Results of SOCRAT code development, validation and applications for NPP safety assessment under severe accidents

Bolshov, L. A.; Dolganov, K.; Kiselev, A. S.; Strizhov, V. F.

doi:10.1016/j.nucengdes.2018.11.013

Cited by 31 publications

(4 citation statements)

References 22 publications

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“…The results presented in Figs. [9][10][11][12] show that the process of concrete ablation advances faster in the upper part of the concrete cavity than in the lower part, due to a higher temperature, and lower natural convection.…”

Section: Thermal Boundary Conditionsmentioning

confidence: 99%

“…They forecasted the molten corium cooling behavior and the change in its properties. Different models and/or simulators have been developed in the past by a number of scientists; Bradley et al, [5] developed CORCON-Mod3, MEDICIS and MAAP were modelled by Crange et al, [10], SOCRAT was developed by Bolshov and Strizhov [11], Polidoro et al, [12] modeled CORIUM-2D and Farmer [4] developed CORQUENCH. Using CORQUENCH, Khurshid et al, [13] and [14] analyzed different parameters affecting concrete ablation.…”

Section: Introductionmentioning

confidence: 99%

“…For example, the relocation of molten corium has been studied via the Euler-based CFD approach by Kucala et al, [22], Ye et al, [23] and Na et al, [24]. Based on the Euler method, Bolshov and Valery [26] developed a multi-component environment to solve the mass, momentum, and energy equations based on the conservation principles for separate components. In their approach each component occupies a sub-area with specified boundaries in the numerical domain.…”

Section: Introductionmentioning

confidence: 99%

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A Thermal Approach for Modelling the Concrete Ablation during Molten Corium-Concrete Interaction

Khurshid,

Hassan,

Faizan

et al. 2024

Preprint

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An accident in a nuclear power plant involving a reactor core meltdown would result in instigation of molten corium, which is a mixture of nuclear fuel, claddings and structural components. In this paper, an enthalpy porosity model is proposed to comprehensively analyze the ablation of concrete during the Molten Corium and Concrete Interaction (MCCI) process. The developed numerical model is an extension of the enthalpy-porosity model and is termed as Corium-Concrete Enthalpy Porosity Model (CCEPM). The developed CCEPM Computational Fluid Dynamic (CFD) model can predict natural convection, melting and solidification. The developed model simplifies the complex phenomena of concrete ablation and its melting by incorporating the multi-region approach. The model was implemented in OpenFOAM, by developing a new solver which couples the buoyant-driven natural convection and conjugate heat transfer solvers. The thermal modelling and heat transfer capabilities of the developed solver were verified against experimental data sets. Additionally, the effect of various boundary conditions, concrete thermal conductivity and intensity of decay heat were analyzed to study their impact on concrete ablation. We observed significant low concrete ablation, controlled temperature and velocity field, for the water-cooled boundary condition. Accordingly, the ablation of concrete decreased by 17% just by imposing the water-cooled boundary condition. Similarly, when the thermal conductivity was decreased to 0.43 and 0.13 W/m. K, the ablation of concrete reduced by 38% and 75%, respectively. Furthermore, early cooling of molten corium to decrease decay heat was found to be an effective strategy to mitigate the concrete ablation successfully by 20%.

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Section: Thermal Boundary Conditionsmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

Section: Introductionmentioning

confidence: 99%

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A Thermal Approach for Modelling the Concrete Ablation during Molten Corium-Concrete Interaction

Khurshid,

Hassan,

Faizan

et al. 2024

Preprint

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“…В нем приводятся также соответствующие результаты прецизионных оценок выгорания топлива, накопления в нем радионуклидов и мощности ОТВ, выполненные при помощи кода ORIGEN2 по состоянию на момент аварии и спустя различное время после нее . Эти данные далее будут использоваться в качестве референтных для результатов моделирования с помощью отечественного интегрального кода СОКРАТ/В3 [6] .…”

Section: метод анализа и исходные данныеunclassified

Integrally Estimated Nuclide Inventory and Decay Heat from Potential Nuclear Waste Represented by the Spent Fuel from the Fukushima Daiichi Unit 1

Dolganov

2023

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Assuming an accident that may damage spent fuel assemblies at an NPP, nuclear fuel in the spent fuel pool can be considered as radioactive waste. Data on its nuclide composition and heat generation is seen important for the safety assessment of its storage and potential radiological consequences in case of an accident. The SOCRAT/V3 integral code designed to simulate severe accidents at nuclear power plants with VVER power units is able to estimate the nuclide inventory of irradiated fuel based on a one-through calculation and a built-in model. This code was used for a computational analysis of the spent fuel from the Fukushima Daiichi Unit 1, namely, evaluated was its nuclide composition and its decay heat release. The results were compared with the reference data from JAEA (Japan). The calculations demonstrate that in case of severe accident simulations, the use of an integral approach causes no large errors in the estimated nuclide inventory. For most dose-forming radionuclides in the spent fuel, relative deviation of activity from the reference data does not exceed 15%, which appears to be lower than the typical uncertainty under source term predictions. Despite certain overestimation tendency that can be noticed, the decay heat levels calculated for the spent fuel pool appear to be quite similar to those presented in the JAEA report differing by no more than 16% in the first years after the accident, and by 5% ten years after the accident.

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Estimation of system code SOCRAT/V3 accuracy to simulate the heat transfer in a pool of volumetrically heated liquid on the basis of BAFOND experiments

Dolzhenkov

Dolganov

Kapustin

et al. 2021

Annals of Nuclear Energy

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Results of SOCRAT code development, validation and applications for NPP safety assessment under severe accidents

Cited by 31 publications

References 22 publications

A Thermal Approach for Modelling the Concrete Ablation during Molten Corium-Concrete Interaction

A Thermal Approach for Modelling the Concrete Ablation during Molten Corium-Concrete Interaction

Integrally Estimated Nuclide Inventory and Decay Heat from Potential Nuclear Waste Represented by the Spent Fuel from the Fukushima Daiichi Unit 1

Estimation of system code SOCRAT/V3 accuracy to simulate the heat transfer in a pool of volumetrically heated liquid on the basis of BAFOND experiments

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