MELCOR 2.2-ASTEC V2.2 crosswalk study reproducing SBLOCA and CSBO scenarios in a PWR1000-like reactor part I: Analysis of RCS thermal-hydraulics and in-vessel phenomena

Giuli, M. Di; Malicki, Mateusz; Dejardin, P.; Corda, V.; Swaidan, Ali

doi:10.1016/j.nucengdes.2021.111248

Cited by 5 publications

(3 citation statements)

References 7 publications

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“…Other plant details were also modelled, including the mai steam line with safety valves, the pressurizer with safety valves and accident valves, th The COR package model has nineteen axial levels and six rings. The active core region has twelve axial levels (7)(8)(9)(10)(11)(12)(13)(14)(15)(16)(17)(18) and one non-active fuel level above and below the active core (6 and 19). The core support plate is modelled by level 5, and the levels below are part of the lower plenum.…”

Section: Gen-iii Pwr Modelmentioning

confidence: 99%

“…After the Fukushima Daiichi accident, much attention was paid to study the hydrogen production process in all LWR reactors. Recently, several studies were published about the in-vessel progression and the hydrogen production during severe accidents, including analytical case studies, uncertainty studies, and sensitivity studies or crosswalks with different severe accident codes, e.g., [5][6][7][8][9][10][11][12][13]. A single severe accident integral code run provides deterministic results, which has to be treated with a large margin of confidence.…”

Section: Introduction 1hydrogen During a Severe Accidentmentioning

confidence: 99%

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Uncertainty and Sensitivity Analysis of the In-Vessel Hydrogen Generation for Gen-III PWR and Phebus FPT-1 with MELCOR 2.2

2021

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In this study, uncertainty and sensitivity analyses were performed with MELCOR 2.2.18 to study the hydrogen generation (figure-of-merit (FoM)) during the in-vessel phase of a severe accident in a light water reactor. The focus of this work was laid on a large generation-III pressurized water reactor (PWR) and a double-ended hot leg (HL) large break loss of coolant accident (LB-LOCA) without a safety injection (SI). The FPT-1 Phebus integral experiment emulating LOCA was studied, where the experiment outcomes were applied for the plant scale modelling. The best estimate calculations were supplemented with an uncertainty analysis (UA) based on 400 input-decks and Latin hypercube sampling (LHS). Additionally, the sensitivity analysis (SA) utilizing the linear regression and linear and rank correlation coefficients was performed. The study was prepared with a new open-source MELCOR sensitivity and uncertainty tool (MelSUA), which was supplemented with this work. The FPT-1 best-estimate model results were within the 10% experimental uncertainty band for the final FoM. It was shown that the hydrogen generation uncertainties in PWR were similar to the FPT-1, with the 95% percentile being covered inside a ~50% band and the 50% percentile inside a ~25% band around the FoM median. Two different power profiles for PWR were compared, indicating its impact on the uncertainty but also on the sensitivity results. Despite a similar setup, different uncertainty parameters impacted FoM, showing the difference between scales but also a significant impact of boundary conditions on the sensitivity analysis.

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Section: Gen-iii Pwr Modelmentioning

confidence: 99%

Section: Introduction 1hydrogen During a Severe Accidentmentioning

confidence: 99%

Uncertainty and Sensitivity Analysis of the In-Vessel Hydrogen Generation for Gen-III PWR and Phebus FPT-1 with MELCOR 2.2

2021

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“…The system included a reactor facility, a major loop, a reactor compartment and system, and the necessary control signals. Giuli 7 simulated thermal‐hydraulic phenomena of a reactor coolant system following a small‐break LOCA and complete station blackout in a pressurized water reactor (PWR) in Belgium using MELCOR and ASTEC to evaluate the impact on the management strategy for a severe reactor accident. Albright 8 studied the material interaction after a severe accident with a station blackout of a boiling water reactor using MELCOR to conduct an uncertainty analysis based on the established model.…”

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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Study of the DVI-LOCA in the AP1000-like reactor with MELCOR code

Włostowski,

Darnowski

2024

Annals of Nuclear Energy

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MELCOR 2.2-ASTEC V2.2 crosswalk study reproducing SBLOCA and CSBO scenarios in a PWR1000-like reactor part I: Analysis of RCS thermal-hydraulics and in-vessel phenomena

Cited by 5 publications

References 7 publications

Uncertainty and Sensitivity Analysis of the In-Vessel Hydrogen Generation for Gen-III PWR and Phebus FPT-1 with MELCOR 2.2

Uncertainty and Sensitivity Analysis of the In-Vessel Hydrogen Generation for Gen-III PWR and Phebus FPT-1 with MELCOR 2.2

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

Study of the DVI-LOCA in the AP1000-like reactor with MELCOR code

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