CSNI1024 CORA-W2. (Abstract last modified 01-FEB-1995)
1.
NAME OF EXPERIMENT - CORA-W2. 2.
COMPUTER FOR WHICH PROGRAM IS DESIGNED AND OTHER MACHINE VERSION PACKAGES AVAILABLE -
To request or retrieve programs click on the one of the active versions below.
A password and special authorization is required. Explanation of the status codes.
Machines used:
Package-ID Orig.Computer Test Computer
CSNI1024/01
3.
DESCRIPTION OF TEST FACILITY - 4.
DESCRIPTION OF TEST - 6.
PHENOMENA TESTED - 9.
STATUS 10.
REFERENCES - 11.
TEST DESIGNATION - W2. 12.
PROGRAMMING LANGUAGE -CSNI1024/01: 15.
ESTABLISHMENT - 16.
MATERIAL AVAILABLE -CSNI1024/01: 17.
CATEGORIES - Keywords: DATA, FUEL DAMAGE, LOSS-OF-COOLANT ACCIDENT, PWR REACTORS, WWER REACTORS
Program-name Package-ID Status
CORA-W2 CSNI1024/01 Report
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
- ISP 36 was organized as a "blind¿ 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
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
CSNI1024/01: 01-FEB-1995 Report Only
CSNI1024/01:
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
Gesellschaft für Anlagen und
Reaktorsicherheit (GRS) mbH
Schwertnergasse 1
D-50667 KOELN
Germany
NEA/CSNI/R(95)20 report
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