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Maximizing accuracy in severe accident simulations: An In-depth analysis of sampling methods for melcor code during station black-out in WWER-1000

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NUCLEAR ENGINEERING AND DESIGN 2025年 435卷

作者： Gharari, R. Ahangari, R. Hasanifard, E. Roostaii, B. Nucl Sci & Technol Res Inst Reactor & Nucl Safety Res Sch Tehran Iran Islamic Azad Univ Fac Engn Sci & Res Branch Tehran Iran

This study explores the uncertainty in severe accident analysis of nuclear power plants, focusing on the thermal-hydraulic parameters of the WWER1000 reactor during station black-out scenarios. The primary challenge in such analyses lies in the inherent uncertainties associated with the simplifying assumptions embedded in computational codes. To address this, five prominent sampling methods-simple generation sampling, Latin hypercube sampling, SOBOL, HALTON, and LHS-SOBOL-are applied to improve the melcor code's predictive accuracy. The results highlight that SGS produces non-uniform input distributions and performs poorly in comparison to the other methods. Additionally, key parameters, including hydrogen production, maximum containment pressure, and reactor pressure vessel rupture time, show significant variability, ranging from 150 to 602 kg, 14,100 to 28,000 s, and 0.62 to 1.05 MPa, respectively, with a 95 % confidence level. The findings emphasize the importance of selecting robust sampling techniques for enhancing the reliability of nuclear safety assessments.

关键词： Severe accident analysis Uncertainty analysis melcor code Nuclear power plants Sampling methods

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A sensitivity study on cut frequency of FFTBM for quantitative assessment of thermal-hydraulic modeling by melcor code

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PROGRESS IN NUCLEAR ENERGY 2019年第0期115卷 62-73页

作者： Saghafi, Mandi Univ Bonab Fac Engn Dept Mech Engn Bonab *** East Azerbaijan Iran

In this paper, accuracy sensitivity of melcor code in prediction of Thermal-Hydraulic (TH) behavior of test facilities of Nuclear Power Plants (NPP) is quantitatively analyzed. melcor is a fast-running code with integrated structure, and it is extensively employed for modeling of severe accidents in NPPs. Fast Fourier Transform Based Method (FFTBM) is a tool to determine differences between code predictions and measurements in frequencies. Cut frequency plays an important role in quantitative evaluation of TH calculations by FFTBM. In the current research, a sensitivity study has been carried out by different cut frequencies. Suitable cut frequency removes incorrect influence of higher noise frequencies of investigated signals, while avoiding loss of information in accuracy indices of FFTBM. Determined cut frequencies for both original FFTBM and improved FFTBM by signal mirroring, have a good agreement. Accuracy indices calculated with the determined cut frequency, provide more consistent understanding of calculation accuracy variations during the modeled transients.

关键词： Nuclear power plant Thermal-hydraulics FFTBM melcor code Cut frequency

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The influence of the water ingression and melt eruption model on the melcor code prediction of molten corium-concrete interaction in the APR-140 0 reactor cavity

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NUCLEAR ENGINEERING AND TECHNOLOGY 2022年第4期54卷 1508-1515页

作者： Amidu, Muritala A. Addad, Yacine Khalifa Univ Sci & Technol Emirates Nucl Technol Ctr ENTC Dept Nucl Engn POB 127788 Abu Dhabi U Arab Emirates

In the present study, the cavity module of the melcor code is used for the simulation of molten corium concrete interaction (MCCI) during the late phase of postulated large break loss of coolant (LB-LOCA) accident in the APR1400 reactor design. Using the molten corium composition data from previous melcor Simulation of APR1400 under LB-LOCA accident, the ex-vessel phases of the accident sequences with long-term MCCI are recalculated with stand-alone cavity package of the melcor code to investigate the impact of water ingression and melt eruption models which were hitherto absent in melcor code. Significant changes in the MCCI behaviors in terms of the heat transfer rates, amount of gases released, and maximum cavity ablation depths are observed and reported in this study. Most especially, the incorporation of these models in the new release of melcor code has led to the reduction of the maximum ablation depth in radial and axial directions by-38% and-32%, respectively. These impacts are substantial enough to change the conclusions earlier reached by researchers who had used the older versions of the melcor code for their studies. and it could also impact the estimated cost of the severe accident mitigation system in the APR1400 reactor. (c) 2021 Korean Nuclear Society, Published by Elsevier Korea LLC. This is an open access article under the CC BY-NC-ND license (http://***/licenses/by-nc-nd/4.0/).

