Research on thermal hydraulic behavior of small-break LOCAs in AP1000

Wang, W.W.; Su, G.H.; Tian, Wenxi; Qiu, Suizheng

doi:10.1016/j.nucengdes.2013.06.004

Cited by 24 publications

(4 citation statements)

References 25 publications

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“…In the beginning, the NOTRUMP flowrate appears higher that the RELAP5, which might be caused by the differences in the loss coefficient of the break model. At the start of DVI break, only liquid flow discharges from the downcomer occurs, to be followed by steam flow as the void fraction in the core increases [18], as shown by NOTRUMP calculation results. That phenomenon (differences in the liquid and vapor phase of break flow) can not be shown in the present RELAP5 model.…”

Section: Transient Simulationmentioning

confidence: 78%

Performance Analysis of AP1000 Passive Systems during Direct Vessel Injection (DVI) Line Break

Ekariansyah

Widodo

2016

Atom Indo.

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Generation II Nuclear Power Plants (NPPs) have a design weakness as shown by the Fukushima accident. Therefore, Generation III+ NPPs are developed with focus on improvements of fuel technology and thermal efficiency, standardized design, and the use of passive safety system. One type of Generation III+ NPP is the AP1000 that is a pressurized water reactor (PWR) type that has received the final design acceptance from US-NRC and is already under construction at several sites in China as of 2015. The aim of this study is to investigate the behavior and performance of the passive safety system in the AP1000 and to verify the safety margin during the direct vessel injection (DVI) line break as selected event. This event was simulated using RELAP5/SCDAP/Mod3.4 as a best-estimate code developed for transient simulation of light water reactors during postulated accidents. This event is also described in the AP1000 design control document as one of several postulated accidents simulated using the NOTRUMP code. The results obtained from RELAP5 calculation was then compared with the results of simulations using the NOTRUMP code. The results show relatively good agreements in terms of time sequences and characteristics of some injected flow from the passive safety system. The simulation results show that the break of one of the two available DVI lines can be mitigated by the injected coolant flowing, which is operated effectively by gravity and density difference in the cooling system and does not lead to core uncovery. Despite the substantial effort to obtain an apropriate AP1000 model due to lack of detailed geometrical data, the present model can be used as a platform model for other initiating event considered in the AP1000 accident analysis.

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Section: Transient Simulationmentioning

confidence: 78%

Performance Analysis of AP1000 Passive Systems during Direct Vessel Injection (DVI) Line Break

Ekariansyah

Widodo

2016

Atom Indo.

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“…In the natural circulation phase, the prediction results of development model 43 are the lowest, and the prediction result of development model 50 is slightly higher than the base model. Compared with the results of other researchers, 26,27 the activation of the development model 43 may cause deviations for the prediction of critical mass flow in the natural circulation phase. In the later phase of the accident, the activation of development model 50 may produce unreasonable oscillation.…”

Section: Model Application Analysismentioning

confidence: 86%

Best estimate plus uncertainty analysis of a small‐break LOCA on an advanced Generation‐III pressurized water reactor

Yang

Deng

2020

Int J Energy Res

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“…e typical advanced passive plant has been originated from AP600/AP1000 which adopts PCCS design [8][9][10][11][12][13]. Wang et al simulated and analyzed cold leg (CL) small-break LOCAs for AP1000 using RELAP5/MOD3.4 codes, including 2-inch, 4-inch, and 8-inch breaks, and the peak cladding temperatures during the process could meet the safety criterion [14][15][16][17]. Di Giuli et al modeled all the PCCS for the AP1000 and studied a 10-inch small-break LOCA scenario using CESAR codes; when compared to the Westinghouse-developed NOTRUMP code as reference showing a reasonable agreement [18].…”

Section: Introductionmentioning

confidence: 99%

RELAP5 Foresight Thermal-Hydraulic Analysis of Hypothesis Passive Safety Injection System under LOCA for an Existing NPP in China

Sun

Wang

et al. 2020

Science and Technology of Nuclear Installations

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Qinshan nuclear power plant is the first large Chinese-designed nuclear power station based on pressurized water reactor, and the second generation main stream active safety injection system is adopted for Qinshan nuclear power plant. In this paper, a novel passive safety injection system (PSIS) has been proposed for ocean-based Qinshan Phase One nuclear power plant to replace the original active one. The PSIS contains high-pressure, medium-pressure, and lower-pressure safety injection systems, and a two-stage automatic depressurization system. To evaluate the system performance, small-break LOCA has been investigated using RELAP5. Various break sizes and locations including 2-inch, 10-inch cold leg break, and double-ended direct vessel injection line break were analyzed. Key safety parameters such as safe injection mass flow rates, coolant level of the core, system pressure, and fuel cladding temperature were monitored during the accident process. The results illustrate that the performance of the safety injection system can guarantee the effective core cooling and submerged under different LOCA even with only half of the safety injection system, which can fulfill the single failure criteria. The thermal-hydraulic analysis for the Qinshan passive safety injection system is significant to master the related technologies for Chinese engineer and develop the Chinese-designed third-generation nuclear power plants, and the PSIS can guarantee the reactor submerged under LOCA even plus the station block out accident.

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Research on thermal hydraulic behavior of small-break LOCAs in AP1000

Cited by 24 publications

References 25 publications

Performance Analysis of AP1000 Passive Systems during Direct Vessel Injection (DVI) Line Break

Performance Analysis of AP1000 Passive Systems during Direct Vessel Injection (DVI) Line Break

Best estimate plus uncertainty analysis of a small‐break LOCA on an advanced Generation‐III pressurized water reactor

RELAP5 Foresight Thermal-Hydraulic Analysis of Hypothesis Passive Safety Injection System under LOCA for an Existing NPP in China

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