Research about the Hydrogen Removal Verification Method through Porous Medium of Silver Zeolite in Nuclear Power Station

2020 ◽  
Vol 999 ◽  
pp. 31-38
Author(s):  
Qing Bo Bao ◽  
Jian Hu ◽  
Shao Fei Zhou
Keyword(s):  
Nuclear Power Station ◽  
Nuclear Power ◽  
Evaluation Method ◽  
Natural Circulation ◽  
Catalyst Layer ◽  
Severe Accident ◽  
Test Device ◽  
Power Station ◽  
Verification Method ◽  
Silver Zeolite

After severe accident in the nuclear power station, it is necessary to remove the hydrogen timely for the purpose of preventing the containment integrity from breach. This report has investigated and studied the role of silver zeolite in the reaction of hydrogen and oxygen. According to the catalyst role, the principle test device for hydrogen removal with silver zeolite is provided. The force of natural circulation for principle test device is created by the Chimney Effect, which is the result of different density between the internal and external of the device. Also, this report suggests the flowing capability calculation method of up-thrust about the mixture gas passing through the catalyst layer of silver zeolite. The evaluation method of hydrogen removal efficiency with silver zeolite is described. Finally, this report gives the method of CFX numeric analog and the specific simulating steps for the layer of silver zeolite using for catalytic role.

Research on Estimating Methods and Application of Fuel Rods Defect

2017 ◽  
Author(s):  
Huaibin Li ◽  
Yaru Fu ◽  
Lanfang Mao ◽  
Qiliang Mei
Keyword(s):  
Radiation Dose ◽  
Nuclear Power Station ◽  
Nuclear Power ◽  
Evaluation Method ◽  
Fission Products ◽  
Fuel Management ◽  
Power Station ◽  
Fuel Rods ◽  
Reactor Coolant ◽  
The World

When fuel rods have defects, the fission products in the fuel rods will come out and enter the reactor coolant through the defects. The release of the fission products will result in the increase of radiation dose, the indeterminacy of the fuel management and will influence the economics and safety of the nuclear power station. Based on the analysis of the typical nuclides activities in reactor coolant, the evaluation of the defect fuel rods can be realized. This paper studied the related analysis around the world and determined the methods to evaluate the number, the defect type (open or tight) and the burn-up of the defect fuel rods. The evaluation method of this paper can be used to evaluate the defects of fuel rods, and can provide valuable information for the fuel management and dose analysis, and also can be a useful technical support to the operation of nuclear power station.

Long-Term Cooling Strategy for the Primary Containment Vessel of the Kashiwazaki-Kariwa Nuclear Power Station in a Severe Accident

2017 ◽  
Author(s):  
Satoshi Kawaguchi ◽  
Satoshi Mizuno ◽  
Yoshihiro Oyama
Keyword(s):  
Heat Exchanger ◽  
Nuclear Power Station ◽  
Nuclear Power ◽  
Heat Removal ◽  
Severe Accident ◽  
Power Station ◽  
Working Temperature ◽  
Containment Vessel ◽  
Portable Pump

This paper explains the strategy of our company (Tokyo Electric Power, TEPCO) regarding means of long-term heat removal from the primary containment vessel (PCV) of Units 6 and 7 (ABWR) of the Kashiwazaki-Kariwa Nuclear Power Station in a severe accident. If the PCV continues in a high-temperature state for a long time, the strength of the PCV concrete will decline, and the risk of being affected by an earthquake will increase. Therefore, it is crucial for safety to cool the PCV and reduce its temperature to the maximum working temperature or lower. TEPCO provides a means of cooling the reactor pressure vessel (RPV) and PCV called the alternative coolant circulation system (ACCS). This system uses the heat exchanger of the residual heat removal (RHR) system, the make up water condensate (MUWC) pump, and alternative heat exchanger vehicles. By using these measures, it is possible reduce temperature in the PCV over the long term to the maximum working temperature (design value) or less, even in severe accident scenarios such as a large LOCA + ECCS function failures + SBO (station blackout). This function has quite high reliability, but in a scenario where these measures cannot be used, expectations are placed on the filtered vent (FV). However, due to FV characteristics, it is impossible to reduce to below the saturation temperature of 100°C at atmospheric pressure using FV alone, and it will be necessary in the medium/long-term to cool the PCV while also restoring the cooling equipment. Therefore, the following restoration operation of PCV cooling and its dose evaluation were studied. (1) RPV heat removal by restoring the RHR system (2) RPV and PCV heat removal using a portable pump employing a portable heat exchanger (3) RPV and PCV heat removal using the suppression pool water clean up system (SPCU) employing portable heat exchangers (4) RPV heat removal using the clean up water system (CUW) By clarifying beforehand issues such as feasibility of these systems, the on-site environment for restoration measures, and the necessary gear/systems, the authors were able to secure means of long-term cooling of the PCV, and further enhance PCV reliability.

