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CSNI0077 SEMISCALE/TEST1011.

SEMISCALE/TEST1011, Large break LOCA blowdown starting from isothermal conditions of the primary coolant loop

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1. NAME OR DESIGNATION:  SEMISCALE/TEST1011.
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2. COMPUTERS
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Program name Package id Status Status date
SEMISCALE/TEST1011 CSNI0077/01 Report 10-JAN-1975

Machines used:

Package ID Orig. computer Test computer
CSNI0077/01 Many Computers
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3. DESCRIPTION OF TEST FACILITY

The system consisted of a pressure vessel with simulated reactor internals, an intact loop with a steam generator, pump, and pressurizer, a blowdown loop with rupture assemblies, a simulated steam generator, and a simulated pump and a pressure suppression system with a suppression tank and header. The semiscale apparatus was designed to simulate a 100% double-ended offset shear of the cold leg of a pressurized water reactor to investigate the system response to a Loss of Coolant Accident (LOCA). The break areas in Test lOll were reduced by 20% so that the system behavior would be significantly different from the previous semiscale tests.
  
Facility:
- 1 1/2 Loop Semiscale System consisted of a pressure vessel with internals, including 9 electrically heated fuel rod simulators;
- 1 intact loop with steam generator, simulated pump and bypass line
- 1 broken loop simulator
- emergency core cooling injection system
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4. DESCRIPTION OF TEST

Scaling Information:
- volume to break area equal to volume to break area of a large PWR
- other parameters not scaled, facility should only yield data for comparison to analyses
  
Parameters offered for Comparison:
- 2 pressures ( upper plenum and upstream break nozzle) ; containment pressure
- 4 differential pressures (hot leg-cold leg, across pump, pressure vessel containment)
- 6 local fluid densities
- 7 flow rates and including break flow
- liquid level
- time to start high pressure injection system (HPIS) and low pressure injection system (LPIS)
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6. PHENOMENA TESTED

Simulation of a Loss of Coolant Accident (LOCA) by the Semiscale Test Facility starting from isothermal conditions of the primary coolant loop.
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9. STATUS
Package ID Status date Status
CSNI0077/01 10-JAN-1975 Report Only
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10. REFERENCES
CSNI0077/01, included references:
- D.J. Barnum:
Comparative Analyses of Standard Problems - Standard Problem 2
Semiscale Test 1011 (Interim report I-296-75-1 1975)
- G.G. Loomis:
Summary Report Semiscale Mod-2A Heat Loss Characterization Test Series
EGG-SEMI-5448 (May 1981)
- E.M. Feldman and D.J. Olson:
Semiscale Mod-1 Program and System Description for the Blowdown
Heat Transfer Tests (ANCR-1230, August 1975)
- R.T. French:
An Evaluation of Piping Heat Transfer Piping Flow Regimes,
and Steam Generator Heat Transfer for the Semiscale Mod-1
Isothermal Tests (ANCR-1229, August 1975)
- M.T. Leonard:
RELAP5 Standard Model Description for the Semiscale Mod-2A System
EGG-SEMI-5692 (December 1981)
- D.W. Golden:
Steady State Control Model for the Standard RELAP5 Semiscale
Mod-2A System Model
RE-A-82-023 (April 1982)
- G.G. Loomis:
Summary of the Semiscale Program (1965-1986)
NUREG/CR-4945 - EGG-2509 (July 1987)
- John S. Martinelli:
"Semiscale Loss-of-Power Test Results"
11th Water Reactor Safety Research Information Meeting,
October 24-28, 1983
ECC-M-22283 Preprint
- L.J. Ball et al.:
Semiscale Program Description (TREE-NUREG-1210, May 1987)
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11. TEST DESIGNATION:  1011.
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12. PROGRAMMING LANGUAGE(S) USED
No specified programming language
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15. ESTABLISHMENT

     Aerojet Nuclear Company
     Idaho National Engineering Laboratory
     Idaho Falls, Idaho 83401
     U.S.A.
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16. MATERIAL AVAILABLE
CSNI0077/01
Interim report I-296-75-1
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17. CATEGORIES
  • Y. Integral Experiments Data, Databases, Benchmarks

Keywords: data, isothermal, loss-of-coolant accident, primary coolant loop.