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International standard problem ISP36. Cora-W2 experiment on severe fuel damage for a Russian type PWR

Abstract

An OECD/NEA-CSNI International Standard Problem (ISP) has been performed on the experimental comparison basis of the severe fuel damage experiment CORA-W2. The out-of-pile experiment CORA-W2 was executed in February 1993 at he Forschungszentrum Karlsruhe. The objective of this experiment was the investigation of the behavior of a Russian type PWR fuel element (VVER-1000) during early core degradation. The main difference between a Western type and a Russian type PWR bundle is the B{sub 4}C absorber rod instead of AgInCd. Measured quantities ar boundary conditions, bundle temperature, hydrogen generation and the final bundle configurations after cooldown. The ISP was conducted as a blind exercise. Boundary conditions were estimated using ATHLET-CD. Six different severe accident codes were used. The comparisons between experimental and analytical results were grouped by codes and examined separately. The thermal behavior up to significant oxidation has been predicted quite well. Larger deviations have been observed for the oxidation-induced temperature escalation, both time of onset and maximum temperature as well. The bundle behavior is greatly influenced by chemical interactions involving B{sub 4}C absorber rod material, which failed relatively early at low temperature due to eutectic interaction between B{sub 4}C and SS cladding as well as the SS guide tube.  More>>
Authors:
"NONE"
Publication Date:
Dec 31, 1996
Product Type:
Technical Report
Report Number:
NEA-CSNI-R-95-20
Reference Number:
SCA: 210200; 990200; PA: AIX-28:037563; EDB-97:078525; NTS-97:012729; SN: 97001796022
Resource Relation:
Other Information: DN: 36 refs.; PBD: 1996
Subject:
21 NUCLEAR POWER REACTORS AND ASSOCIATED PLANTS; 99 MATHEMATICS, COMPUTERS, INFORMATION SCIENCE, MANAGEMENT, LAW, MISCELLANEOUS; REACTOR SAFETY EXPERIMENTS; TEST REACTORS; A CODES; BORON CARBIDES; CHEMICAL REACTION KINETICS; COMPUTERIZED SIMULATION; DECOMPOSITION; DISSOLUTION; FUEL ELEMENTS; FUEL-CLADDING INTERACTIONS; GUIDE TUBES; I CODES; INTERSTITIAL HYDROGEN GENERATION; K CODES; M CODES; MELTING; OXIDATION; PWR TYPE REACTORS; R CODES; REACTOR CORE DISRUPTION; RISK ASSESSMENT; RUSSIAN FEDERATION; S CODES; STAINLESS STEELS; TEMPERATURE RANGE 1000-4000 K; THERMAL DEGRADATION; URANIUM DIOXIDE; ZIRCALOY 4; ZIRCONIUM; ZIRCONIUM OXIDES
OSTI ID:
484122
Research Organizations:
Nuclear Energy Agency, 75 - Paris (France). Committee on the Safety of Nuclear Installations
Country of Origin:
NEA
Language:
English
Other Identifying Numbers:
Other: ON: DE97625637; TRN: XN9600131037563
Availability:
INIS; OSTI as DE97625637
Submitting Site:
INIS
Size:
149 p.
Announcement Date:

Citation Formats

International standard problem ISP36. Cora-W2 experiment on severe fuel damage for a Russian type PWR. NEA: N. p., 1996. Web.
International standard problem ISP36. Cora-W2 experiment on severe fuel damage for a Russian type PWR. NEA.
1996. "International standard problem ISP36. Cora-W2 experiment on severe fuel damage for a Russian type PWR." NEA.
@misc{etde_484122,
title = {International standard problem ISP36. Cora-W2 experiment on severe fuel damage for a Russian type PWR}
abstractNote = {An OECD/NEA-CSNI International Standard Problem (ISP) has been performed on the experimental comparison basis of the severe fuel damage experiment CORA-W2. The out-of-pile experiment CORA-W2 was executed in February 1993 at he Forschungszentrum Karlsruhe. The objective of this experiment was the investigation of the behavior of a Russian type PWR fuel element (VVER-1000) during early core degradation. The main difference between a Western type and a Russian type PWR bundle is the B{sub 4}C absorber rod instead of AgInCd. Measured quantities ar boundary conditions, bundle temperature, hydrogen generation and the final bundle configurations after cooldown. The ISP was conducted as a blind exercise. Boundary conditions were estimated using ATHLET-CD. Six different severe accident codes were used. The comparisons between experimental and analytical results were grouped by codes and examined separately. The thermal behavior up to significant oxidation has been predicted quite well. Larger deviations have been observed for the oxidation-induced temperature escalation, both time of onset and maximum temperature as well. The bundle behavior is greatly influenced by chemical interactions involving B{sub 4}C absorber rod material, which failed relatively early at low temperature due to eutectic interaction between B{sub 4}C and SS cladding as well as the SS guide tube. Regarding the complex material interaction larger differences can be recognized between calculated and measured results because of inappropriate models for material relocation and solidification processes and the lack of models describing the interactions of absorber rod materials with the fuel rods. For the total amount of H{sub 2} generated, acceptable agreement could be achieved, if the total of oxidized zirconium was calculated correctly. The oxidation of stainless steel components and B{sub 4}C were not treated. In general the confidence in code predictions decreases with processing core damage. 36 refs.}
place = {NEA}
year = {1996}
month = {Dec}
}