Severe accident analysis of the Qinshan Nuclear Power Plant and evaluation of boundary conditions for ex-vessel heat transfer

Hengra, Nasir Hayat; Ilyas, Muhammad; Inayat, Mansoor Hameed

doi:10.1016/j.pnucene.2021.104032

Cited by 5 publications

(2 citation statements)

References 14 publications

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“…Continued advancements in coding and numeric allow this code to run a broader range of transients, including low-pressure transients with non-condensable gases, such as those encountered during mid-loop operations in LWRs, pool-type reactors, or spent fuel storage facilities. [21][22][23] Two-phase thermal-hydraulic models…”

Section: Relap5/scdapmentioning

confidence: 99%

Thermal‐hydraulic investigation of a proposed concept for a passive containment cooling system and evaluation of heat exchanger performance

Ilyas

Hengra

Nauman

et al. 2022

Proceedings of the Institution of Mechanical Engineers, Part A:

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Reduction in ambient pressure within the containment of a light water reactor in the event of a hypothetical Loss of Coolant Accident (LOCA) is crucial for containment integrity. This article introduces the distinct concept of a Passive Containment Cooling System (PCCS). In the proposed PCCS concept, the containment is divided into two compartments with appropriate volumes. An external heat exchanger is connected between the two compartments of the containment for heat transfer and pressure reduction in the case of LOCA. This concept has unique characteristics and provides many advantages over previous designs. A RELAP5 model was developed to perform thermal-hydraulic analysis of the proposed PCCS concept. The performance of the PCCS after its deployment on a small reactor (998.6 MWth) has been assessed under various conditions following LOCA. The results indicate that partitioning of the containment creates a differential pressure that acts as a driving force for flow through the PCCS heat exchanger. Consequently, a decrease in long-term containment pressure is observed. In the case of equal volume fractions in the two containment compartments, the PCCS reduces long-term pressure in the containment by 16.5%. Sensitivity analysis was conducted in which the flow area of the PCCS inlet line, the heat transfer area, and the size of the heat exchanger tank were varied. The PCCS heat transfer capacity increases as these parameters are increased. However, the former is less sensitive than the latter two. The performance of the PCCS heat exchanger under steady-state conditions was validated using RELAP5 simulations, empirical correlations, and analytical models. The evaluation of heat exchanger parameters by various methods (presented in Appendix 1 ) was found in close agreement.

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Section: Relap5/scdapmentioning

confidence: 99%

Thermal‐hydraulic investigation of a proposed concept for a passive containment cooling system and evaluation of heat exchanger performance

Ilyas

Hengra

Nauman

et al. 2022

Proceedings of the Institution of Mechanical Engineers, Part A:

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“…Numerous parameters are monitored in nuclear power plant systems, and the number of alarm variables grows exponentially in the event of a failure, along with the propagation of the failure [4][5]. Therefore, it is often difficult for operators to make accurate judgments about possible faults in nuclear power plants and the effects of faults, which may lead to incorrect fault intervention actions by operators, which in turn may lead to the deterioration of faults and bring about unnecessary reactor shutdown events and even radioactive leakage [6][7].…”

Section: Introductionmentioning

confidence: 99%

ECOC-based integrated learning method for fault diagnosis in nuclear power plants

Sheng,

Mu,

Zhang

2023

Applied Mathematics and Nonlinear Sciences

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The fault diagnosis system of nuclear power plants plays an important role in ensuring the safety and economy of nuclear power plant operations. This paper first analyzes typical faults of nuclear power plants and their phenomena, and fault samples are obtained. A comprehensive study of the structure of the nuclear power plant system, its working mode and the association between each subsystem is carried out to analyze the monitoring parameters and fault characteristics and establish the fault data set. Secondly, an IFWA (Improved Fireworks Algorithm - Integrated Learning) algorithm is proposed to assess the severity of faults in the first circuit of a nuclear power plant. Finally, the fault diagnosis module is divided into three units according to the functional logic, i.e., condition monitoring unit, fault identification unit, and fault severity assessment unit. The results show that the diagnostic accuracy of the IFWA algorithm is 94.25% for SGTR in the single-fault diagnosis experiment and 96.25% for SGTR-LOCA in the multiple-fault diagnosis experiment. It shows that the IFWA algorithm proposed in this paper has the optimal performance capability when applied to nuclear power plant fault diagnosis and effectively assists managers in diagnosing faults and giving maintenance recommendations.

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Assessment of corium concrete interaction and impact of water ingression mechanism using top flooding strategy for a small PWR with MELCOR 2.2

Hengra

Song²,

Kim³

et al. 2023

Annals of Nuclear Energy

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Severe accident analysis of the Qinshan Nuclear Power Plant and evaluation of boundary conditions for ex-vessel heat transfer

Cited by 5 publications

References 14 publications

Thermal‐hydraulic investigation of a proposed concept for a passive containment cooling system and evaluation of heat exchanger performance

Thermal‐hydraulic investigation of a proposed concept for a passive containment cooling system and evaluation of heat exchanger performance

ECOC-based integrated learning method for fault diagnosis in nuclear power plants

Assessment of corium concrete interaction and impact of water ingression mechanism using top flooding strategy for a small PWR with MELCOR 2.2

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