• Nuclear Engineering and Technology
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• v.11 no.3
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• pp.195-202
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• 1979

In this paper, a cause-consequence chart is applied to evaluate the probability that the containment spray system in a nuclear power plant may not be woring properly, given a demand for spryaing at loss of coolant accident (LOCA). It is shown how the diagram provides a basis for calculating two probability measures for malfunctioning of this system, in case the test policy of the system is taken into account, i.e., average probability that the containment spray cannot be established, and average probability that the containment spray is established : spray stops before the required operating time is over.

• Nuclear Engineering and Technology
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• v.48 no.5
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• pp.1140-1153
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• 2016

Since the inception of nuclear power as a commercial energy source, safety has been recognized as a prime consideration in the design, construction, operation, maintenance, and decommissioning of nuclear power plants. The release of radioactivity to the environment requires the failure of multiple safety systems and the breach of three physical barriers: fuel cladding, the reactor cooling system, and containment. In this study, nuclear reactor containment pressurization has been modeled in a large break-loss of coolant accident (LB-LOCA) by programming single-cell and multicell models in MATLAB. First, containment has been considered as a control volume (single-cell model). In addition, spray operation has been added to this model. In the second step, the single-cell model has been developed into a multicell model to consider the effects of the nodalization and spatial location of cells in the containment pressurization in comparison with the single-cell model. In the third step, the accident has been simulated using the CONTAIN 2.0 code. Finally, Bushehr nuclear power plant (BNPP) containment has been considered as a case study. The results of BNPP containment pressurization due to LB-LOCA have been compared between models, final safety analysis report, and CONTAIN code's results.

• Nuclear Engineering and Technology
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• v.15 no.3
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• pp.179-187
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• 1983

The pH of containment spray and sump water following a LOCA for KNU 5'||'&'||'6 and KNU 1 was calculated to see if pH design criteria of containment spray system established by USNRC were met. The pH calculations have been made for the two cases; maximum pH and minimum pH. For KNU 5'||'&'||'6, results showed that long term sump pH values calculated for the maximum pH and minimum pH case well met the pH requirement of at least 8.5 and spray pH for the maximum case slightly exceeded the range of design criteria (8.5 to 11.0). For KNU 1, pH requirement of long term sump pH was also met, however, spray pH value for the maximum pH case was very largely greater than that of current pH requirement. (No pH requirement of containment spray water has been established at the time of designing KNU 1) In order to find the design parameters of containment spray system which are expected to meet the spray pH requirement, several calculations were wade, by changing the input parameters to "LCCAPH". Finally, it was shown that the boric acid concentration in RWST (refueling water storage tank), which was the primary sources of containment spray water during injection mode, be maintained the range of 2750 ppm to 2850 ppm, or tile flow rate of NaOH added to spray water he kept between 10 gpm to 24 gpm.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2477-2487
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• 2021

To improve safety analysis technology for a nuclear reactor containment considering an interaction between a reactor coolant system (RCS) and containment, this study aims at an experimental investigation on the integrated simulation of the RCS and containment, with an integral effect test facility, ATLAS-CUBE. For a realistic simulation of a pressure and temperature (P/T) transient, the containment simulation vessel was designed to preserve a volumetric scale equivalently to the RCS volume scale of ATLAS. Three test cases for a steam line break (SLB) transient were conducted with variation of the initial condition of the passive heat sink or the steam flow direction. The test results indicated a stratified behavior of the steam-gas mixture in the containment following a high-temperature steam injection in prior to the spray injection. The test case with a reduced heat transfer on the passive heat sink showed a faster increase of the P/T inside the containment. The effect of the steam flow direction was also investigated with respect to a multi-dimensional distribution of the local heat transfer on the passive heat sink. The integral effect test data obtained in this study will contribute to validating the evaluation methodology for mass and energy (M/E) and P/T transient of the containment.

