MELCOR severe accident analysis for a natural circulation small modular reactor

2017 ◽  
Vol 100 ◽  
pp. 197-208 ◽  
Author(s):  
Longze Li ◽  
Tae Woon Kim ◽  
Yapei Zhang ◽  
Shripad T. Revankar ◽  
Wenxi Tian ◽  
...  
Keyword(s):  
Natural Circulation ◽  
Severe Accident ◽  
Accident Analysis ◽  
Small Modular Reactor

Sensitivity Analysis of the SBLOCA Induced Severe Accident for a Natural Circulation Small Modular Reactors

2018 ◽  
Author(s):  
Longze Li ◽  
Jue Wang ◽  
Yapei Zhang ◽  
G. H. Su
Keyword(s):  
Sensitivity Analysis ◽  
Heat Sink ◽  
Natural Circulation ◽  
Management Strategies ◽  
Severe Accident ◽  
Severe Accidents ◽  
Small Modular Reactor ◽  
Fraction Increase ◽  
Crack Position

The natural circulation small modular reactor (NCSMR) is a 330 MW reactor which has no reactor coolant pumps (RCP) and no active safety injection systems at all. The reactor is mainly comprised of the reactor pressure vessel (RPV) with integral pressurize r and steam generator. RPV is enclosed by a vacuumed pressure containment vessel (PCV) and the PCV is submerged in the underground containment pool. A MELCOR model and corresponding input deck are developed for the RPV, PCV, and containment pool. The containment pool takes the role of ultimate heat sink (UHS) in accident situations. The containment pool may crack and leak in some critical accidents as the earthquake, leading to the severe accident of the reactor. A TMI-2 like SBLOCA in the RPV (stuck open RVVs) along with the containment pool crack (loss of ultimate heat sink) is simulated in the work. So me key parameters as the RRVs stuck open fraction, the PCV-SRVs open or not, the containment pool crack position would have large influence on the severe accident sequence. The sensitivity of these parameters to the accident sequence is analyzed in the work. According to the simulation results, the RPV pressure decreased with the RRVs stuck open. The depressurization of RPV accelerated with the RPV-SRV open fraction increase. The PCV pressure increased after that. Two cases as the PCV-SRV open after PCV pressure increase to 5 MPa, and PCV break while the RV d id not open, are analysis. The coolant discharge mass flo wrate in RPV and PCV were different in two cases, leading to the different degradation situation of the core. Since the containment pool is so important for the accident mitigation, sensitivity analysis is done for the containment pool crack position in the pool. The work will be meaningful in gaining an insight into the detailed process involved. One of the final goals of this work would be to identify appropriate accident management strategies and countermeasures for the potential extreme natural hazard induced severe accidents during the design process of NCSMR.

scholarly journals RELAP5 SIMULATION FOR SEVERE ACCIDENT ANALYSIS OF RSG-GAS REACTOR

2018 ◽  
Vol 20 (1) ◽  
pp. 23 ◽  
Author(s):  
Andi Sofrany Ekariansyah ◽  
Endiah Puji Hastuti ◽  
Sudarmono Sudarmono
Keyword(s):  
Fuel Element ◽  
Secondary Flow ◽  
Research Reactor ◽  
Natural Circulation ◽  
Safety Analysis ◽  
Reactor Power ◽  
Severe Accident ◽  
Accident Analysis ◽  
Plate Temperature ◽  
The One

