scholarly journals Corrosion Mitigation Activities Performed After the Fukushima Daiichi Accident

CORROSION ◽  
10.5006/2695 ◽  
2017 ◽  
Vol 74 (5) ◽  
pp. 577-587 ◽  
Author(s):  
Yuichi Fukaya ◽  
Toshifumi Hirasaki ◽  
Katsuhiko Kumagai ◽  
Teruhisa Tatsuoka ◽  
Kenro Takamori ◽  
...  
Keyword(s):  
Dissolved Oxygen ◽  
Nuclear Power ◽  
Cooling Water ◽  
Chloride Ion ◽  
Oxygen Removal ◽  
Ion Concentration ◽  
Localized Corrosion ◽  
Mitigation Measures ◽  
Spent Fuel ◽  
Fukushima Daiichi

This synopsis describes corrosion issues and mitigation activities shortly after the Fukushima Daiichi Nuclear Power Station accident. An earthquake of magnitude 9.0 occurred on March 11, 2011; the subsequent tsunami removed the cooling capacity of fuels in both the reactors and spent fuel pools (SFPs). Seawater was temporarily used for emergency fuel cooling, which induced various corrosion issues. Just after the accident, the temperatures within the reactors of Units 1 to 3 increased to several hundred degrees Celsius and the water quality of the cooling water seems to have become similar to that of concentrated seawater. To stabilize the fuel cooling, corrosion mitigation actions were required for mainly carbon steel components. The following corrosion mitigation measures were applied to the reactors: (a) temperature decreases, (b) dissolved oxygen removal from feedwater via deaeration, (c) dissolved oxygen removal from cooling water in the reactors via nitrogen gas injection, (d) salt removal from cooling water, and (e) sterilization of feedwater by hydrazine addition. The temperatures of SFP water in Units 1 to 4 were between 47°C and 93°C just after the accident. The maximum chloride ion concentration was approximately 2,000 ppm and the pH was in the range from 7.5 to 11.2. The mitigation of localized corrosion of the stainless steel pool liners and alkaline corrosion of the aluminum fuel racks was the top priority. In addition to (a), (b), and (d) listed above, (f) dissolved oxygen removal and sterilization by hydrazine addition and (g) pH control were applied to the SFPs. In the six years since the accident, no major corrosion problems have yet arisen. However, continued efforts to increase plant stability are underway for the long-term goal of decommissioning.

Life Prediction Model for Nuclear Power Plant Intake Structure Exposed to Chloride Environment Based on Similarity Principle

2020 ◽  
Author(s):  
Sen Chen ◽  
Zhen Wang ◽  
Jiangtao Zhang ◽  
Dahua Cai ◽  
Jiyun Zhou
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Salt Spray ◽  
Chloride Ion ◽  
Ion Concentration ◽  
Test Results ◽  
Ion Diffusion ◽  
Similarity Principle ◽  
Intake Structure

Abstract Three environmental simulation experiments for accelerating the chloride ingress were designed based on the similarity principle and the actual Nuclear Power Plant (NPP) intake structure service environment parameters including alternate drying-wetting condition, immersion condition and salt spray condition. Different experiment conditions were designed for each three-simulation experiment condition based on the similarity principle. The temperature, chloride ion concentration and dry-wet cycle time ratio were selected parameters for alternation wetting-drying experiments. The temperature and chloride ion concentration were selected parameters for the salt spray experiment and immersion experiment. The distribution of chloride ion concentration along the depth of the concrete samples were measured every 30 days. The testing results showed that chloride ion diffusion depth is maximum under the dry-wet cycle test, chloride ion diffusion depth is minimum under salt spray test. The curves of the chloride ion concentration along depth increased firstly and then decreased, which is consistent with the two-dimension diffusion law. Finally, a life prediction model that can be used to predict the concrete structure of the existing coastal nuclear power plant was developed based on the test results and field test results.

Probabilistic Assessment of Countermeasures for Loss of Ultimate Heat Sink to Spent Fuel Pool

2017 ◽  
Author(s):  
Bumpei Fujioka ◽  
Naoki Hirokawa ◽  
Daisuke Taniguchi
Keyword(s):  
Heat Sink ◽  
Nuclear Power ◽  
Water System ◽  
Cooling Water ◽  
Severe Accident ◽  
Probabilistic Assessment ◽  
Spent Fuel ◽  
The Past ◽  
Plant Safety ◽  
Spent Fuel Pool

In the Fukushima Dai-ichi nuclear power station, Loss of Ultimate Heat Sink (LUHS) was caused by the great east japan earthquake and the subsequent tsunami [1]. It resulted in severe accident in three units. In that time, fuel damage in Spent Fuel Pool (SFP) were prevented by the various countermeasures such as makeup by pump truck and recovery of injection systems /cooling water system. In the past, Probabilistic Safety Assessment (PSA) has been developed with a focus on the reactor. After the accident, it has been acknowledged that SFP PSA is important to enhance the plant safety. In this study, probabilistic assessment is performed to suggest countermeasures for LUHS to SFP.

