A study of different cases of VVER reactor core flooding in a large break loss of coolant accident

Bezrukov, Y.; Schekoldin, Vladimir Ivanovitc; Zaitsev, S. I.; Churkin, A. N.; Lisenkov, E.

doi:10.1051/epjn/e2015-50005-9

Cited by 5 publications

(3 citation statements)

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“…The PZR space in the four-region model is divided into regions of continuous vapor, discrete water, discrete vapor, and continuous water (Baghban et al, 2016;Bezrukov et al, 2016;Moghanaki & Rahgoshay, 2014;Zhong et al, 2019). This model considers the effects of vapor bubbles and water droplets' residence in the continuous water and vapor regions.…”

Section: Introductionmentioning

confidence: 99%

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A developed analytical model for the pressurizer unit in nuclear power plants

Sheta

Ali

Fikry

et al. 2021

Journal of Radiation Research and Applied Sciences

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The pressure control in the pressurized water reactor (PWR) primary loop is key to secure the safe operation. The pressurizer (PZR) unit is responsible for attaining this task. Thus, the PZR unit's modeling is an important issue for the tracking control purpose and performance analysis concerning the turbine load. This paper develops a mathematical model to accurately predict the PZR unit pressure in the normal and load power changes. The model is a nonequilibrium three region model developed based on the thermodynamics of mass and energy for the water and steam in the PZR. The variations of the pressure and the temperature due to thermodynamic variables (enthalpy, density) and the mechanical work effect are considered in the developed model. The model also addresses other thermodynamic processes such as bulk flashing, spray condensation, interface condensation, wall condensation, and rainout flow. The inlet and outlet flow rates of the primary circuit (PC) and the average temperature and the hot leg temperature are included in the developed model. Based on data generated from the VVER-1200 simulator, parameter estimation, verification, and validation of the developed model through load power changes are achieved. Besides, a comparison with other models given in literature has been performed. This comparison indicated that the developed model is more sensitive against the minimum variation of PZR dynamics. Finally, a closed-loop PZR pressure control system, including a conventional Proportional Integral Derivative (PID) controller, was designed to test the control purpose's developed model. The simulation results over typical load change transients have been demonstrated the feasibility, effectiveness, and accuracy of the developed nonlinear model of PZR for dynamic modeling and control purposes.

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Section: Introductionmentioning

confidence: 99%

“…2-Using the simulator personal computer transient analyzer VVAR −1200 (PCTRAN VVER-1200), the parameter estimation, the validation, and the verifications of the developed model were carried out (Bezrukov et al, 2016;Fyza et al, 2019;Ibrahim et al, 2013;Khan & Islam, 2019;Mollah, 2018).…”

Section: Introductionmentioning

confidence: 99%

A developed analytical model for the pressurizer unit in nuclear power plants

Sheta

Ali

Fikry

et al. 2021

Journal of Radiation Research and Applied Sciences

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“…This tool has been used by many researchers for various accident analyses. Bezrukov et al studied different reactor core re-flooding cases for VVER reactors during a loss of coolant accident [28]. Fyza et al analyzed the thermal hydraulic parameters of a VVER-1200 nuclear power plant during a loss of coolant accident accompanied by loss of external power supply [29].…”

Section: Introductionmentioning

confidence: 99%

A Pctran-Based Investigation on the Effect of Inadvertent Control Rod Withdrawal on the Thermal-Hydraulic Parameters of a Vver-1200 Nuclear Power Reactor

Khan¹,

Islam²

2019

Acta mech. Malays.

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In this work, the effect of inadvertent control rod withdrawal on the thermal-hydraulic parameters of VVER-1200 type nuclear power reactor has been investigated. Personal Computer Transient Analyzer (PCTRAN) has been used to simulate the transient responses associated with the situation. An attempt has been made to find out the maximum amount of positive reactivity that may be inserted instantaneously without causing serious damage to any of the essential reactor components. Four instantaneous positive reactivity insertion case studies have been performed in the range of 2.5-10% Δk/k in order to identify the limiting condition for ensuring safety of the nuclear facility. Results show that for all the cases, prompt criticality is observed due to positive reactivity insertion higher than 0.66% Δk/k. For positive reactivity insertion up to 7.5% Δk/k, the maximum core thermal power is observed to be less than 160% of the rated core thermal power. The peak cladding temperature is observed to be less than 800 o C with no clad failure. However, for 10% Δk/k reactivity insertion, core thermal power has been observed to reach nearly 360% of the rated core thermal power. Also, a 76% probability of cladding failure is recorded due to peak cladding temperature reaching nearly 1400 o C. To identify the maximum limit of instantaneous positive reactivity insertion, a "Factor of Safety" based analysis has been conducted. Results from the analysis reveal that maximum limit is around 8.8% Δk/k. This finding is supported by the results from further simulations.

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