Incorporation of the Radioactive Interinfluence in Multi-Unit Seismic PRA

2018 ◽  
Author(s):  
Shuhei Matsunaka ◽  
Chikahiro Sato ◽  
Manabu Watanabe
Keyword(s):  
Nuclear Power Station ◽  
Nuclear Power ◽  
Power Plants ◽  
Evaluation Method ◽  
Common Factor ◽  
Radioactive Material ◽  
Power Station ◽  
Core Damage ◽  
Accident Management ◽  
Multiple Units

Kashiwazaki-Kariwa nuclear power station of TEPCO is the largest nuclear power station in the world, and it has seven nuclear power plants. As the experience at Fukushima Daiichi nuclear power station accident in March 2011 involving concurrent core damage at multiple units, it is considered that the risk derived from hazards of Earthquakes and Tsunamis is relatively significant in Japan, and these events have a high likelihood of damaging multiple units simultaneously. Therefore, it is very important to grasp the multi-unit specific risk. Although there are some unique accident scenarios of Multi-Unit PRA, this paper focuses on the influence of radioactive materials released outside the containment vessel on the accident management of the adjacent unit. The events including core damage and loss of containment function should be considered as the causes of the release of radioactive substances, and operator’s operation or the like should be considered as objects to be adversely affected by them. It is necessary to incorporate that into PRA to confirm the effect on risk. It is very difficult in terms of the maturity of evaluation method and the calculation load to accurately incorporate consequences derived from time series of various events and complicated interaction into PRA model. Therefore, as the first step in evaluating the risk of influence of radioactive material release on the accident management, some streamlining efforts are implemented according to the purpose. For example, Kashiwazaki-Kariwa unit 6 and unit 7 were set as the target units for model simplification. We also assume the earthquake as the initiating event due to the strong common factor for multi units. Whether or not to be operable in the adjacent plant is set conservatively based on deterministic evaluation. PRA taking into consideration the radiation influence by multi-unit accident is compared with normal PRA. Some kind of Core Damage Frequency (CDF) such as CDF1 (Core Damage Frequency at which the damage of one or the other of two unit occur), CDF2 (CDF at which the damage of both of units occur) and CDFTOTAL (CDF at which the damage of one or more units occur: CDF1 + CDF2) are quantified, and the degree of this issue is provided. Although the change of CDFTOTAL was insignificant, the necessity of further study was shown from the viewpoint of the amount and timing of radioactive substance released due to an approximately 1.5-fold increase in CDF2.

scholarly journals A METHOD FOR EVALUATING THE MAXIMUM TSUNAMI LOADINGS ON SEAWALLS

2018 ◽  
pp. 91
Author(s):  
Takuya Toriyama ◽  
Kisaburo Azuma ◽  
Hiroshi Moritani ◽  
Nobuo Ishida
Keyword(s):  
Froude Number ◽  
Nuclear Power Station ◽  
Nuclear Power ◽  
Evaluation Method ◽  
Power Station ◽  
Wave Pressure ◽  
Flume Tests ◽  
Proper Design

A seawall in a nuclear power station is one of the important structures to protect from tsunami. Estimation of tsunami loadings on structures is an important part of the proper design of seawalls. In this study, hydraulic flume tests was conducted to investigate the characteristics of tsunami loadings. Correlations between the loading on a seawall and the Froude number as characteristics of the tsunami flow were investigated. Finally, we proposed a new evaluation method to evaluate the design wave pressure on a seawall. A new evaluation method can predict the design wave pressure on a seawall with taking the characteristics of the tsunami flow into consideration.

