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CSNI1024 CORA-W2.
last modified: 01-FEB-1995 | catalog | new | search |

CSNI1024 CORA-W2.

CORA-W2, Experiment on Severe Fuel Damage for a VVER-type PWR

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1. NAME OF EXPERIMENT:  CORA-W2.
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2. COMPUTERS
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Program name Package id Status Status date
CORA-W2 CSNI1024/01 Report 01-FEB-1995

Machines used:

Package ID Orig. computer Test computer
CSNI1024/01 Many Computers
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3. DESCRIPTION OF TEST FACILITY

CORA test facility operated at Kernforschungszentrum Karlsruhe serving to study the behaviour of PWR fuel elements under severe accident conditions
- fuel rod bundle with heated and unheated rods under controlled thermal-hydraulic boundary conditions, high temperature radiation shield surrounding the bundle
- heated fuel rods consist of 6 mm diameter tungsten rod surrounded by UO2 annular pellets and Zr-Nb cladding material, arranged in hexagonal array to represent VVER type fuel elements
- two absorber rods added to the bundle to simulate interaction of fuel rods with absorber rod materials
- steam supply to provide superheated steam
- slow cooldown phase terminated the experiment
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4. DESCRIPTION OF TEST

- ISP 36 was organized as a "blid" exercise, only thermal initial and boundary conditions were given
- heat-up phase predicted quite well, however larger deviations observed for onset of oxidation induced temperature escalation
- modeling of the interaction of control rod or spacer material with fuel cladding not possible
- hydrogen production and release rates not described correctly, several codes did not properly treat oxidation of steel components and boron carbide oxidation
- final core blockage predicted by only some calculations
- code user influence important
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6. PHENOMENA TESTED

The objectives were:
Analysis of the heat-up and meltdown phases of a VVER-type fuel element in the CORA-test facility
- specific emphasis on reliability and accuracy of severe accident computer codes
- investigation into the thermal and mechanical behaviour of a fuel bundle at high temperatures (e.g. formation of blockages, fragmentation of rods)
- study of physico-chemical processes during core degradation (e.g., oxidation of cladding and other metallic components, hydrogen formation)
- 'blind' post-test analyses
  
Scaling Information:
-heated length of rods 1,000 mm, rod dimensions and hexagonal arrangement corresponding to original VVER fuel
- typical separate effects tests
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9. STATUS
Package ID Status date Status
CSNI1024/01 01-FEB-1995 Report Only
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10. REFERENCES
CSNI1024/01, included references:
M. Firnhaber, et al.;
OECD/NEA-CSNI INTERNATIONAL STANDARD PROBLEM No. 36;
CORA-W2 Experiment on Severe Fuel Damage for a Russian-type PWR
Comparison Report;
NEA/CSNI/R(95)20, February 1996; also referenced as OCDE/GD(96)19
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11. TEST DESIGNATION:  W2.
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12. PROGRAMMING LANGUAGE(S) USED
No specified programming language
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15. ESTABLISHMENT

Gesellschaft fur Anlagen und
Reaktorsicherheit (GRS) mbH
Schwertnergasse 1
D-50667 KOELN
Germany
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16. MATERIAL AVAILABLE
CSNI1024/01
NEA/CSNI/R(95)20 report
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17. CATEGORIES
  • Y. Integral Experiments Data, Databases, Benchmarks

Keywords: VVER type reactor, data, fuel damage, loss-of-coolant accident, pwr reactors.