Separate-effect tests on zirconium cladding degradation in air ingress situations

Duriez, Christian; Steinbrück, M.; Ohâi, D.; Meleg, T.; Birchley, J.; Haste, T.

doi:10.1016/j.nucengdes.2008.10.017

Cited by 52 publications

(32 citation statements)

References 10 publications

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“…The comparison of the results of this study with recent studies in steam [3], oxygen [4], and air [16,17] shows strong differences in oxidation behavior up to about 1050°C. On the other hand, the pre-transition oxidation kinetics up to 1050°C and the overall oxidation at higher temperatures are quite comparable and less dependent on composition and boundary conditions.…”

Section: Discussionsupporting

confidence: 59%

Oxidation of Advanced Zirconium Cladding Alloys in Steam at Temperatures in the Range of 600–1200 °C

Vér²,

2011

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The oxidation kinetics of the classical pressurized water reactors (PWR) cladding alloy Zircaloy-4 have been extensively investigated over a wide temperature range from operational conditions to beyond design basis accident (BDBA) temperatures. In recent years, new cladding alloys optimized for longer operation and higher burn-up are used in Western light water reactors (LWR). This paper presents the results of thermo-gravimetric tests with Zircaloy-4 as the reference material, Duplex DX-D4, M5 Ò (both AREVA), ZIRLO TM (Westinghouse), and the Russian E110 alloy. All materials were investigated in isothermal and transient tests in a thermal balance with steam furnace. Post-test analyses were performed by lightmicroscopy and neutron radiography for investigation of the hydrogen absorbed by the metal. Strong and varying differences (up to 800%) in oxidation kinetics between the alloys were found at up to 1000°C, where the breakaway effect plays a role. Less but significant differences (ca. 30%) were observed at 1100 and 1200°C. Generally, the M5 Ò alloy revealed the lowest oxidation rate over the temperature range investigated whereas the behavior of the other alloys was considerably dependent on temperature. A strong correlation was found between oxide scale structure and amount of absorbed hydrogen.

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Section: Discussionsupporting

confidence: 59%

Oxidation of Advanced Zirconium Cladding Alloys in Steam at Temperatures in the Range of 600–1200 °C

Vér²,

2011

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“…Oxidation of the cladding, rods, and other components in the core constructed in zirconium base alloy, by steam, is a critical issue in LWR accident producing severe core damage. During a severe accident in a nuclear reactor, several works are focusing on core degradation by metal core components oxidation by air or steam (e.g., [15][16][17][18]), on the other hand, Royl et al [19] have made a hydrogen risk analysis during severe accident using computational fluid dynamics (CFD) codes, to obtain localized detailed information and supplement the results of lumped parameter codes, which focus on global or average effect.…”

Section: Simulations and Discussionmentioning

confidence: 99%

Severe Accident Simulation of the Laguna Verde Nuclear Power Plant

Espinosa-Paredes

Camargo-Camargo²,

Nuñez-Carrera

2012

Science and Technology of Nuclear Installations

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The loss-of-coolant accident (LOCA) simulation in the boiling water reactor (BWR) of Laguna Verde Nuclear Power Plant (LVNPP) at 105% of rated power is analyzed in this work. The LVNPP model was developed using RELAP/SCDAPSIM code. The lack of cooling water after the LOCA gets to the LVNPP to melting of the core that exceeds the design basis of the nuclear power plant (NPP) sufficiently to cause failure of structures, materials, and systems that are needed to ensure proper cooling of the reactor core by normal means. Faced with a severe accident, the first response is to maintain the reactor core cooling by any means available, but in order to carry out such an attempt is necessary to understand fully the progression of core damage, since such action has effects that may be decisive in accident progression. The simulation considers a LOCA in the recirculation loop of the reactor with and without cooling water injection. During the progression of core damage, we analyze the cooling water injection at different times and the results show that there are significant differences in the level of core damage and hydrogen production, among other variables analyzed such as maximum surface temperature, fission products released, and debris bed height.

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“…al. [8] , a part of the Severe Accident Research network of the European Commission discussed the effects of the zirconium cladding degradation in situations that involved air ingress. These experiments expanded upon previous ones that only included steam as the oxidizing environment.…”

Section: Yearly and Final Progress Report September 10 2010 De-fg07-mentioning

confidence: 99%

Thermal Cooling Limits of Sbotaged Spent Fuel Pools

Hughes¹,

Lin²

2010

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Submitted to:DOE Project Manager: Nancy Elizondo, elizonna@id.deo.gov Procurement Services Division: psdrept@id.doe.govProject Objective: To develop the understanding and predictive measures of the post "loss of water inventory" hazardous conditions as a result of the natural and/or terrorist acts to the spent fuel pool of a nuclear plant. This includes the thermal cooling limits to the spent fuel assembly (before the onset of the zircaloy ignition and combustion), and the ignition, combustion, and the subsequent propagation of zircaloy fire from one fuel assembly to others. Background:A National Academy of Sciences expert panel issued a public plea in 2005 to increase the security measures to safeguard the spent fuel pools of the nation's nuclear plants. The panel strongly recommended increased research to assess the Maximum-Credible Scenarios (MCS) and to improve the current understanding of the hazardous conditions, as well as the thermal runaway condition of zircaloy ignition and combustion that would follow the loss of spent fuel pool water. This research is critical to improving the security of the nation's nuclear plants as well as the spent fuel infrastructure of the next generation (Generation IV) nuclear units because they all share the common characteristic of having to cope with extended years of spent fuel pool cooling.

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Separate-effect tests on zirconium cladding degradation in air ingress situations

Cited by 52 publications

References 10 publications

Oxidation of Advanced Zirconium Cladding Alloys in Steam at Temperatures in the Range of 600–1200 °C

Oxidation of Advanced Zirconium Cladding Alloys in Steam at Temperatures in the Range of 600–1200 °C

Severe Accident Simulation of the Laguna Verde Nuclear Power Plant

Thermal Cooling Limits of Sbotaged Spent Fuel Pools

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