Core loss during a severe accident (COLOSS)

Adroguer, B.; Bertrand, Frédèric; Chatelard, P.; Cocuaud, N.; Dorsselaere, J.P. Van; Bellenfant, L.; Knocke, D.; Bottomley, D.; Vrtílková, Věra; Bělovský, L.; Mueller, K.; Hering, W.; Homann, Ch.; Krauss, W.; Miassoedov, A.; Schanz, G.; Steinbrück, M.; Stuckert, J.; Hózer, Zoltán; Bandini, G.; Birchley, J.; Berlepsch, T.v.; Kleinhietpass, I.; Buck, M.; Benitez, J. A. Fernandez; Virtanen, E.; Marguet, Serge; Azarian, G.; Caillaux, A.; Plank, Helmut; Boldyrev, A.V.; Veshchunov, M.S.; Kobzar, V.; Zvonarev, Yu.A.; Goryachev, A. V.

doi:10.1016/s0029-5493(02)00344-8

Cited by 32 publications

(21 citation statements)

References 2 publications

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“…A limited number of reflooding experiments with fuel rod bundles or B 4 C control blades (see Section 2.2.4 on core reflooding) have shown that the oxidation of B 4 C may contribute to the production of both hydrogen and other gases such as CO, CO 2 , and large amounts of B-compound aerosols [15]. The latter gases are mainly produced during reflooding or the final rapid cooling phase.…”

Section: Oxidation Of B 4 C In Steammentioning

confidence: 99%

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In-Vessel Core Degradation

2012

Nuclear Safety in Light Water Reactors

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Section: Oxidation Of B 4 C In Steammentioning

confidence: 99%

“…Several experiments and models related to this subject were led during the EC projects CIT and COLOSS. The results are available in several reports, and a good overview may be found, respectively, in [23] and [15].…”

Section: Fuel Rod Dissolutionmentioning

confidence: 99%

In-Vessel Core Degradation

2012

Nuclear Safety in Light Water Reactors

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“…Adroguer [9] published the findings from a number of experiments including the effects of hydrogen production, oxidization of cladding, and effects of molten zircaloy on the cladding geometry. From these experiments, software codes were established and implemented in programs to predict the effects of a LOCA.…”

Section: Modeling Of the Heat-up Of Zircaloy-4 Cladding And Burn-up Cmentioning

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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“…Accordingly, the RPV failure is determined as time-and temperature-dependent process. In disclosing the complexity of the structural behaviors, several experimental programs such as FOREVER, USNRC/SNL, CORVIS, and LIVE had been performed (Willschutz et al, 2001(Willschutz et al, , 2006Nicolas et al, 2003;Adroguer et al, 2003;Gaus-Liu et al, 2010;Buck et al, 2010), the results of which validated the predictability on the failure time, mode and site by using the finite element method (FEM). The already conducted experiment span a wide range of pressure (0.2-10 MPa) and temperature gradient (130-1500 • C) (Theofanous et al, 1997).…”

Section: Introductionmentioning

confidence: 99%

Structural integrity investigation for RPV with various cooling water levels under pressurized melting pool

et al. 2018

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Abstract. The strategy denoted as "in-vessel retention (IVR)" is widely used in severe accident (SA) management by most advanced nuclear power plants. The essence of IVR mitigation is to provide long-term external water cooling in maintaining the reactor pressure vessel (RPV) integrity. Actually, the traditional IVR concept assumed that RPV was fully submerged into the water flooding, and the melting pool was depressurized during the SA. The above assumptions weren't seriously challenged until the occurrence of Fukushima accident on 2011, suggesting the structural behavior had not been appropriately assessed. Therefore, the paper tries to address the structure-related issue on determining whether RPV safety can be maintained or not with the effect of various water levels and internal pressures created from core meltdown accident. In achieving it, the RPV structural behaviors are numerically investigated in terms of several field parameters, such as temperature, deformation, stress, plastic strain, creep strain, and total damage. Due to the presence of high temperature melt on the inside and water cooling on the outside, the RPV failure is governed by the failure mechanisms of creep, thermal-plasticity and plasticity. The creep and plastic damages are interacted with each other, which further accelerate the failure process. Through detailed investigation, it is found that the internal pressure as well as water levels plays an important role in determining the RPV failure time, mode and site.

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Core loss during a severe accident (COLOSS)

Cited by 32 publications

References 2 publications

In-Vessel Core Degradation

In-Vessel Core Degradation

Thermal Cooling Limits of Sbotaged Spent Fuel Pools

Structural integrity investigation for RPV with various cooling water levels under pressurized melting pool

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