VVER-1000/V446 spent fuel pool risk assessment and support through portable mitigating equipment. (15th June 2021)

• Record Type: Journal Article
• Authors: Afshar, N.; Pirouzmand, A.; Faghihi, F.
• Is Part Of: Annals of nuclear energy. Volume 156(2021)
• Publication Date: 2021-06-15
• Subjects: VVER-1000/V446 nuclear reactor -- Spent fuel pool -- Level 1 PSA -- Fuel damage frequency -- Portable equipment -- Risk assessment; Nuclear energy -- Periodicals; Nuclear engineering -- Periodicals; 621.4805
• DOI: 10.1016/j.anucene.2021.108204
• Languages: English
• ISSNs: 0306-4549

Abstract

Highlights: Level 1 PSA is applied to evaluate the fuel damage frequency in spent fuel pool of BNPP-1. A stress test program is conducted by applying available in-site portable equipment. The fuel damage frequency is calculated in the normal long-term storage and the refueling modes. The results indicate that the most dominant initiating events are the loss of offsite power and loss of service water. Abstract: The Spent Fuel Pool (SFP) accident at the Fukushima Daiichi unit 4 demonstrated the need for identifying and analyzing possible accident sequences and subsequent application of severe accident mitigating systems in LWR (Light Water Reactor) SFPs. In this study, Level 1 Probabilistic Safety Assessment (PSA) is applied to evaluate the Fuel Damage Frequency (FDF) in SFP of Bushehr Nuclear Power Plant unit-1 (BNPP-1). This paper specifically focuses on assessing the SFP risk using Portable Mitigating Equipment (PME) including a reliable robust Portable Diesel Generator (PDG), a Portable Diesel Pump (PDP), and an SFP-spray system proposed to maintain steady decay heat removal. PME is the recommended equipment proposed by IAEA -following the Fukushima disaster- to mitigate accident consequences. In this study, the fault/event trees have been applied to conduct PSA using the SAPHIRE code. The results indicate that the most dominant Initiating Events (IEs) in the reference plant are the Loss of Offsite Power (LOOP) and Loss of Service Water (LOSW), both simultaneously affecting the reactor core and the SFP. There is a significant decrease in calculated FDF – after implementing PDG and PDP- in respect of the reference plant from 3.086E to 6 to 2.166E−7 in the normal storage mode. Heavy load drop, which is a low probability/high consequence IE, becomes crucially important in cases where PME is employed as it directly leads to fuel damage. It is expected that the analyses set forth here would provide efficient guidelines and insights into the strengths and weaknesses inherent in the design and operation of SFP in VVER-1000/V-446 nuclear power plant.