Analyses of liquid entrainment through ADS-4 in AP1000 during a typical small break LOCA transient

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

doi:10.1016/j.anucene.2013.05.006

Cited by 16 publications

(2 citation statements)

References 7 publications

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“…Plenty of institutions have contributed a lot on the Design Basis Condition transients with different simulation tools such as ATHLET, RETRAN, TRAC, CATHARE, RELAP5, and TRACE for di?erent reactor types for the variations of reactor thermal hydraulic parameters. A large amount of knowledge about thermohydraulic processes in the reactor cooling systems under different operational statuses and accident conditions has been gained based on a vast number of experimental and analytical studies (Aksan, 2008;Umminger et al, 2010;Kozmenkov and Rohde, 2013;Asmolov et al, 2014;Wang et al, 2015;Bestion, 2017). The recent studies focus more on the performance of passive safety systems and the increased safety margins of newly developed reactors, as well as the code to code comparisons (Nevo et al, 2012;Wang et al, 2012;Hu et al, 2013Hu et al, , 2014Salehi and Jahanfarnia, 2016;Yousif et al, 2017).…”

Section: Simulation Tools Used For Hpr1000mentioning

confidence: 99%

Study on Typical Design Basis Conditions of HPR1000 With Nuclear Safety Analysis Code ATHLET

et al. 2020

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The third-generation nuclear power plant Hua-long Pressurized Reactor (HPR1000) is developed based on the experience of Chinese commercial Nuclear Power Plant (NPP) designs, construction, operation and maintenance. It improves the concept of defense in depth and strengthens severe accident prevention and mitigation strategies. The HPR1000 has implemented a number of active and passive innovative safety systems and accident management procedures for design basis conditions, e.g., the employment of Medium Pressure Rapid Cooldown (MCD) and Atmospheric Steam Dump System (ASDS) for the activation of Middle Head Safety Injection (MHSI), the application of Secondary Passive Residual Heat Removal System (SPRHR) for the residual heat removal. In the article, calculations are carried out for HPR1000 nuclear power plant with nuclear system safety analysis code ATHLET (Analysis of Thermal-Hydraulics of Leaks and Transient) 3.1 (Lerchl et al., 2016). By means of conservative deterministic safety analysis approach, transient analyses concerning selected typical design basis conditions, i.e., Large Break Loss-Of-Coolant Accident (LB-LOCA), Small Break Loss-Of-Coolant Accident (SB-LOCA), Steam Generator Tube Rupture accident (SGTR), and Feed water Line Break (FLB) are performed. The ATHLET results are also compared with the results performed by CGN-CNPTRI (China General Nuclear-China Nuclear Power Technology Research Institute) with their own code LOCUST with similar assumptions. The comparisons indicate that, although some discrepancies are detected, the trends of system responses predicted by the two codes are generally in agreement with each other for different accident scenarios. The results also demonstrate that the acceptance criteria for each accident can be met with significant safety margin. Thus, the effectiveness of safety system configuration and accident management procedures is guaranteed.

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Section: Simulation Tools Used For Hpr1000mentioning

confidence: 99%

Study on Typical Design Basis Conditions of HPR1000 With Nuclear Safety Analysis Code ATHLET

et al. 2020

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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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Experimental investigation of liquid entrainment in vertical up tee branch with double-end gas inlets

Sun

Zhang

Xiang

et al. 2015

Annals of Nuclear Energy

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Analyses of liquid entrainment through ADS-4 in AP1000 during a typical small break LOCA transient

Cited by 16 publications

References 7 publications

Study on Typical Design Basis Conditions of HPR1000 With Nuclear Safety Analysis Code ATHLET

Study on Typical Design Basis Conditions of HPR1000 With Nuclear Safety Analysis Code ATHLET

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

Experimental investigation of liquid entrainment in vertical up tee branch with double-end gas inlets

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