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[Program Service | Recent programs]

CSNI1023 CORA-13. (Abstract last modified 01-JUL-1993)


CORA-13 experiment on severe fuel damage, core degradation and quench

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Abstract Table of Contents:

NAME | COMPUTER | PROBLEM | SOLUTION | RESTRICTIONS | CPU | FEATURES | AUXILIARIES | STATUS | REFERENCES | REQUIREMENTS | LANGUAGE | OPERATING SYSTEM | OTHER RESTRICTIONS | AUTHOR | MATERIAL | CATEGORIES |

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1. NAME OF EXPERIMENT - CORA-13.


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.

Program-name         Package-ID  Status
CORA-13              CSNI1023/01 Report

Machines used:
Package-ID     Orig.Computer                       Test Computer
CSNI1023/01                                        

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
- two absorber rods added to the bundle to simulate interaction of fuel rods with absorber rod materials
- steam supply to provide superheated steam
- refill or quench phase added
Scaling Information: heated length of rods 1,000 mm, rod dimensions and pitch corresponding to original PWR fuel


4. DESCRIPTION OF TEST - Severe core damage experiment with quenching, of a full length simulated fuel bundle, heated with internal tungsten heaters. Investigations of the material relocation dynamics, quench phenomena and hydrogen generation.
The out-of-pile experiment CORA-13 was executed in November 1990.
In the experiments the decay heat is simulated by electrical heating. Great emphasis is given to the fact that the test bundles contain all materials used in light-water reactor fuel elements, to investigate the different material interactions. Pellets, cladding, grid spacers, absorber rods and the pertinent guide tubes are typical of those of commercial LWRs with respect to their compositions and radial dimensions.


6. PHENOMENA TESTED -
Objectives:
- Analysis of the heat-up and meltdown phases of a PWR 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).


9. STATUS
CSNI1023/01: 01-JUL-1993 Report Only


10. REFERENCES  - 
CSNI1023/01:
- M. Firnhaber et al.:
ISP ¿ 31 OECD/NEA-CSNI 
CORA-13 Experiment on severe fuel damage
NEA/CSNI/R(93)17 - GRS-106 ¿ KfK 5287 (July 1993)


11. TEST DESIGNATION - CORA-13.


12. PROGRAMMING LANGUAGE -CSNI1023/01: 


15. ESTABLISHMENT -
GRS
Koeln und Muenchen
Germany
KfK
Karlsruhe
Germany


16. MATERIAL AVAILABLE -CSNI1023/01:
Documentation in PDF format


17. CATEGORIES  - 

- Y. Integral Experiments Data, Databases, Benchmarks


Keywords: DATA, LOSS-OF-COOLANT ACCIDENT, TRANSIENTS

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