The research reactor in the world is to be known safer than power reactor due to its simpler design related to the core and operational chararacteristics. Nevertheless, potential hazards of research reactor to the public and the environment can not be ignored due to several special features. Therefore the level of safety must be clearly demonstrated in the safety analysis report (SAR) using safety analysis, which is performed with various approaches and methods supported by computational tools. The purpose of this research is to simulate several accidents in the Indonesia RSG-GAS reactor, which may lead to the fuel damage, to complement the severe accident analysis results that already described in the SAR. The simulation were performed using the thermal hydraulic code of RELAP5/SCDAP/Mod3.4 which has the capability to model the plate-type of RSG-GAS fuel elements. Three events were simulated, which are loss of primary and secondary flow without reactor trip, blockage of core subchannels without reactor trip during full power, and loss of primary and secondary flow followed by reactor trip and blockage of core subchannel. The first event will harm the fuel plate cladding as showed by its melting temperature of 590 °C. The blockage of one or more subchannels in the one fuel element results in different consequences to the fuel plates, in which at least two blocked subchannels will damage one fuel plate, even more the blockage of one fuel element. The combination of loss of primary and secondary flow followed by reactor trip and blockage of one fuel element has provided an increase of fuel plate temperature below its melting point meaning that the established natural circulation and the relative low reactor power is sufficient to cool the fuel element.Keywords: loss of flow, blockage, fuel plate, RSG-GAS, RELAP5 SIMULASI RELAP5 UNTUK ANALISIS KECELAKAAN PARAH PADA REAKTOR RSG-GAS. Reaktor riset di dunia diketahui lebih aman dari pada reaktor daya karena desainnya yang lebih sederhana pada teras dan karakteristika operasinya. Namun demikian, potensi bahaya reaktor riset terhadap publik dan lingkungan tidak bisa diabaikan karena beberapa fitur tertentu. Oleh karena itu, level keselamatan reaktor riset harus jelas ditunjukkan dalam Laporan Analisis Keselamatan (LAK) dalam bentuk analisis keselamatan yang dilakukan dengan berbagai macam pendekatan dan metode dan didukung dengan alat komputasi. Tujuan penelitian ini adalah untuk mensimulasikan beberapa kecelakaan parah pada reaktor RSG-GAS yang dapat menyebabkan kerusakan bahan bakar untuk memperkuat hasil analisis kecelakaan parah yang sudah ada dalam LAK. Simulation dilakukan dengan program perhitungan RELAP5/SCDAP/Mod3.4 yang memiliki kemampuan untuk memodelkan elemen bahan bakar tipe pelat di RSG-GAS. Tiga kejadian telah disimulasikan yaitu hilangnya aliran primer dan sekunder dengan kegagalan reaktor untuk dipadamkan, tersumbatnya beberapa kanal pendingin bahan bakar pada daya penuh, dan hilangnya aliran primer dan sekunder yang diikuti dengan tersumbatnya beberapa kanal pendingin bahan bakar setelah reaktor padam. Kejadian pertama akan membahayakan pelat bahan bakar dengan naiknya temperatur kelongsong hingga titik lelehnya yaitu 590 °C. Tersumbatnya satu atau beberapa kanal pada satu elemen bahan bakar menyebabkan konsekuensi yang berbeda pada pelat bahan bakar, dimana paling sedikit tersumbatnya 2 kanal akan merusak satu pelat bahan bakar, apalagi tersumbatnya satu elemen bahan bakar. Kombinasi antara hilangnya aliran pendingin primer dan sekunder yang diikuti dengan tersumbatnya satu kanal bahan bakar setelah reaktor dipadamkan menyebabkan naiknya temperatur kelongsong di bawah titik lelehnya yang berarti sirkulasi alam yang terbentuk dan daya yang terus turun cukup untuk mendinginkan elemen bahan bakar.Kata kunci: kehilangan aliran, penyumbatan, pelat bahan bakar, RSG-GAS, RELAP5

Analysis of the natural circulation flow map uncertainties in an integral small modular reactor

2021 ◽  
Vol 378 ◽  
pp. 111156
Author(s):  
Seyed Ali Hosseini ◽  
Reza Akbari ◽  
Amir Saeed Shirani ◽  
Francesco D'Auria
Keyword(s):  
Natural Circulation ◽  
Circulation Flow ◽  
Small Modular Reactor ◽  
Flow Map

scholarly journals Simulation of the Westinghouse AP1000 Response to SBLOCA Using RELAP/SCDAPSIM

2014 ◽  
Vol 2014 ◽  
pp. 1-9 ◽  
Author(s):  
Ayah Elshahat ◽  
Timothy Abram ◽  
Judith Hohorst ◽  
Chris Allison
Keyword(s):  
Driving Forces ◽  
Natural Circulation ◽  
Safety System ◽  
Operating Conditions ◽  
Severe Accident ◽  
Passive Safety ◽  
System Component ◽  
Pressurized Water ◽  
Loss Of Coolant Accident ◽  
Accident Conditions

Great interest is given now to advanced nuclear reactors especially those using passive safety components. The Westinghouse AP1000 Advanced Passive pressurized water reactor (PWR) is an 1117 MWe PWR designed to achieve a high safety and performance record. The AP1000 safety system uses natural driving forces, such as pressurized gas, gravity flow, natural circulation flow, and convection. In this paper, the safety performance of the AP1000 during a small break loss of coolant accident (SBLOCA) is investigated. This was done by modelling the AP1000 and the passive safety systems employed using RELAP/SCDAPSIM code. RELAP/SCDAPSIM is designed to describe the overall reactor coolant system (RCS) thermal hydraulic response and core behaviour under normal operating conditions or under design basis or severe accident conditions. Passive safety components in the AP1000 showed a clear improvement in accident mitigation. It was found that RELAP/SCDAPSIM is capable of modelling a LOCA in an AP1000 and it enables the investigation of each safety system component response separately during the accident. The model is also capable of simulating natural circulation and other relevant phenomena. The results of the model were compared to that of the NOTRUMP code and found to be in a good agreement.

