Deposition Fractions of Fission Product and Heavy Elements on Soil in Fukushima Dai-Ichi Nuclear Power Plant Accident

2013 ◽  
Author(s):  
Toru Yamamoto
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Soil Samples ◽  
Heavy Elements ◽  
Severe Accident ◽  
Accident Analysis ◽  
Radioactivity Measurement ◽  
Power Station ◽  
The Core

Based on radioactivity measurement of soil samples in the site of Fukushima Dai-Ichi Nuclear Power Station, radioactivity of Sr, Nb, Mo, Tc, Ru, Ag, Te, I, Cs, Ba, La, Pu, Am, and Cm isotopes were compiled as radioactivity ratios to 137Cs. By exponentially fitting or averaging, the radioactivity ratios at the core shutdown were estimated. They were divided by those of the fuel of the core at the shutdown to obtain a deposited radioactivity fractions of the nuclides as relative values to 137Cs, which also correspond to deposition fractions of the elements as relative values to Cs. They were estimated to be orders of 10−4 to 10−3 for Sr, 10−4 for Nb, 10−2 to 10−1 for Mo, 10−1 for Ag, 10−1 to 100 for Te, 100 for I, 10−3 for Ba, 10−6 to 10−5 for Pu, 10−6 to 10−5 for Am, and 10−6 for Cm. The observed radioactivity ratios to 137Cs were compared with those obtained by severe accident analysis to assess the validation of the analysis.

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.

Melting Simulation Using a Particle Method With Angular Momentum Conservation

2017 ◽  
Author(s):  
Masahiro Kondo ◽  
Shota Ueda ◽  
Koji Okamoto
Keyword(s):  
Angular Momentum ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Particle Method ◽  
Momentum Conservation ◽  
Severe Accident ◽  
Angular Momentum Conservation ◽  
The Core ◽  
Melting And Solidification

To analyze the core degradation and relocation behavior of melts in a severe accident of nuclear power plant, the melting and solidification in the complexed geometry is to be calculated. For the calculation of such complexed behavior, a new particle method conserving angular momentum is proposed and applied for the melting simulation. When solid melts, it may move like a rigid body. The angular momentum conservation is important to capture such kind of motion. The potential of the new particle method was confirmed with a calculation of the melting in dam break geometry and cantilever geometry.

Modeling of Dryout in Core Debris Beds of Conical and Cylindrical Geometries

2012 ◽  
Author(s):  
Eveliina Takasuo ◽  
Ville Hovi ◽  
Mikko Ilvonen ◽  
Stefan Holmström
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Flow Field ◽  
Nuclear Power ◽  
Experimental Results ◽  
Severe Accident ◽  
Accident Analysis ◽  
Simulation Results ◽  
Particle Bed ◽  
Good Agreement

A porous particle bed consisting of core debris may be formed as a result of a core melt accident in a nuclear power plant. The coolability of conical (heap-like) and cylindrical (evenly-distributed) ex-vessel debris beds have been investigated in the COOLOCE experiments at VTT. The experiments have been modeled by using the MEWA severe accident analysis code. The main objectives of the modeling were (1) to validate the simulation results against the experiments by comparing the dryout power density predicted by the code to the experimental results and (2) to evaluate the effect of geometry on the coolability by examining the flow field and the development of dryout in the two geometries. In addition to the MEWA simulations, 3D demonstration calculations of the particle bed dryout process have been performed using the in-house code PORFLO. It was found that the simulation and experimental results are in a relatively good agreement. The results suggest that the coolability of the conical debris bed is poorer than that of the cylindrical bed due to the greater height of the conical configuration.