关键词： Severe accident MCCI melcor code APR1400 reactors Water ingression model Melt eruption model

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Severe accident analysis of plant-specific spent fuel pool to support a SFP risk and accident management

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ANNALS OF NUCLEAR ENERGY 2016年 89卷 70-83页

作者： Ahn, Kwang-Il Shin, Jae-Uk Kim, Won-Tae Korea Atom Energy Res Inst Integrated Risk Assessment Div Daedeok Daero 989-111 Taejon 305353 South Korea RETech Co LTD Heungdeok IT Valley13Heungdeok 1 Ro Yongin 446908 Gyeonggi Do South Korea

The Fukushima accident that happened on March 11, 2011 has raised the possibility of severe accidents in an SFP under beyond design basis external events (BDBEEs) that might lead to a loss of all safety functions and need a way to cope effectively with such an event through a SFP risk and accident management. From a SFP safety point of view, the low decay heat of the fuel assemblies and large water inventory in an SFP may make the accident processes slow compared to an accident in the reactor core, but a huge number of fuel assemblies stored inside it and no containment in a SFP building (especially for PWR types) might be exposed to much greater risk. A quantitative analysis for SFP accidents can give more insights into which aspects play dominant roles in an accident progression. To date, however, few studies have been made for severe plant-level SFP accidents. This paper provides the plant-specific melcor analysis results for potential severe accidents that can be expected in a typical PWR. The SFP of the reference plant is located adjacent to the primary containment (outside primary containment), and its bottom is placed above the plant grade. Three representative accident scenarios (loss of cooling accident, loss of coolant inventory, and complete loss of coolant accident) and two different reactor operating modes (normal operation and refuelling mode) have been considered for the analysis. From the risk and accident management point of view, phenomenological analysis was mainly focused on (a) the level of the coolant, (b) the cladding temperature, (c) the mass of hydrogen generated during accident progressions, and (d) MCCI (Molten Cerium Concrete Interaction). (C) 2015 Elsevier Ltd. All rights reserved.

关键词： Spent fuel pool Beyond design basis external events (BDBEEs) Risk and accident management Operation mode Severe accident analysis melcor code

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Analysis of loss of cooling and loss of coolant severe accident scenarios in VVER-1000/V446 spent fuel pool

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ANNALS OF NUCLEAR ENERGY 2020年 138卷 107205-107205页

作者： Omidifard, P. Pirouzmand, A. Hadad, K. Sahin, Sumer Shiraz Univ Sch Mech Engn Dept Nucl Engn Shiraz Iran Bahcesehir Univ Fac Engn & Nat Sci Istanbul Turkey

The spent fuel storage system is designed to store and cool the spent fuel within several years with account for scheduled refueling and emergency unloading of the entire reactor core. The Fukushima nuclear disaster in 11 March 2011 showed that the safe storage of spent fuel assemblies in the spent fuel pool (SFP) is very vital since many radioactive fission products could be simply released into the environment if a severe accident occurred in this equipment. In this study, severe accident scenarios at the VVER-1000/V446 spent fuel pool are studied using the melcor 1.8.6 severe accident analysis code. Since the most likely events of the SFP are the loss of cooling and loss of coolant, scenarios leading to such accidents (i.e. six scenarios for different leakage rate and leak elevation) are modeled and investigated. The most important results of this study are the calculation of the occurrence time of the key events, the amount of generated hydrogen that has great importance in development of hydrogen mitigation strategies, and the amount of radioactive materials released into the containment for each scenario. (C) 2019 Elsevier Ltd. All rights reserved.

