CSNI1024 CORA-W2.
CORA-W2, Experiment on Severe Fuel Damage for a VVER-type PWR
NAME OF EXPERIMENT, COMPUTER, DESCRIPTION OF TEST FACILITY, DESCRIPTION OF TEST, PHENOMENA TESTED, STATUS, REFERENCES, TEST DESIGNATION, LANGUAGE, ESTABLISHMENT, MATERIAL, CATEGORIES
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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
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
- 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
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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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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Keywords: VVER type reactor, data, fuel damage, loss-of-coolant accident, pwr reactors.