Stepwise integral scaling method for severe accident analysis and its application to corium dispersion in direct containment heating

1994 ◽  
Vol 151 (1) ◽  
pp. 223-234 ◽  
Author(s):  
M. Ishii ◽  
G. Zhang ◽  
H.C. No ◽  
F. Eltwila
Keyword(s):  
Severe Accident ◽  
Accident Analysis ◽  
Scaling Method ◽  
Direct Containment Heating

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.

Numerical simulation of CS28-1 experiment by using CANDU severe accident analysis code, CAISER

2021 ◽  
Vol 150 ◽  
pp. 107820
Author(s):  
Jun Ho Bae ◽  
Dong Gun Son ◽  
Ki Hyun Kim ◽  
Jun Young Kang ◽  
Yong Mann Song ◽  
...  
Keyword(s):  
Numerical Simulation ◽  
Severe Accident ◽  
Accident Analysis

BEPU Approach in the CANDU 6 Severe Accident Analysis

2019 ◽  
Author(s):  
Robert Musoiu ◽  
Roxana-Mihaela Nistor-Vlad ◽  
Ilie Prisecaru ◽  
Chris Allison
Keyword(s):  
Severe Accident ◽  
Accident Analysis

Onset of Gas Blowthrough During Melt Expulsion From a Hole in a RPV: Numerical Analysis Using the SIMMER Code

2006 ◽  
Author(s):  
Frank Kretzschmar
Keyword(s):  
Numerical Analysis ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Severe Accident ◽  
Core Material ◽  
Test Facility ◽  
Computer Codes ◽  
Direct Containment Heating ◽  
Heating Up

In the case of a severe accident in a nuclear power plant there is a residual risk, that the Reactor Pressure Vessel (RPV) does not withstand the thermal attack of the molten core material, of which the temperature can be about 3000 K. For the analysis of the processes governing melt dispersal and heating up of the containment atmosphere of a nuclear power plant in the case of such an event, it is important to know the time of the onset of gas blowthrough during the melt expulsion through the hole in the bottom of the RPV. In the test facility DISCO-C (Dispersion of Simulant Corium-Cold) at the FZK /6/, experiments were performed to furnish data for modeling Direct Containment Heating (DCH) processes in computer codes that will be used to extrapolate these results to the reactor case. DISCO-C models the RPV, the Reactor Coolant System (RCS), cavity and the annular subcompartments of a large European reactor in a scale 1:18. The liquid type, the initial liquid mass, the type of the driving gas and the size of the hole were varied in these experiments. We present results for the onset of the gas blowthrough that were reached by numerical analysis with the Multiphase-Code SIMMER. We compare the results with the experimental results from the DISCO-C experiments and with analytical correlations, given by other authors.

Preliminary Severe Accident Analysis of Fusion-Fission Hybrid Reactor Using the MELCOR Code

2014 ◽  
Author(s):  
Zhang Dabin ◽  
Zhiwei Zhou ◽  
Heng Xie ◽  
Tang Yang
Keyword(s):  
Cooling System ◽  
Safety Performance ◽  
Severe Accident ◽  
Hybrid Reactor ◽  
Accident Analysis ◽  
Analysis Model ◽  
Decay Heat ◽  
Calculation Results ◽  
Core Meltdown ◽  
Dry Out

The fusion-fission hybrid conceptual reactor is a proposed means of generating power, which adopts a water cooled fission blanket based on ITER. Due to the water cooled fission blanket, safety performance of the hybrid reactor should be considered, including decay heat remove, core uncovered, core meltdown, core degradation, radioactivity releases, etc., similar with the PWRs. The main goal of this work is to develop the fission blanket model by using MELCOR code, and to evaluate the safety performance under severe accidents preliminarily. Based on MELCOR 1.8.5, some modification is made for the severe accident analysis of fission blanket. Using modified MELCOR code, an analysis model is developed for the fission blanket and the cooling loop. The strategy of the In-Vessel Retention (IVR) using the ex-vessel cooling method is evaluated during a large break LOCA. The calculation results describes the main phenomena during the severe accident progression, including core dry out, meltdown, relocation, etc.. Simulation result is shown that the decay heat in the fission zone can be removed out by the ex-vessel cooling system merely, and the fuel max temperature will not reach the melting point.

Accident Analysis of Fukushima Daiichi Nuclear Power Plant Unit 2 by the SAMPSON Severe Accident Code

2014 ◽  
Author(s):  
Atsuo Takahashi ◽  
Marco Pellegrini ◽  
Hideo Mizouchi ◽  
Hiroaki Suzuki ◽  
Masanori Naitoh
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Severe Accident ◽  
Accident Analysis ◽  
Unknown Boundary ◽  
Long Time ◽  
Fukushima Daiichi ◽  
Nuclear Power Plant Unit

The transient process of the accident at the Fukushima Daiichi Nuclear Power Plant Unit 2 was analyzed by the severe accident analysis code, SAMPSON. One of the characteristic phenomena in Unit 2 is that the reactor core isolation cooling system (RCIC) worked for an unexpectedly long time (about 70 h) without batteries and consequently core damage was delayed when compared to Units 1 and 3. The mechanism of how the RCIC worked such a long time is thought to be due to balance between injected water from the RCIC pump and the supplied mixture of steam and water sent to the RCIC turbine. To confirm the RCIC working conditions and reproduce the measured plant properties, such as pressure and water level in the pressure vessel, we introduced a two-phase turbine driven pump model into SAMPSON. In the model, mass flow rate of water injected by the RCIC was calculated through turbine efficiency degradation the originated from the mixture of steam and water flowing to the RCIC turbine. To reproduce the drywell pressure, we assumed that the torus room was flooded by the tsunami and heat was removed from the suppression chamber to the sea water. Although uncertainties, mainly regarding behavior of debris, still remain because of unknown boundary conditions, such as alternative water injection by fire trucks, simulation results by SAMPSON agreed well with the measured values for several days after the scram.

Sign in / Sign up

Export Citation Format

Share Document