关键词： Severe accident Spent fuel pool VVER-1000/V446 melcor code

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Assessment of new severe accident mitigation systems on containment pressure of the WWER1000/V446

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ANNALS OF NUCLEAR ENERGY 2020年 148卷 107691-107691页

作者： Gharari, R. Kazeminejad, H. Kojouri, N. Mataji Hedayat, A. Vand, M. Hassan Nucl Sci & Technol Res Inst NSTRI POB 11365-3486 Tehran Iran

Investigation of Severe Accidents (SAs) to manage containment pressure and hydrogen production are the objectives of the present work. A novel SA management (SAM) strategy for containment pressure management is evaluated during Station Black-Out (SBO), Station Black-Out with Large Break Loss of Coolant Accident (SBO-LBLOCA) and Station Black-Out with Small Break Loss of Coolant Accident (SBO-SBLOCA) in the WWER1000/V446 nuclear reactor by tuning the spray parameters using melcor1.8.6 code and CONTAIN2.0 code for validation. Results show that the spray system with 0.44 m(3)/s flow rate, droplet temperature of 55 degrees C and duration about 7955 s associated with 49 Hydrogen Auto-catalyst Recombiners (HARs) can keep the containment integrity against overpressure during the course of the SAs. The peak containment pressure during the SBO, SBO-LBLOCA and SBO-SBLOCA decreased from 0.92, 0.52, 0.66 MPa to 0.43, 0.38, and 0.37 MPa by the proposed mitigation strategy, respectively. (C) 2020 Elsevier Ltd. All rights reserved.

关键词： Severe accident HAR Spray system melcor code CONTAIN code

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Strategic analysis on sizing of flooding valve for successful accident management of small modular reactor

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NUCLEAR ENGINEERING AND TECHNOLOGY 2024年第3期56卷 949-958页

作者： An, Hyo Jun Park, Jae Hyung Song, Chang Hyun Lee, Jeong Ik Kim, Yonghee Kim, Sung Joong Hanyang Univ Dept Nucl Engn 222 Wangsimni Ro Seoul 04763 South Korea Korea Adv Inst Sci & Technol Dept Nucl & Quantum Engn 291 Daehak Ro Daejeon 34141 South Korea Hanyang Univ Inst Nano Sci & Technol 222 Wangsimni Ro Seoul 04763 South Korea

In contrast to all -time flooded small modular reactor (SMR) systems, an in-kind flooding safety system (FSS) has been proposed as a passive safety system applicable to small modular reactors (SMRs) that adopt a metal containment vessel (MCV). Under transient conditions, the FSS can provide emergency cooling to dry reactor cavities and sustain long -term coolability using re-acquired evaporated steam in the reactor building on demand. When designing an FSS, the effect of the flooding flow area is vital as it affects the overall accident sequence and safety. Therefore, in this study, a melcor model of a reference SMR is developed and numerical analysis is performed under postulated accident scenarios. Without flooding, the MCV pressure of the reactor module exceeds the design pressure before core damage. To prevent core damage, an emergency flooding strategy is devised using various flow path parameters and requirements to ensure an adequate emergency coolant supply before the core damage is investigated. The results indicate that a flow area exceeding 0.02 m2 is required in the FSS to prevent MCV overpressure and core damage. This study is the first to report a strategic analysis for appropriately sizing an FSS flooding valve applicable to innovative SMRs.

关键词： Small modular reactor Flooding safety system Accident management Passive safety system melcor code

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Analysis of Dose Rate Around Molten Corium Deposited on the Seabed after a Severe Bottom of the Ship Melt-Through Nuclear Accident 3

Analysis of Dose Rate Around Molten Corium Deposited on the ...

