scholarly journals Risk due to creep of prestressed concrete at moderate temperature

2019 ◽  
Vol 281 ◽  
pp. 01007
Author(s):  
Thierry Vidal ◽  
Hugo Cagnon ◽  
Nam Nguyen ◽  
Jean-Michel Torrenti ◽  
Alain Sellier
Keyword(s):  
Internal Pressure ◽  
Prestressed Concrete ◽  
Nuclear Power ◽  
Power Plants ◽  
Severe Accident ◽  
Experimental Program ◽  
Strong Increase ◽  
Creep Tests ◽  
Accident Conditions ◽  
Creep Deformations

This study is a part of a French national project dealing with the mechanical behaviour of the containment vessel of French Nuclear Power Plants in case of a severe accident. The accident conditions are characterized by the increases of internal pressure, +0.5 MPa, and of temperature, up to 180°C, during two weeks. Heating can induce a strong increase of creep deformations kinetics leading to prestressing losses of concrete. Associated to internal pressure, tensile stress could occur in some areas of the structure and the potential cracking could affect the containment capacity of the vessel. One of the objectives of the project was thus to provide original creep data to develop accurate models, taking into account the coupled effects of temperature, desiccation and damage, and able to predict the behaviour of prestressed concrete structures in such insitu conditions. A wide experimental program consisted of numerous creep tests under various thermo-hydro-mechanical conditions in the values range of the accident. The presented results concern uniaxial compressive and flexural creep tests respectively performed on concrete cylinders and prestressed concrete beams, at 20°C and 40°C without desiccation.

Modelling Basic Creep of Concrete at Elevated Temperatures and Stresses

2016 ◽  
Vol 711 ◽  
pp. 879-884
Author(s):  
Jean Michel Torrenti
Keyword(s):  
Prestressed Concrete ◽  
Nuclear Power ◽  
Power Plants ◽  
Elevated Temperatures ◽  
Severe Accident ◽  
Basic Creep ◽  
The Third ◽  
Cooling Agent ◽  
Concrete Confinement ◽  
Kinetics Of

The prestressed concrete confinement vessel is the third and last barrier in Nuclear Power Plants (NPP). In case of a severe accident (loss of cooling agent of the reactor for instance), pressure and temperature will increase in the nuclear vessel (0,5 MPa and 180°C during 2 weeks). Due to elevated temperatures, the evolution of basic creep will be accelerated. In this case, due to internal pressure, some tensile stresses could appear in specific parts of the structure and induce cracking. The modelling of basic creep and its couplings with temperature is very important for the safety of the structure (tightness of the concrete vessel). Here we present a model considering the following elements: a coupling between creep and damage is introduced, kinetics of basic creep is affected by temperature by the means of an Arrhenius thermo-activation, damage due to the increase of temperature is taken into account. The model is compared with the available experimental results. This work is a part of the MACENA project.

Experimental Investigation Into Buckling Failure of Slender Metal Plates in Severe Accident Conditions

2014 ◽  
Author(s):  
Byeongnam Jo ◽  
Wataru Sagawa ◽  
Koji Okamoto
Keyword(s):  
High Speed ◽  
Nuclear Power ◽  
Power Plants ◽  
Failure Load ◽  
Severe Accident ◽  
Steel Columns ◽  
Wide Range ◽  
Buckling Failure ◽  
The Difference ◽  
Accident Conditions

Buckling failure load of stainless steel columns under compressive stress was experimentally measured in severe accident conditions, which addresses the accidents in Fukushima Daiichi nuclear power plants. Firstly, buckling failure load defined as load which causes failure of the column (plastic collapse) was measured in a wide range of temperatures from 25 °C up to 1200 °C. The load values measured in this study were compared to numerical estimations by eigenvalue simulations (for an ideal column) and by nonlinear simulations (for a column with initial bending). Two different methods for measurement of the buckling failure load were employed to examine the effect of thermal history on buckling failure. Different load values were obtained from two methods in high temperature conditions over 800 °C. The difference in the buckling failure load between two methods increased with temperature, which was explained by the effect of creep at high temperatures. Moreover, the influence of asymmetric temperature profiles along a plate column was also explored with regard to the failure mode and the buckling failure load. In present study, all of the buckling processes were visualized by a high speed camera.

