Simulation of the small modular reactor severe accident scenario response to SBO using MELCOR code

2016 ◽  
Vol 86 ◽  
pp. 87-96 ◽  
Author(s):  
Shasha Yin ◽  
Yapei Zhang ◽  
Wenxi Tian ◽  
Suizheng Qiu ◽  
G.H. Su ◽  
...  
Keyword(s):  
Severe Accident ◽  
Accident Scenario ◽  
Small Modular Reactor

Core to surge-line energy transport in a severe accident scenario

1995 ◽  
Vol 154 (2) ◽  
pp. 119-132 ◽  
Author(s):  
M. di Marzo ◽  
K. Almenas ◽  
S. Gopalnarayanan
Keyword(s):  
Energy Transport ◽  
Severe Accident ◽  
Line Energy ◽  
Accident Scenario ◽  
Surge Line

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.

Severe Accident Context Quantification for Long-Term Station Blackout in Boiling Water Reactor Nuclear Power Plants

2018 ◽  
Vol 4 (2) ◽  
Author(s):  
Gueorgui I. Petkov ◽  
Monica Vela-Garcia
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Boiling Water ◽  
Severe Accident ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Accident Scenario ◽  
Station Blackout ◽  
Fukushima Daiichi

The realistic study of dynamic accident context is an invaluable tool to address the uncertainties and their impact on safety assessment and management. The capacities of the performance evaluation of teamwork (PET) procedure for dynamic context quantification and determination of alternatives, coordination, and monitoring of human performance and decision-making are discussed in this paper. The procedure is based on a thorough description of symptoms during the accident scenario progressions with the use of thermo-hydraulic (TH) model and severe accident (SA) codes (melcor and maap). The opportunities of PET procedure for context quantification are exemplified for the long-term station blackout (LT SBO) accident scenario at Fukushima Daiichi #1 and a hypothetic unmitigated LT SBO at peach bottom #1 boiling water reactor (BWR) reactor nuclear power plants (NPPs). The context quantification of these LT SBO scenarios is based on the IAEA Fukushima Daiichi accident report, “State-of-the-Art Reactor Consequence Analysis” and TH calculations made by using maap code at the EC Joint Research Centre.

Wall Materials Effects on Sheltered Indoor Doses From an SMR Hypothetical Accident Release

2021 ◽  
Author(s):  
Yamato Sugitatsu ◽  
Shripad T. Revankar
Keyword(s):  
Emergency Planning ◽  
Severe Accident ◽  
Air Concentration ◽  
Loss Of Coolant Accident ◽  
Energy Needs ◽  
Small Modular Reactor ◽  
Station Blackout ◽  
Wall Materials ◽  
Set Up ◽  
Radiological Dose

Abstract Small modular reactors (SMRs) are expected as a suitable candidate to fulfill energy needs in the future. The regulation of the emergency planning zone (EPZ) has been a controversial issue. The possibility of smaller EPZs because of their small core size and passive safety functions has still under discussion. The major emergency responses to radiological incidents in the early phase are evacuation from the area and sheltering-in-place within a building. Comparison between the dose incurred during evacuation and that with sheltering-in-place is necessary to consider the proper protective actions. This study focuses on effect of wall materials on indoor doses for sheltered population from small modular reactor severe accident. The source term came from loss of coolant accident or station blackout, and the time change of air concentration and the ground deposition data was calculated with RASCAL, a software developed by NRC to provide dose projection around the plant. Then general one-story and two-story houses were set up, and 6 wall materials were selected for calculating indoor doses. Cloudshine and groundshine were calculated with Monte Carlo methods, and the shielding function of each house was evaluated by comparing the indoor dose with outdoor dose. The result will be a basis for calculating the radiological dose for sheltered cases in case of nuclear emergency for SMR, which will be valuable to have a more effective emergency planning.

scholarly journals RESULTS OF A PHENOMENA IDENTIFICATION AND RANKING TABLE (PIRT) EXERCISE FOR A SEVERE ACCIDENT IN A SMALL MODULAR HIGH-TEMPERATURE GAS-COOLED REACTOR

2019 ◽  
Vol 8 (2) ◽  
pp. 159-169
Author(s):  
David William Hummel ◽  
Yu-Shan Chin ◽  
Andrew Prudil ◽  
Anthony Williams ◽  
Eugene Masala ◽  
...  
Keyword(s):  
High Temperature ◽  
Current Knowledge ◽  
Safety Analysis ◽  
Severe Accident ◽  
Specific Interest ◽  
Worst Case ◽  
Severe Accidents ◽  
Small Modular Reactor ◽  
Physical Phenomena ◽  
High Temperature Gas

Canada has attracted specific interest from developers of nonwater-cooled small modular reactor (SMR) technologies, including concepts based on high-temperature gas-cooled reactors (HTGRs). It is anticipated that some research and development (R&D) will be necessary to support safety analysis and licensing of these reactors in Canada. The Phenomena Identification and Ranking Table (PIRT) process is a formalized method in which a panel of experts identifies which physical phenomena are most relevant to the reactor safety analysis and how well understood these phenomena are. The PIRT process is thus a tool to assess current knowledge levels and (or) predictive capabilities of models, thus providing direction to a focused R&D program. This paper summarizes the results of a PIRT process performed by a panel of experts at Canadian Nuclear Laboratories for a limiting or “worst-case” accident scenario at a generic HTGR-type SMR. Suggestions are given regarding the highest priority R&D items to support severe accidents analysis of these reactors.

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