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NEA-0055 STDY-3.

STDY-3, Steady-State Parallel Channel Thermal Analysis of PWR

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1. NAME OR DESIGNATION OF PROGRAM:  STDY-3.
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2. COMPUTERS
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Program name Package id Status Status date
STDY-3 NEA-0055/01 Tested 01-APR-1966

Machines used:

Package ID Orig. computer Test computer
NEA-0055/01 IBM 7040 IBM 7040
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3. NATURE OF PHYSICAL PROBLEM SOLVED

A programme designed for the thermal analysis of a pressurized water nuclear reactor during steady-state operation.
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4. METHOD OF SOLUTION

Performs a complete steady-state, parallel channel thermal analysis of a rectangular water channel core with a  plate-type fuel element. The digital code performs enthalpy and temperature rise calculations for the code average channel from input parameters. Two-phase core pressure drop calculations are performed using Bettis developed semi-empirical correlations. These  calculations include the pressure drop effects of spatial acceleration, elevation, friction, and entrance and exit losses.
Hot channel pressure drop is computed by modifying the normal channel pressure drop by an input controlled plenum equation. To compute the hot channel flow where the flow vs pressure drop curve may contain inflections, a table of hot channel flow vs pressure drop is constructed by a stepwise reduction of flow from 100 percent of the average channel value to a flow producing steam quality in excess of 100 percent in the hot channel. The flow vs pressure drop  tableis then searched and interpolated for a pressure drop value within two percent of the hot channel pressure drop as determined by the plenum effects on the normal channel. The flow corresponding to  this pressure drop is considered the hot channel flow. Once the hot  channel flow is determined, enthalpy, temperature and steam quality  values are printed. Metal surface temperatures, mean centerline temperatures, and DNB (Departure from Nucleate Boiling) ratio calculations are made for each point from input properties and design equations.
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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM

The programme is capable of solving steady-state problems, including one and two pass cores, up to two hot channels per pass, up to thirty axial sections  per channel, and the parameter variation necessary for core safety margin calculations. For safety margin calculations, up to five values each of system pressure, flow, inlet temperature and power may be supplied for the same core. Each parameter is varied independently through the set of values supplied, and then returned  to its original value while another parameter is varied.
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6. TYPICAL RUNNING TIME

Typical computing time for a two-pass core containing a hot channel in each pass is 72 minutes.
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7. UNUSUAL FEATURES: UNUSUAL FEATURES OF THE PROGRAM
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8. RELATED OR AUXILIARY PROGRAMS: RELATED AND AUXILIARY PROGRAMS
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9. STATUS
Package ID Status date Status
NEA-0055/01 01-APR-1966 Tested at NEADB
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10. REFERENCES

- R.S. Pyle:
STDY-3 - A Program for the Thermal Analysis of a Pressurized Water    Nuclear Reactor during Steady-State Operation
  WAPD-TM-213 (June 1960).
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11. HARDWARE REQUIREMENTS: MACHINE REQUIREMENTS
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12. PROGRAMMING LANGUAGE(S) USED
Package ID Computer language
NEA-0055/01 FORTRAN-IV
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13. SOFTWARE REQUIREMENTS: OPERATING SYSTEM OR MONITOR UNDER WHICH PROGRAM IS EXECUTED
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14. OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS

ANY OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS
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15. NAME AND ESTABLISHMENT OF AUTHOR

The library has received the FORTRAN II version of this programme through the CETIS programmotheque. The FORTRAN IV version has been prepared by FIAT.
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16. MATERIAL AVAILABLE
NEA-0055/01
File name File description Records
NEA0055_01.001 SOURCE & DATA FORTRAN 4 1226
NEA0055_01.002 OUTPUT 130
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17. CATEGORIES
  • H. Heat Transfer and Fluid Flow

Keywords: enthalpy, heat transfer, pwr reactors, steady-state conditions, thermal analysis.