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.

scholarly journals Extended Station Blackout Coping Capabilities of APR1400

2014 ◽  
Vol 2014 ◽  
pp. 1-10 ◽  
Author(s):  
Sang-Won Lee ◽  
Tae Hyub Hong ◽  
Mi-Ro Seo ◽  
Young-Seung Lee ◽  
Hyeong-Taek Kim
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Water Injection ◽  
Severe Accident ◽  
Injection Strategy ◽  
Safety Improvement ◽  
External Injection ◽  
Station Blackout ◽  
The Government ◽  
External Water

The Fukushima Dai-ichi nuclear power plant accident shows that an extreme natural disaster can prevent the proper restoration of electric power for several days, so-called extended SBO. In Korea, the government and industry performed comprehensive special safety inspections on all domestic nuclear power plants against beyond design bases external events. One of the safety improvement action items related to the extended SBO is installation of external water injection provision and equipment to RCS and SG. In this paper, the extended SBO coping capability of APR1400 is examined using MAAP4 to assess the effectiveness of the external water injection strategy. Results show that an external injection into SG is applicable to mitigate an extended SBO scenario. However, an external injection into RCS is only effective when RCS depressurization capacity is sufficiently provided in case of high pressure scenarios. Based on the above results, the technical basis of external injection strategy will be reflected on development of revised severe accident management guideline.

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.

Education and Exercise for Severe Accident Management at KANSAI

2017 ◽  
Author(s):  
Takumi Kawahara ◽  
Tsugio Shiozawa ◽  
Tsutomu Nishioka ◽  
Yukimoto Shimominami ◽  
Hiroaki Onooka
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Exercise Program ◽  
Severe Accident ◽  
Staff Members ◽  
Power Company ◽  
Safety Improvement ◽  
Management Ability ◽  
Accident Management ◽  
Response Organization

In case of a severe accident, all the staff members of a nuclear power plant (NPP), members of the Emergency Response Organization (ERO) on site as well as operators in the main control room (MCR) are required to take necessary actions to mitigate the consequential effects of the accident. Therefore, Nuclear Engineering Ltd (NEL) has been implementing education and exercise for severe accident (SA) management both at nuclear power plants and Nuclear Power Division of Kansai Electric Power Company (KANSAI) since FY 2014. For the education of commanders who take a lead at the ERO in case of an accident, table top exercise is provided by using simulators developed by NEL, including functions to respond to a SA involving core melt. Continuous implementation of this education and exercise program is expected to enhance KANSAI’s severe accident management ability and their voluntary safety improvement activities in the future.

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.

Kinetic Study of Containment Retention Factor (CRF) for Large Dry Containment Under Radiation Load

2014 ◽  
Author(s):  
Khurram Mehboob ◽  
Kwangheon Park ◽  
Rehan Khan ◽  
Majid Ali ◽  
Raheel Ahmed
Keyword(s):  
Kinetic Study ◽  
Nuclear Power ◽  
Power Plants ◽  
Heat Removal ◽  
Fission Products ◽  
Retention Factor ◽  
Severe Accident ◽  
Spray System ◽  
Filtration System ◽  
Kinetic Dependency

The Nuclear Power Plants (NPPs) have been built on the concept of Defense in depth. The severe accident causes the failure of fission product barriers and let the fission products to escape into environment. The containment is the last barrier to the fission products. Thus, the containment is installed with engineering safety features (ESFs) i.e. spray system, heat removal system, recirculation filtration system; containment filtered venting system (CFVS), and containment exhaust filtration system. In this work, kinetic study of the containment retention factor (CRF) has been carried out for a large dry PWR containment considering 1000 MWe PWR. The computational modeling and simulation have been carried out by developing a kinetic code in MATLAB, which uses the fractions of activity airborne into the containment after the accident. The Kinetic dependency of CRF on containment filtration systems, spray system with caustic and boric acid spray has been carried out. For noble gases, iodine and aerosols, the CRF increases with the increase in exhaust rate. While, CRF for iodine first increases then start reducing with containment spray flow rate. The Kinetic dependency of CRF has also been studied for boric and caustic spray.

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.

Sign in / Sign up

Export Citation Format

Share Document