Severe accident analysis for a typical PWR using the MELCOR code

Li, Longze; Chen, Yongzheng; Tian, Wenxi; Su, G.H.; Qiu, Suizheng

doi:10.1016/j.pnucene.2013.10.014

Cited by 46 publications

(3 citation statements)

References 7 publications

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“…MELCOR and MAAP were used to compare partial results. Li 11 used MELCOR code to simulate a severe accident at a 1000‐MW PWR, analyzed the influence of three conditions on a station blackout, and compared the results. Moon 12 used MELCOR to conduct hydrogen safety studies on the reactor as the reactor hydrogen explosion damaged the integrity of the containment and resulting in a large amount of radionuclide leakage.…”

Section: Introductionmentioning

confidence: 99%

Effect analysis of break size on source term release and radioactive consequences of marine nuclear reactor during loss of coolant accident

Zhao

Zou²,

et al. 2022

Intl J of Energy Research

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Summary A simulation calculation model in the case of a severe accident such as a loss of coolant accident (LOCA) was established and calculated by MELCOR code in a marine pressurized water reactor. The effects of severe accidents on the hydrogen source term, release of the radionuclide source term, and the accident process were investigated, and the radiation consequences were analyzed. LOCA sensitivity analysis was conducted with different break sizes; the release and diffusion of source terms were analyzed. The results indicated that the hydrogen yield in the reactor core was dependent on the temperature, and on the residual water in the reactor core, which has no direct relationship with the break size. The break size directly affected the accident progression. However, when the break exceeded a certain size, it did not affect the LOCA process. The break size may not change to change the release amount and migration rule of radionuclides such as Xe and I, and the fission product CsI, as final containment is in dynamic equilibrium in LOCA conditions. The contents of fission products released into the containment vessel, major loop, cavity, and the environment are sensitive to the break size. However, no special correlation is observed between the amount released and the break size. Highlights The hydrogen yield in the reactor core was dependent on the temperature, and on the residual water in the reactor core, which has no direct relationship with the break size. The break size may not change to change the release amount and migration rule of radionuclides such as Xe, I, and CsI, as final containment is in dynamic equilibrium in LOCA conditions. No particular correlation is observed between the amount of radionuclide released and the break size.

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Section: Introductionmentioning

confidence: 99%

Effect analysis of break size on source term release and radioactive consequences of marine nuclear reactor during loss of coolant accident

Zhao

Zou²,

et al. 2022

Intl J of Energy Research

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“…For improving the NPPs safety, the study of severe accidents has become increasingly important [1]. The normal operation and accident in nuclear reactor has been modeled and studied with different codes along of the years [2][3][4][5]. The reference [6] demonstrated the importance of the selfinitialization of the steady-state on the simulations of transient responses and accidents sequences timing in a PWR modeling on MELCOR.…”

Section: Introductionmentioning

confidence: 99%

MELCOR steady state calculation of the generic PWR of 40MWth

2021

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After the two most significant nuclear accidents in history – the Chernobyl Reactor Four explosion in Ukraine(1986) and the Fukushima Daiichi accident in Japan (2011) –, the Final Safety Analysis Report (FSAR) included a new chapter (19) dedicated to the Probabilistic Safety Assessment (PSA) and Severe Accident Analysis (SAA), covering accidents with core melting. FSAR is the most important document for licensing of siting, construction, commissioning and operation of a nuclear power plant. In the USA, the elaboration of the FSAR chapter 19 is according to the review and acceptance criteria described in the NUREG-0800 and U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.200. The same approach is being adopted in Brazil by National Nuclear Energy Commission (CNEN). Therefore, the FSAR elaboration requires a detailed knowledge of severe accident phenomena and an analysis of the design vulnerabilities to the severe accidents, as provided in a PSA – e.g., the identification of the initiating events involving significant Core Damage Frequency (CDF) are made in the PSA Level 1. As part of the design and certification activities of a plant of reference, the Laboratory of Risk Analysis, Evaluating and Management (LabRisco), located in the University of São Paulo (USP), Brazil, has been preparing a group of specialists to model the progression of severe accidents in Pressurized Water Reactors (PWR), to support the CNEN regulatory expectation – since Brazilian Nuclear Power Plants (NPP), i.e., Angra 1, 2 and 3, have PWR type, the efforts of the CNEN are concentrated on accidents at this type of reactor. The initial investigation objectives were on completing the detailed input data for a PWR cooling system model using the U.S. NRC MELCOR 2.2 code, and on the study of the reference plant equipment behavior – by comparing this model results and the reference plant normal operation main parameters, as modeled with RELAP5/MOD2 code.

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“…However, in the third generation of an advanced pressurized-water reactor (AP) series NPP, the spray system was removed and a passive containment cooling system (PCCS) was used to remove erosol (Fu et al, 2017). Simultaneously, in the event of station blackout (SBO), the nuclear power plant loses off-site power and this can lead to a severe accident (Li et al, 2014). If that happens, the concentration and distribution of erosols in containment are mainly dependent on natural deposition.…”

Section: Introductionmentioning

confidence: 99%

Experimental Study on the Natural Deposition Characteristics of Erosols in Containment

Tian¹,

et al. 2020

Front. Energy Res.

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Against the background of the natural deposition of erosols in containment under severe accident conditions, a relevant experimental facility was designed and constructed to study the decay constant of the natural deposition of erosols. TiO2 was used as the erosol medium, and the decay constant of erosols was studied. The results show that when erosol concentration is < 15,000/cm3, the decay constant of erosols increases with the increase of the particle size. However, when it is ≥ 15,000/cm3, because the agglomeration of small particles into larger particles accelerates the deposition of particles with smaller size, the decay constant of erosols decreases first and then increases with the increase of the particle size. In addition, the effect of electrostatic on the decay constant of the natural deposition of erosols was studied. When the electrostatic effect is present, the decay constant of erosols increases significantly. Finally, the natural deposition characteristics of erosols were obtained.

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Severe accident analysis for a typical PWR using the MELCOR code

Cited by 46 publications

References 7 publications

Effect analysis of break size on source term release and radioactive consequences of marine nuclear reactor during loss of coolant accident

Effect analysis of break size on source term release and radioactive consequences of marine nuclear reactor during loss of coolant accident

MELCOR steady state calculation of the generic PWR of 40MWth

Experimental Study on the Natural Deposition Characteristics of Erosols in Containment

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