A User's Guide to the PLTEMP/ANL Code

Olson, Arne P.¹  Kalimullah, M.¹ 

[1]Argonne National Lab. (ANL), Argonne, IL (United States)

关键词: P CODES;    DEPARTURE NUCLEATE BOILING;    FUEL PLATES;    THERMAL HYDRAULICS;    RESEARCH REACTORS;    FUEL ASSEMBLIES;    MATHEMATICAL SOLUTIONS;    NATURAL CONVECTION;    FORTRAN;    HEAVY WATER;    TWO-DIMENSIONAL SYSTEMS;    COOLANTS;    TUBES;    LAMINAR FLOW;    GEOMETRY;    PLATES;    SAFETY MARGINS;    STEADY-STATE CONDITIONS;    TWO-DIMENSIONAL CALCULATIONS;    CORRELATIONS;    INSTABILITY;    PERFORMANCE;    SLABS;    TEMPERATURE DISTRIBUTION;    REACTOR CORES;    FORCED CONVECTION;    SUBCOOLED BOILING;   

DOI  :  10.2172/1234525
RP-ID  :  ANL/RERTR/TM--11-22 Rev.1-Version 4.2
PID  :  OSTI ID: 1234525
Others  :  Other: 121096
Others  :  TRN: US1600226

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来源: SciTech Connect

【 摘 要 】

PLTEMP/ANL V4.2 is a FORTRAN program that obtains a steady-state flow and temperature solution for a nuclear reactor core, or for a single fuel assembly. It is based on an evolutionary sequence of ''PLTEMP'' codes in use at ANL for the past 20 years. Fueled and non-fueled regions are modeled. Each fuel assembly consists of one or more plates or tubes separated by coolant channels. The fuel plates may have one to five layers of different materials, each with heat generation. The width of a fuel plate may be divided into multiple longitudinal stripes, each with its own axial power shape. The temperature solution is effectively 2-dimensional. It begins with a one-dimensional solution across all coolant channels and fuel plates/tubes within a given fuel assembly, at the entrance to the assembly. The temperature solution is repeated for each axial node along the length of the fuel assembly. The geometry may be either slab or radial, corresponding to fuel assemblies made of a series of flat (or slightly curved) plates, or of nested tubes. A variety of thermal-hydraulic correlations are available with which to determine safety margins such as Onset-of- Nucleate boiling (ONB), departure from nucleate boiling (DNB), and onset of flow instability (FI). Coolant properties for either light or heavy water are obtained from FORTRAN functions rather than from tables. The code is intended for thermal-hydraulic analysis of research reactor performance in the sub-cooled boiling regime. Both turbulent and laminar flow regimes can be modeled. Options to calculate both forced flow and natural circulation are available. A general search capability is available (Appendix XII) to greatly reduce the reactor analyst???s time.