Technology-- Fuel Elements
ACCIDENTS
COMPUTER CODES
ENERGY TRANSFER
F CODES
FAILURES
FLOW BLOCKAGE
FUEL ELEMENTS
FUEL RODS
FUEL-CLADDING INTERACTIONS
HEAT TRANSFER
LOSS OF COOLANT
POWER-COOLING-MISMATCH ACCIDENTS
REACTOR ACCIDENTS
REACTOR COMPONENTS
REACTORS
THERMAL REACTORS
TRANSIENTS