Study on Typical Initial Event of ATWS for Small Modular Reactor

2016 ◽  
Author(s):  
Zhang Dan ◽  
Ran Xu ◽  
Qiu Zhifang ◽  
Zhou Ke ◽  
Feng Li
Keyword(s):  
Optimum Design ◽  
Safety Analysis ◽  
Accident Analysis ◽  
Pressurized Water ◽  
Control Rod ◽  
Reactor Coolant ◽  
Initial Event ◽  
Small Modular Reactor ◽  
Activation System

The method for ATWS (anticipated transient without scram) analysis was completely developed for commercial pressurized water (PWR) reactor plants, especially for selecting of typical initial events. For accident analysis of ATWS, it is different between PWR and small modular reactor (SMR), as different structures and characters, and it is necessary to study the typical initial events for these reactors. Based on the standard of PWR, the demanding for ATWS analysis was studied and the consequences for typical anticipated transient was calculated using RELAP5/MOD3.2 code, “maintain reactor coolant pressure boundary integrity” was selected as limiting criterion. The results shows for SMR, anticipated transient with the most serious consequence for ATWS are loss of offsite power and inadvertent control rod withdraw event, this conclusion will support to prepare the safety analysis report and optimum design of diversity activation system (DAS) for SMR.

scholarly journals Severe Accident Analysis Code SAMPSON Improvement for IMPACT Project.

2002 ◽  
Vol 45 (3) ◽  
pp. 607-614 ◽  
Author(s):  
Hiroshi UJITA ◽  
Takashi IKEDA ◽  
Masanori NAITOH
Keyword(s):  
Severe Accident ◽  
Accident Analysis

Coupling of a Reactor Analysis Code and a Lower Head Thermal Analysis Solver

2019 ◽  
Vol 5 (1) ◽  
Author(s):  
Hiroshi Madokoro ◽  
Alexei Miassoedov ◽  
Thomas Schulenberg
Keyword(s):  
Structural Analysis ◽  
Message Passing ◽  
Message Passing Interface ◽  
Severe Accident ◽  
Coupled Analysis ◽  
Accident Analysis ◽  
Analysis Tool ◽  
Lower Head ◽  
Calculation Results ◽  
Reactor Pressure

Due to the recent high interest on in-vessel melt retention (IVR), development of detailed thermal and structural analysis tool, which can be used in a core-melt severe accident, is inevitable. Although RELAP/SCDAPSIM is a reactor analysis code, originally developed for U.S. NRC, which is still widely used for severe accident analysis, the modeling of the lower head is rather simple, considering only a homogeneous pool. PECM/S, a thermal structural analysis solver for the reactor pressure vessel (RPV) lower head, has a capability of predicting molten pool heat transfer as well as detailed mechanical behavior including creep, plasticity, and material damage. The boundary condition, however, needs to be given manually and thus the application of the stand-alone PECM/S to reactor analyses is limited. By coupling these codes, the strength of both codes can be fully utilized. Coupled analysis is realized through a message passing interface, OpenMPI. The validation simulations have been performed using LIVE test series and the calculation results are compared not only with the measured values but also with the results of stand-alone RELAP/SCDAPSIM simulations.

scholarly journals Perspectives on a Severe Accident Consequences—10 Years after the Fukushima Accident

2021 ◽  
Vol 2 (4) ◽  
pp. 398-411
Author(s):  
Jinho Song
Keyword(s):  
Nuclear Reactor ◽  
Severe Accident ◽  
Accident Analysis ◽  
Fukushima Accident ◽  
The Public ◽  
Analysis Methodology ◽  
International Attention ◽  
Nuclear Reactor Safety ◽  
Term Monitoring

Scientific issues that draw international attention from the public and experts during the last 10 years after the Fukushima accident are discussed. An assessment of current severe accident analysis methodology, impact on the views of nuclear reactor safety, dispute on the safety of fishery products, discharge of radioactive water to the ocean, status of decommissioning, and needs for long-term monitoring of the environment are discussed.

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