European validation of the integral code ASTEC (EVITA)

Allelein, Hans-Josef; Neu, K.; Dorsselaere, J.P. Van; Müller, Kirill; Kostka, P.; Barnák, M.; Matejovič, Peter; Buján, Antonio Fernández de; Slaby, J.

doi:10.1016/s0029-5493(02)00346-1

Cited by 33 publications

(8 citation statements)

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“…In Europe, the development of the integral SA code ASTEC was carried by eight European countries from 2000 to 2003 [60]. The ASTEC code is composed of 10 modules to predict an entire SA sequence from the initiating event to fission product release out of the containment.…”

Section: Sa Analysis Codementioning

confidence: 99%

Review of thermal-hydraulic researches in severe accidents in light water reactors

Kataoka

2013

Journal of Nuclear Science and Technology

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The Tohoku Region Pacific Coast Earthquake and subsequent severe accident (SA) in Fukushima Daiichi Nuclear Power Station caused unprecedented disaster in Japan. Before this accident, considerable researches on SAs had been carried out in Japan. However, unfortunately, such researches could not prevent the accident due to the unexpected huge Tsunami. However, the researches on SAs become more and more important in order to make clear the causes of the accident in Fukushima and improve the safety of nuclear power plants in Japan. In view of this, review on researches on thermal hydraulics in SAs in light water reactors was carried out. Important thermal-hydraulic phenomena in SAs were identified. Research activities on each phenomenon were surveyed mainly based on the articles published in Journal of Nuclear Science and Technology of Atomic Energy Society of Japan.

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Section: Sa Analysis Codementioning

confidence: 99%

Review of thermal-hydraulic researches in severe accidents in light water reactors

Kataoka

2013

Journal of Nuclear Science and Technology

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“…Several codes for simulating an MCCI have been developed, such as TOLBIAC-ICB, ASTEC/MEDICIS, and COCO; these are used in order to estimate the timing or possibility of damage to the wall of the PCV for reactor safety evaluations [1][2][3]. Moreover, knowing the characteristics of the corium and MCCI products is important for decommissioning severely damaged reactors.…”

Section: Introductionmentioning

confidence: 99%

Thermodynamic evaluation of the solidification phase of molten core–concrete under estimated Fukushima Daiichi nuclear power plant accident conditions

Kitagaki

Yano

Ogino

et al. 2017

Journal of Nuclear Materials

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The solidification phases of molten core-concrete under the estimated molten coreconcrete interaction (MCCI) conditions in the Fukushima Daiichi Nuclear Power Plant Unit 1 were predicted using the thermodynamic equilibrium calculation tool, FactSage 6.2, and the NUCLEA database in order to contribute toward the 1F decommissioning work and to understand the accident progression via the analytical results for the 1F MCCI products. We showed that most of the U and Zr in the molten core-concrete forms (U,Zr)O 2 and (Zr,U)SiO 4 , and the formation of other phases with these elements is limited. However, the formation of (Zr,U)SiO 4 requires a relatively long time because it involves a change in the crystal structure from fcc-(U,Zr)O 2 to tet-(U,Zr)O 2 , followed by the formation of (Zr,U)SiO 4 by reaction with SiO 2. Therefore, the formation of (Zr,U)SiO 4 is limited under quenching conditions. Other common phases are the oxide phases, CaAl 2 Si 2 O 8 , SiO 2 , and CaSiO 3 , and the metallic phases of the Fe-Si and Fe-Ni alloys. The solidification phenomenon of the crust under quenching conditions and that of the molten pool under thermodynamic equilibrium conditions in the 1F MCCI progression are discussed.

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“…ASTEC code has progressively reached a larger European dimension through the European Commission Framework Programme projects: VASA (FP4) (Allelein et al, 2000), EVITA (FP5) (Allelein et al, 2003), SARNET (FP6) (Haste et al, 2009) and SARNET II (Severe Accident Research Network of Excellence) (FP7) (European Commission, 2008). Within these projects, and integrated in a series of other severe accident activities, the code has been validated over a large number of experimental programs performed in in-pile and out-of-pile facilities, at various scales.…”

Section: Introductionmentioning

confidence: 99%

PWR Medium Break LOCA source term analysis using ASTEC code

Luca

Buján

Sangiorgi

2015

Progress in Nuclear Energy

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a b s t r a c tDuring a hypothetical severe accident in a nuclear reactor most of the radioactivity would be released from the damaged core and the broken primary circuit in the form of aerosols and chemical vapours and gases. However the effective radio-toxicity per unit release from the fuel depends both on the released mass and on the chemical form as well. The fission products release and transport in the circuit has been analysed for a Medium Break cold leg Loss Of Coolant Accident (MBLOCA) combined with the total loss of electric power supply (Station Black Out) for a reference PWR 1000 reactor using the severe accident integral code ASTECV1.3 rev2. A detailed description of the reactor model is given. Sensitivity analyses were performed in terms of break location and accumulator availability. The break location has an impact on the carrier gas temperature and in turn on the chemical speciation of the fission products. Also retention is influenced by the different flow pathways establishing in the circuit. The accumulator availability extends the duration of the core degradation process affecting the total production of hydrogen and the cumulative released mass of low, semi and highly volatile fission products. In agreement with the ASTEC equilibrium chemistry model, the Mo/Cs (Rb) molar ratio has the most influence on the fraction of iodine which can reach the containment in the gaseous or highly volatile form. Due to the surplus of Cs, the predicted dominant iodine species to the containment are CsI and its dimers. A 3e4% of iodine which reaches the break exists in gaseous (HI) and highly volatile metal-iodides (I 2 MoO 2 , SnI 2 , SnI 4 ) species.

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European validation of the integral code ASTEC (EVITA)

Cited by 33 publications

References 0 publications

Review of thermal-hydraulic researches in severe accidents in light water reactors

Review of thermal-hydraulic researches in severe accidents in light water reactors

Thermodynamic evaluation of the solidification phase of molten core–concrete under estimated Fukushima Daiichi nuclear power plant accident conditions

PWR Medium Break LOCA source term analysis using ASTEC code

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