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3rd IEEE International Conference on Computer and Communications (ICCC)

作者： Ouyang, Kehan Chen, Wenzhen He, Zhengyao Naval Univ Engn Dept Nucl Energy Sci & Engn Wuhan Hubei Peoples R China

ISBN: (纸本)9781509063529

After a severe nuclear accident in nuclear powered ship, molten corium may pass through the bottom and enter the sea, which will cause severe marine radioactive consequences. There are few reports about the marine radioactive consequences caused by cooled molten corium deposited on seabed. This paper uses melcor code to establish the model to analyze severe nuclear accident induced by the blackout in ship, calculates the amount of radionuclides in the molten corium entered in sea, uses MCNP5 code to calculate the dose rate in sea water around the sedimentary debris. It is found that (1) on surface of debris, dose rate can reach 4.5x10(6) mSv/h;(2) sea water has a good shielding effect on radiation, there are almost not affected by radiation at 150 cm away from the debris;(3) Within 100 cm around the debris, the dose rate is too high, the survival of marine biological resources received great threat. This research can provide technical support for emergency response after a severe nuclear accident in nuclear powered ship and the protection of marine ecological environment.

关键词： severe nuclear accident molten corium dose rate melcor code MCNP5 code

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melcor Analysis of a SPARC Experiment for Spray-PAR Interaction during a Hydrogen Release

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ENERGIES 2020年第21期13卷 5696页

作者： Kim, Hyoung Tae Kim, Jongtae Korea Atom Energy Res Inst KAERI Accid Monitoring & Mitigat Res Team Daeduk Daero 989-111 Daejeon 34057 South Korea

A series of experiments were performed in the SPARC (spray-aerosol-recombiner-combustion) test facility to simulate a hydrogen mitigation system with the actuation of a PAR (passive auto-catalytic re-combiner) and spray system. In this study, the SPARC-SPRAY-PAR (SSP1) experiment is chosen to benchmark the melcor (a lumped-parameter code for severe accident analysis) predictions against test data. For this purpose, firstly we prepared the base input model of the SPARC test vessel, and tested it by a simple verification problem with well-defined boundary conditions. The implementation of a currently used PAR correlation in melcor is shown to be appropriate for the simulation of a PAR actuation experiment. In an SSP1 experiment, the PAR is reacting with hydrogen, and the spray actuation starts as soon as hydrogen injection is complete. The melcor simulation well predicts the pressure behavior and the gas flow affected by operating both a PAR and spray system. However, the local hydrogen concentration measurement near the inlet nozzle is much higher than the volume average-value by melcor, since high jet flow from the nozzle is dispersed in the corresponding cell volume. The experimental reproduction of the phenomena we expect, or, conversely, the identification of phenomena we do not understand, will continue to support the verification of analytical models using experimental data and to analyze the impact of spray on PAR operations in severe accident conditions.

关键词： severe accident containment hydrogen behavior spray PAR SPARC test facility melcor code

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Sensitivity study of thermal-hydraulic nodalization for melcor simulations of severe accidents in a pressurized water reactor

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ANNALS OF NUCLEAR ENERGY 2022年 166卷

作者： Zhao, Nan Chen, Yangli Ma, Weimin Bechta, Sevostian Royal Inst Technol KTH Roslagstullsbacken 21 S-10691 Stockholm Sweden

The CVH package in the melcor code is responsible for modelling the thermal-hydraulic behavior dur -ing severe accident. This work presents a sensitivity study of thermal-hydraulic nodalization in the core region for melcor simulations of postulated severe accidents in a pressurized water reactor (PWR). Three nodal schemes are developed with a good agreement of steady-state parameters. Two accident sce-narios: loss of coolant accident (LOCA) and station blackout (SBO) are simulated. The analysis is focused on the effect of the control volumes (CVs) in the CVH nodalization on the simulation results of in-vessel accident progression, including the core degradation, hydrogen generation and fission products release, etc. It is found that compared with the coarse CVH nodalization (1 CV), the radial refinement of CVH nodalization (7 CVs) leads to different impacts on accident progression in the two scenarios: faster core relocation and more hydrogen generation is predicted in the LOCA scenario, but it is opposite in the SBO scenario. The finest nodal scheme (49 CVs) with refinement in both radial and axial direction tends to predict an earlier occurrence of cladding rupture, RPV failure and faster core relocation, as well as more hydrogen generation. The CVH refinement makes little difference on the radioactive release. (c) 2021 The Author(s). Published by Elsevier Ltd. This is an open access article under the CC BY license (http://***/licenses/by/4.0/).

关键词： Severe accident Numerical simulation melcor code Sensitivity study

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