Phenomenological Analysis of Melt Progression in LWRS: Proposal of an In-Vessel Corium Progression Scenario in the Unit 1 of Fukushima Dai-ichi Nuclear Power Plant

2016 ◽  
Vol 2 (4) ◽  
Author(s):  
Payot Frédéric ◽  
Seiler Jean-Marie
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Large Scale ◽  
Initial Conditions ◽  
Physical Models ◽  
Severe Accident ◽  
Fukushima Accident ◽  
Support Plate ◽  
The Core

In the field of severe accident, the description of corium progression events is mainly carried out using integral calculation codes. However, these tools are usually based on bounding assumptions because of the high complexity of phenomena. The limitations associated with bounding situations [1] (e.g., steady-state situations and instantaneous whole core relocation in the lower head) led CEA to develop an alternative approach to improve the phenomenological description of the melt progression. The methodology used to describe the corium progression was designed to cover the accidental situations from the core meltdown to the molten core–concrete interaction (MCCI). This phenomenological approach is based on the available data (including learnings from TMI-2) on physical models and knowledge about the corium behavior. It provides emerging trends and best-estimate intermediate situations. As different phenomena are unknown, but strongly coupled, uncertainties at large scale for the reactor application must be taken into account. Furthermore, the analysis is complicated by the fact that these configurations are most probably three-dimensional (3D), all the more so because 3D effects are expected to have significant consequences for the corium progression and the resulting vessel failure. Such an analysis of the in-vessel melt progression was carried out for the Unit 1 of the Fukushima Dai-ichi Nuclear Power Plant. The core uncovering kinetics governs the core degradation and impacts the appearance of the first molten corium inside the core. The initial conditions used to carry out this analysis are based on the available results derived from codes such as the MELCOR calculation code [2]. The core degradation could then follow different ways: (1) Axial progression of the debris and the molten fuel through the lower support plate, or (2) lateral progression of the molten fuel through the shroud. On the basis of the Bali program results [3] and the TMI-2 accident observations [4], this work is focused on the consequences of a lateral melt progression (not excluding an axial progression through the support plate). Analysis of the events and the associated time sequence will be detailed. Besides, this analysis identifies some number of issues. Random calculations and statistical analysis of the results could be performed with calculation codes such as LEONAR–PROCOR codes [5]. This work was presented in the frame of the OECD/NEA/CSNI Benchmark Study of the Accident at the Fukushima Dai-ichi Nuclear Power Station (BSAF) project [6]. During the years of 2012 and 2014, the purpose of this project was both to study, by means of severe accident codes, the Fukushima accident in the three crippled units, until 6 days from the reactor shutdown, and to give information about, in particular, the location and composition of core debris.

Sorption and desorption studies of Cs and Sr in contaminated soil samples around Fukushima Daiichi Nuclear Power Plant

2019 ◽  
Vol 20 (1) ◽  
pp. 392-403 ◽  
Author(s):  
Sharayu Kasar ◽  
Suchismita Mishra ◽  
Yasutaka Omori ◽  
Sarata Kumar Sahoo ◽  
Norbert Kavasi ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Contaminated Soil ◽  
Nuclear Power ◽  
Soil Samples ◽  
Fukushima Daiichi

Analysis of Inertization Strategies for the Filtered Containment Venting System in Cofrentes Nuclear Power Plant

2018 ◽  
Vol 4 (3) ◽  
Author(s):  
Kevin Fernández-Cosials ◽  
Gonzalo Jiménez ◽  
César Serrano ◽  
Luisa Ibáñez ◽  
Ángel Peinado
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Computational Model ◽  
Nuclear Power ◽  
Inert Atmosphere ◽  
Severe Accident ◽  
Nitrogen Injection

During a severe accident (SA) in a nuclear power plant (NPP), there are several challenges that need to be faced. To coup with a containment overpressure, the venting action will lower the pressure but it will release radioactivity to the environment. In order to reduce the radioactivity released, a filtered containment venting system (FCVS) can be used to retain iodine and aerosols radioactive releases coming from the containment atmosphere. However, during a SA, large quantities of hydrogen can also be generated. Hydrogen reacts violently with oxygen and its combustion could impair systems, components, or structures. For this reason, to protect the integrity of the FCVS against hydrogen explosions, an inertization system is found necessary. This system should create an inert atmosphere previous to any containment venting that impedes the contact of hydrogen and oxygen. In this paper, the inertization system for Cofrentes NPP is presented. It consists of a nitrogen injection located in three different points. A computational model of the FCVS as well as the inertization system has been created. The results show that if the nitrogen sweeps and the containment venting are properly synchronized, the hydrogen risk could be reduced to a minimum and therefore, the integrity of the FCVS would be preserved.

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