Creation of Automatic System for Monitoring, Forecasting and Control of Condition of Lead-Bismuth (Lead) Coolant and Surfaces of Circuits of Nuclear Power Plants

2009 ◽  
Author(s):  
P. N. Martynov ◽  
R. Sh. Askhadullin ◽  
A. A. Simakov ◽  
A. Yu. Chaban’ ◽  
M. E. Chernov ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Reactor Core ◽  
Heat Removal ◽  
Severe Accident ◽  
Medium Size ◽  
Safety Level ◽  
Nitride Fuel ◽  
Active Safety Systems ◽  
Accident Conditions

Lead-bismuth coolant is preferable for the medium size reactors, since, in contrast to the sodium coolant, it does not interact with water and air, it is radiation resistant, insignificantly activated and it is not combustible [1]. Combination of natural properties of lead-based coolants, mono-nitride fuel, fast reactor neutronics and design approaches used for the reactor core and heat removal system brings SVBR 75/100 NPP [2] to achieve a new safety level and assures its stability without operation of active safety systems even under severe accident conditions. Analysis of possible sequences of the events even under conditions of such severe accidents as addition of total excess reactivity or all pumps trip accompanied by safety system failure leads to the conclusion on that power unit with SVBR 75/100 reactor plant (RP) has high safety level.

scholarly journals Containment Depressurization Capabilities of Filtered Venting System in 1000 MWe PWR with Large Dry Containment

2014 ◽  
Vol 2014 ◽  
pp. 1-10 ◽  
Author(s):  
Sang-Won Lee ◽  
Tae-Hyub Hong ◽  
Yu-Jung Choi ◽  
Mi-Ro Seo ◽  
Hyeong-Taek Kim
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Noble Gas ◽  
Decontamination Factor ◽  
Severe Accident ◽  
Spray System ◽  
Discharge Flow Rate ◽  
Safety Improvement ◽  
Accident Conditions ◽  
Containment Integrity

After the Fukushima Daiichi nuclear power plant accident, the Korean government and nuclear industries performed comprehensive safety inspections on all domestic nuclear power plants against beyond design bases events. As a result, a total of 50 recommendations were defined as safety improvement action items. One of them is installation of a containment filtered venting system (CFVS) or portable backup containment spray system. In this paper, the applicability of CFVS is examined for OPR1000, a 1000 MWe PWR with large dry containment in Korea. Thermohydraulic analysis results show that a filtered discharge flow rate of 15 [kg/s] at 0.9 [MPa] is sufficient to depressurize the containment against representative containment overpressurization scenarios. Radiological release to the environment is reduced to10-3considering the decontamination factor. Also, this cyclic venting strategy reduces noble gas release by 50% for 7 days. The probability of maintaining the containment integrity in level 2 probabilistic safety assessment (PSA) initiating events is improved twofold, from 43% to 87%. So, the CFVS can further improve the containment integrity in severe accident conditions.

Calculation of Evaporation From Fukushima NPP Spent Fuel Pools

2018 ◽  
Author(s):  
Mirza M. Shah
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Severe Accident ◽  
Spent Fuel ◽  
Decay Heat ◽  
Wide Range ◽  
Spent Fuel Pool ◽  
Accident Conditions ◽  
Safety Systems ◽  
Good Agreement

Prediction of evaporation rates from spent fuel pools of nuclear power plants in normal and post-accident conditions is of great importance for the design of safety systems. A severe accident in 2011 Fukushima nuclear power plant caused failure of cooling systems of its spent fuel pools. The post-accident evaporation from the spent fuel pools of Fukushima units 2 and 4 is compared to a model based on analogy between heat and mass transfer which has been validated with a wide range of data from many water pools including a spent fuel pool. Calculations are done with two published estimates of fuel decay heat, one 25 % lower than the other. The model predictions are close to the evaporation using the lower estimate of decay heat. Other relevant test data are also analyzed and found in good agreement with the model.

Pool Scrubbing Efficiency of Hypothesized Filtered Containment Venting System for APR1400 Nuclear Power Plants

2017 ◽  
Author(s):  
Seungwon Seo ◽  
Jungjae Lee ◽  
Yongjin Cho
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Severe Accident ◽  
Accident Analysis ◽  
Radioactive Aerosol ◽  
Water Pool ◽  
Radioactive Materials ◽  
Accident Conditions ◽  
Containment Pressure

For a severe accident (a core melting accident) of nuclear power plants, a heat-up of the molten core might cause a overpressurizing of containment building to be damaged, if there couldn’t be given a proper cooling and/or a depressurizing strategy. In order to depressurize containment building and also to minimize the release of radioactive materials, filtered containment venting system (FCVS) might be used for a one of possible options. For a wet-type FCVS, radioactive aerosol released from molten core could be decontaminated by water pool, which is called pool scrubbing effect. The objective of this study is to find out regulatory insights for evaluating a wet-type FCVS for Korean nuclear power plant, APR1400. MELCOR, which is a severe accident analysis code developed by Sandia National Laboratories, was used for simulating postulated accidents. A full-plant scale calculation was performed considering the accident conditions such as temperature, pressure flow rate from containment to the pool of FCVS, behavior of radioactive materials and decontamination factors (DFs) for them. FCVS was operated with containment pressure set points. The decrease thermal margin between containment atmosphere and the pool of the FCVS influenced the DF, because the decreased amount of the steam due to the lowered thermal margin interrupted the radioactive aerosols and steam condensed.

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