• Journal of ILASS-Korea
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• v.4 no.3
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• pp.42-52
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• 1999

An effectiveness of thermal utilization of a dousing system in the 600 MW PHWR Nuclear Power Plant has been evaluated. The behavior and conditions of water droplet sprayed in a postulated accident conditions in containment configuration has been calculated. In this calculation, two pressure conditions with the consideration of obstruction area and containment wall effect has been established : one being the minimum containment pressure of 7 kPa(g) encountered for dousing shut off and the other being the containment design pressure 124 kPa(g). The results revealed that the effectiveness of the thermal utilization ranges from 93% to 97%. In the analysis on two cases without/with side wall effect in the containment building, the thermal utilization decreases with obstruction area from 89% to 85%, which satisfies the design criteria set for the containment pressure against the accident condition.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.975-981
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• 2016

The containment response during the first 24 hours of a low-pressure severe accident scenario in a nuclear power plant with a two-loop Westinghouse-type pressurized water reactor was simulated with the CONTAIN 2.0 computer code. The accident considered in this study is a large-break loss-of-coolant accident, which is not successfully mitigated by the action of safety systems. The analysis includes pressure and temperature responses, as well as investigation into the influence of spray on the retention of fission products and the prevention of hydrogen combustion in the containment.

• Proceedings of the Korean Nuclear Society Conference
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• 2003.10a
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• pp.165.2-165.2
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• 2003

• Journal of Energy Engineering
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• v.24 no.3
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• pp.96-108
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• 2015

A passive containment cooling system has been designed to remove the heat inside a containment during accidents without external power supply. In this work, the PCCS was introduced in the APR1400 plant to replace the containment spray system and, then, the thermal-hydraulic performance of the PCCS was analyzed using the system thermal-hydraulic computer code, MARS. A double-ended cold-leg break accident, which is known to induce the maximum pressure in the containment, is simulated, where the thermal hydraulics of the PCCS, the reactor coolant system, and the containment are simultaneously simulated. The results of the calculations showed that the PCCS can replace the existing spray system and that the containment building and its internal structure also play a very important role for the heat removal during the accident. Some sensitivity calculations were carried out to evaluate the model uncertainty and the effects of design parameters. The limitations of the PCCS are also discussed.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.2960-2973
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• 2022

Since Fukushima nuclear power plant (NPP) accident in 2011, the importance of research on various severe accident phenomena has been emphasized. Particularly, detailed analysis of combustion risk is necessary following the containment damage caused by combustion in the Fukushima accident. Many studies have been conducted to evaluate the risk of local hydrogen concentration increases and flame propagation using computational code. In particular, the potential for combustion by local hydrogen concentration in specific areas within the containment has been emphasized. In this study, the process of flame propagation generated inside a reactor core to containment during a loss of coolant accident (LOCA) was analyzed using MELCOR 2.1 code. Later in the LOCA scenario, it was expected that hydrogen combustion occurred inside the reactor core owing to oxygen inflow through the cold leg break area. The main driving force of the oxygen intrusion is the elevated containment pressure due to the molten corium-concrete interaction. The thermal and mechanical loads caused by the flame threaten the integrity of the containment. Additionally, the containment spray system effectiveness in this situation was evaluated because changes in pressure gradient and concentrations of flammable gases greatly affect the overall behavior of ignition and subsequent containment integrity.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.695-700
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• 1997

To enhance the safety of nuclear power plants, implementation of accident management has been suggested as one of most important programs. Specially, accident management strategies are suggested as one of key elements considered in development of the accident management program. In this study, generally applicable accident management strategies to domestic nuclear power plants are identified through reviewing several accident management programs for the other countries and considering domestic conditions. Identified strategies are as follows; 1) Injection into the Reactor Coolant System, 2) Depressurize the Reactor Coolant System, 3) Depressurize the Steam Generator, 4) Injection into the Steam Generator, 5) Injection into the Containment, 6) Spray into the Containment, 7) Control Hydrogen in the Containment. In addition, the systems and instrumentation necessary for the implementation of .each strategy are also investigated.