Study on Grey Comprehensive Evaluation Based on the APH of the Site Selection of China Inland Nuclear Power Station

2013 ◽  
Vol 357-360 ◽  
pp. 2810-2817 ◽  
Author(s):  
Yun Na Wu ◽  
Lei Tan ◽  
Ru Hang Xu ◽  
Yi Sheng Yang
Keyword(s):  
Nuclear Power Station ◽  
Site Selection ◽  
Nuclear Power ◽  
Evaluation Method ◽  
Comprehensive Evaluation ◽  
Evaluation Model ◽  
Influence Factors ◽  
Power Station ◽  
Station Location ◽  
Selection Of

The site selection of China inland nuclear power station is key issue in its early preparations, and there are many issues which affect it. However, now the study of the site selection of inland nuclear power station limits to the discussion on its influence factors, and the lack of scientific evaluation method to evaluate the reasonableness of the site selection. This article firstly uses analytical hierarchy process to establish index system for inland nuclear station location; secondly establishes the evaluation model based on the grey comprehensive evaluation method; and finally applies the model to one location of a nuclear power station for instance, which confirm the feasibility of the model. The grey comprehensive evaluation of this article can provide scientific and effective evaluation method for the site selection of inland nuclear power station.

Function of Isolation Condenser of Fukushima Unit-1 Nuclear Power Plant

2012 ◽  
Author(s):  
Masanori Naitoh ◽  
Hiroaki Suzuki ◽  
Hidetoshi Okada
Keyword(s):  
Nuclear Power Station ◽  
Nuclear Power ◽  
Failure Time ◽  
Sea Water ◽  
Water Injection ◽  
Heat Removal ◽  
Severe Accident ◽  
Power Station ◽  
Fukushima Daiichi ◽  
Fire Engine

The Tohoku Region Pacific Coast Earthquake with magnitude 9.0 occurred at 2:46 PM of March 11th, 2011, followed by a huge Tsunami. The Fukushima Daiichi nuclear power station suffered serious damages from the Tsunami, involving core melt and release of large amount of fission products to an environment. The station blackout (SBO) occurred due to submergence of emergency equipment by the sea water. The isolation condenser (IC) was the only device for decay heat removal at the unit-1 of the Fukushima Daiichi nuclear power station after the reactor scram. The IC function was analyzed with a severe accident analysis code SAMPSON. The analysis results showed that (1) core melt resulting in RPV failure occurred since the IC operation was limited because it was not designed as a countermeasure to mitigate severe accident progression in Japan and (2) even assuming the continuous IC operation after the SBO to mitigate severe accident progression, the RPV failure occurred at 18:52, March 12th. However, since the alternate water injection by a fire engine was actually ready to start at 5:46, March 12th, which was earlier than calculated RPV failure time, the RPV failure could be prevented by continuous IC operation.

scholarly journals Important Severe Accident Research Issues After Accident at Fukushima Daiichi Nuclear Power Station

2013 ◽  
Author(s):  
Jun Sugimoto
Keyword(s):  
Nuclear Power Station ◽  
Working Group ◽  
Nuclear Power ◽  
Severe Accident ◽  
Power Station ◽  
Fukushima Accident ◽  
Research Issues ◽  
Containment Vessel ◽  
Accident Research ◽  
Fukushima Daiichi

After the accident at Fukushima Daiichi Nuclear Power Station several investigation committees issued reports with lessons learned from the accident in Japan. Among those lessons, some recommendations have been made on severe accident research. Similar to the EURSAFE efforts under EU Program, review of specific severe accident research items was started before Fukushima accident in working group of Atomic Energy Society of Japan (AESJ) in terms of significance of consequences, uncertainties of phenomena and maturity of assessment methodology. Re-investigation has been started after the Fukushima accident in this working group. Additional effects of Fukushima accident, such as core degradation behaviors, sea water injection, containment failure/leakage and re-criticality have been covered. The review results are categorized in ten major fields; core degradation behavior, core melt coolability/retention in containment vessel, function of containment vessel, source term, hydrogen behavior, fuel-coolant interaction, molten core concrete interaction, direct containment heating, recriticality and instrumentation in severe accident conditions. In January 2012, Research Expert Committee on Evaluation of Severe Accident was established in AESJ in order to investigate severe accident related issues for future LWR development and to propose action plans for future severe accident research, in collaboration with this working group. Based on these activities and also author’s personal view, the present paper describes the perspective of important severe accident research issues after Fukushima accident. Specifically those are investigation of damaged core and components, advanced severe accident analysis capabilities and associated experimental investigations, development of reliable passive cooling system for core/containment, analysis of hydrogen behavior and investigation of hydrogen measures, enhancement of removal function of radioactive materials of containment venting, advanced instrumentation for the diagnosis of severe accident and assessment of advanced containment design which excludes long-term evacuation in any severe accident situations.

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