Chemical-Physical Behavior of Light Water Reactor Core Components Tested Under Severe Reactor Accident Conditions in the CORA Facility

Hofmann, P.; Hagen, S.; Noack, V.; Schanz, G.; Sepold, L.

doi:10.13182/nt118-200

Cited by 46 publications

(16 citation statements)

References 17 publications

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“…CORA core degradation early stage experiments started in the 1980s by Germany's KFK National Lab aiming at International Standard Problem 31. The CORA experiment focused on PWR, BWR, WER core degradation and quench situation research (Schanz et al, 1992;Hofmann et al, 1997). It includes 19 tests; CORA-13 is represented here as a reference case (Firnhaber et al, 1993a).…”

Section: Experimental Research On Core Degradation and Melting Materimentioning

confidence: 99%

“…These accidents caused a large amount of damage to local populations and remind people to respect the severity of their occurrences (Matzke, 1982;Simmons, 2013). Details of several core degradation and material migration experiments are then described (Hofmann et al, 1997;Schwarz et al, 1999). Some of the experiments focus directly on core degradation and material migration phenomenon (Hagen et al, 1996;Hofmann FigURe 1 | Core melting of AP1000 (Zhang et al, 2015a,b Repetto et al, 2003;Van Dorsselaere et al, 2006), while other experiments deal with separate effects testing of core degradation and material migration phenomenon (Thomsen, 1998;Steinbrück et al, 2010).…”

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confidence: 99%

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Review on Core Degradation and Material Migration Research in Light-Water Reactors

Wang

Shi

et al. 2018

Front. Energy Res.

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Section: Experimental Research On Core Degradation and Melting Materimentioning

confidence: 99%

mentioning

confidence: 99%

Review on Core Degradation and Material Migration Research in Light-Water Reactors

Wang

Shi

et al. 2018

Front. Energy Res.

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“…The behaviour of fuel rods in high temperature steam, the quenching at high temperature, the role of control rods and the phenomena related to air ingress scenarios (Hózer, 2003) were addressed in the earlier tests. Integral severe accidents tests with electrically heated fuel bundles have been carried out in the CORA (Schanz et al, 1992;Hofmann et al, 1997), QUENCH (Hofmann et al, 1998;Miassoedov et al, 2008;Steinbrück et al, 2006;) and PARAMETER ) tests facilities and in some of these tests the high temperature bundles were quenched by cold water after intense oxidation. Similarly to the previous CODEX tests, the CORA, the QUENCH and the PARAMETER experiments simulated relatively fast heat-up of fuel rods and none of the tests was performed in hydrogen rich steam.…”

Section: Integral Testsmentioning

confidence: 99%

Experimental simulation of the Paks-2 cleaning tank incident through separate effect and integral tests

Hózer¹,

Horváth²,

Kunstár³

et al. 2011

Nuclear Engineering and Design

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Following the ex-core cleaning of VVER type fuel assemblies, a severe fuel damage took place at the Paks NPP. Assemblies with high decay-heat remained without cooling for seven hours. As a common result of high temperature oxidation and final water quenching the assemblies were broken and the fuel rods fragmented.In order to improve the understanding of the phenomena that took place during the Paks-2 incident, several series of separate effect and integral tests have been carried out. The separate effect tests investigated the phenomena related to the upward shift of fuel assemblies, to the role of the spacer grid and the spring in the loss of cladding integrity, and to the importance of high hydrogen content in the vessel's atmosphere on the embrittlement of zirconium. Integral tests were carried out in the CODEX facility to simulate the whole scenario of the incident using electrically heated fuel rods.3

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“…Analysis of the TMI-2 accident [2] and results of various integral inpile and out-of-pile experiments (CORA [3], LOFT [4], PHEBUS [5], PBF [6], PARAMETER [7]) have shown that before the water succeeds in cooling the fuel pins there could be an enhanced oxidation of the Zircaloy cladding and other core components that in turn causes a sharp increase in temperature, hydrogen production and fission product release. At temperatures of 1000°C and above, various physico-chemical processes occur in steam or/and air containing atmospheres, for instance: -Strong exothermic oxidation of zirconium alloys used as fuel rod cladding materials in steam or air -Hydrogen uptake of zirconium alloys during steam oxidation -Eutectic interaction between different reactor materials, as, e.g.…”

Section: Introductionmentioning

confidence: 99%

Application of neutron radiography to study material processes during hypothetical severe accidents in nuclear reactors

et al. 2011

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Neutron radiography was applied for investigations of nuclear fuel cladding and control rod behaviour during steam oxidation at temperatures between 1123 and 1673 K under severe nuclear accident conditions. This article gives an overview of these investigations. At KIT, loss of coolant and severe nuclear accidents were experimentally simulated. Post-test examinations of damaged control rods were performed. In order to separate the effect of the neutron absorber and control rod structure materials, radiographs taken with different neutron spectra were combined. It could be seen clearly from these data that local melting resulting from the eutectic reaction between the stainless steel control rod cladding and the Zircaloy-4 guide tube. The uptake of hydrogen during steam oxidation and its diffusion in Zircaloy-4 was investigated in ex situ and in situ radiography experiments at temperatures of 1123 to 1673 K. The kinetics of hydrogen uptake and diffusion was determined. The oxide layer morphology strongly influences the hydrogen concentration in steam oxidized zirconium alloys. Differences of nearly one order of magnitude were found in samples withdrawn from large scale QUENCH experiments. The hydrogen diffusion coefficients were determined for various temperatures.Whereas the diffusion coefficients at 1123 and 1173 K agree well with values expected from literature values for pure Zr, at higher temperatures a faster diffusion was found. The determined activation energy of the hydrogen diffusion is about 10 % higher than published values in the literature.

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Chemical-Physical Behavior of Light Water Reactor Core Components Tested Under Severe Reactor Accident Conditions in the CORA Facility

Cited by 46 publications

References 17 publications

Review on Core Degradation and Material Migration Research in Light-Water Reactors

Review on Core Degradation and Material Migration Research in Light-Water Reactors

Experimental simulation of the Paks-2 cleaning tank incident through separate effect and integral tests

Application of neutron radiography to study material processes during hypothetical severe accidents in nuclear reactors

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