Simulation of a control rod ejection accident in a VVER-1000/V446 using RELAP5/Mod3.2

Tabadar, Z.; Hadad, Kamal; Nematollahi, Mohammadreza; Jabbari, Masoud; Khaleghi, M.; Hashemi-Tilehnoee, M.

doi:10.1016/j.anucene.2012.02.018

Cited by 36 publications

(4 citation statements)

References 4 publications

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“…At present the accidents for VVER-1000 reactors are thoroughly investigated where the researchers investigated the basis design improvement for VVER-1000 reactors [5][6][7][8][9][10][11][12][13][14][15]. In reviewing the literature, RELAP5 loop code found a validated tool for thermal hydraulic analysis of nuclear reactors [16,17].…”

Section: Introductionmentioning

confidence: 99%

Thermal-hydraulic analysis of VVER-1000 residual heat removal system using RELAP5 code, an evaluation at the boundary of reactor repair mode

Tabadar

Jabbari

Khaleghi

et al. 2018

Alexandria Engineering Journal

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Removing the residual heat from a nuclear reactor is an important safety aspect of thermal hydraulic analysis. In this study, a typical VVER-1000 reactor residual heat removal system has been evaluated using RELAP5 thermal hydraulic loop code during cool-down. Reactor cooling down starts from hot state temperature and then continues to the cool-down stages with 130°C and 70°C, respectively. The second stage of cooling down is the boundary of the reactor repair condition. Main cooling pump head, steam generator (SG) water level, system pressure, the level of coolant in the pressurizer (PRZ), and the temperature of fuel element are examined in a steady state condition. PRZ level, primary circuit and secondary circuit pressure/temperature, and SG water level are evaluated during 32,000 s after cool-down scenario. By comparison, it concluded that the results of RELAP5 code are in agreement with plant experimental data and final safety analysis report (FSAR). Thus, it is proved that the studied reactor is capable to remove the residual heat generated during shutdown. Moreover, RELAP5 is properly recommended for analysis of the VVER-1000 pressurized water reactor during cool-down. Ó 2017 Faculty of Engineering, Alexandria University. Production and hosting by Elsevier B.V. This is an open access article under the CC BY-NC-ND license (http://creativecommons.org/licenses/by-nc-nd/4.0/).

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Section: Introductionmentioning

confidence: 99%

Thermal-hydraulic analysis of VVER-1000 residual heat removal system using RELAP5 code, an evaluation at the boundary of reactor repair mode

Tabadar

Jabbari

Khaleghi

et al. 2018

Alexandria Engineering Journal

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“…OTSG can be modeled in different ways. When OTSG is modeled by using existing system codes, such as RELAP5, which use fixed boundaries, sometimes they allow errors within one control volume range on the secondary side boundaries [4][5]. However, researchers extensively analyzed U-tube steam generators by using RELAP5 [6].…”

Section: mentioning

confidence: 99%

Thermal Hydraulic Modeling of Once-Through Steam Generator by Two-Fluid U-Tube Steam Generator Code

et al. 2017

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The THERMIT U-tube steam generator (THERMIT-UTSG) code was used for evaluation for the parametric study of a scaled once-through pressurized water reactor steam generator (OTSG) made by Babcock & Wilcox. The results of the code were compared to the experimental data of the 19-tube OTSG and a simple heat transfer code that was developed by Osakabe. The main calculated thermodynamic parameters were primary-secondary fluid temperatures, tube wall internal and external temperatures that were subjected to primary and the secondary fluid, and the secondary fluid vapor quality. The assessed code can be used for modeling the OTSGs with some modification. The results of THERMIT-UTSG were in agreement with the experimental results and the prediction of Osakabe's numerical model.

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“…For a majority of these studies, Monte Carlo codes (e.g. MCNP (Erfani Nia et al, 2012;Tabadar et al, 2012, Hadad et al 2008), Discrete Ordinate codes (e.g. WIMSD-4 (Hadad and Ayobian, 2006;Hadad et al, 2009)) and Multi-Group Diffusion codes (e.g.…”

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confidence: 99%

PARCS cross-section library generator; part one: Development and verification

Porhemmat

Hadad

Faghihi

2015

Progress in Nuclear Energy

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Simulation of a control rod ejection accident in a VVER-1000/V446 using RELAP5/Mod3.2

Cited by 36 publications

References 4 publications

Thermal-hydraulic analysis of VVER-1000 residual heat removal system using RELAP5 code, an evaluation at the boundary of reactor repair mode

Thermal-hydraulic analysis of VVER-1000 residual heat removal system using RELAP5 code, an evaluation at the boundary of reactor repair mode

Thermal Hydraulic Modeling of Once-Through Steam Generator by Two-Fluid U-Tube Steam Generator Code

PARCS cross-section library generator; part one: Development and verification

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