Lessons Learned From the Operation of Large-Scale Reprocessing-Recycling Facilities in France: What Is in It for China?

2017 ◽  
Author(s):  
Jean-Pierre Gros
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Large Scale ◽  
Central Government ◽  
Fuel Cycle ◽  
Public Information ◽  
Lessons Learned ◽  
Mitigation Measures ◽  
Spent Fuel ◽  
The Impact

AREVA has been running since decades nuclear reprocessing and recycling installations in France. Several industrial facilities have been built and used to this aim across the time. Following those decades and with the more and more precise monitoring of the impact of those installations, precise data and lessons-learned have been collected that can be used for the stakeholders of potential new facilities. China has expressed strong interest in building such facilities. As a matter of fact, the issue of accumulation of spent fuel is becoming serious in China and jeopardizes the operation of several nuclear power plants, through the running out of space of storage pools. Tomorrow, with the extremely high pace of nuclear development of China, accumulation of spent fuel will be unbearable. Building reprocessing and recycling installations takes time. A decision has to be taken so as to enable the responsible development of nuclear in China. Without a solution for the back end of its nuclear fuel cycle, the development of nuclear energy will face a wall. This is what the Chinese central government, through the action of its industrial CNNC, has well understood. Several years of negotiations have been held with AREVA. Everybody in the sector seems now convinced. However, now that the negotiation is coming to an end, an effort should be done towards all the stakeholders, sharing actual information from France’s reference facilities on: safety, security, mitigation measures for health protection (of the workers, of the public), mitigation measures for the protection of the environment. Most of this information is public, as France has since years promulgated a law on Nuclear transparency. China is also in need for more transparency, yet lacks means to access this public information, often in French language, so let’s open our books!

Strategies to Ensure Safety Measures for ABWRs in Kashiwazaki-Kariwa Nuclear Power Station

2016 ◽  
Author(s):  
Satoshi Mizuno ◽  
Toshihiro Matsuo ◽  
Shinichi Kawamura
Keyword(s):  
Nuclear Power Station ◽  
Nuclear Power ◽  
Mitigation Measures ◽  
Circulation System ◽  
Power Station ◽  
Safety Measures ◽  
Core Damage ◽  
Upper Limit ◽  
Fukushima Daiichi ◽  
Containment Integrity

This paper describes strategies to ensure safety measures for ABWRs in Kashiwazaki-Kariwa Nuclear Power Station reflecting the lessons learned from the Fukushima Daiichi accident. The accident and response actions were analysed to extract lessons. Based on the lessons policies to enhance safety and containment integrity were derived. Firstly with considerations of multiple failures of the Fukushima Daiichi accident, defense in depth (DID) was enhanced by applying more diverse safety measures. For this purpose, in addition to refurbishing safety measures for beyond design base events, safety measures for each DID layer was enhanced not only by strengthen robustness to single failure but also by strengthen diversity and by physical separation. Especially capacities for ensuring containment integrity have been drastically improved by installing diverse safety measures for the third layer and the fourth layer of DID including the alternative coolant circulation system. Secondly phased approach was introduced in choosing mitigation measures considering timing of the response actions and required reliabilities. The basic concept of phased approach is that we have to select safety measures based on the assumption that possible and available measures and their required reliability are different depending on time to spare. Lastly performance requirements was clarified for containment vessel and its auxiliary systems after core damage. The basic policy was to define clearly twice design pressure of PCV as its upper limit pressure and to define 200 degree C as PCV upper limit temperature, and then to define function requirements for other auxiliary systems in order to contain fission products in the PCV and to make them decay as long as possible not only by enhancing PCV capacity but by utilizing auxiliary systems to ensure PCV integrity after core damage. Various safety measures were implemented based on these policies and applied at ABWRs in Kashiwazaki-Kariwa Nuclear Power Station.

scholarly journals Leakage Position Estimation of Cooling Water Using a Stereo Camera for Fukushima Daiichi Nuclear Power Plant

2021 ◽  
Vol 11 (17) ◽  
pp. 7796
Author(s):  
Hao Xu ◽  
Ren Komatsu ◽  
Hanwool Woo ◽  
Atsushi Yamashita ◽  
Hajime Asama
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling Water ◽  
Position Estimation ◽  
Stereo Image ◽  
Inspection System ◽  
Stereo Camera ◽  
Radiation Noise ◽  
Fukushima Daiichi

In this study, a new method of estimating the leakage positions of cooling water using a stereo camera for the Fukushima Daiichi Nuclear Power Plant (FDNPP) is proposed. A stereo camera mounted on an inspection system with a rotating base was inserted into the pedestal below the reactor pressure vessel (RPV), and the waterdrops from the leakage position were captured using a stereo camera. We estimated the leakage positions by triangulating the waterdrop trajectory lines in the stereo image. The main contribution of this study is the extraction and matching of the waterdrop trajectories in a stereo image in the FDNPP. The radiation noise is intense inside the pedestal because of the presence of fuel debris. Therefore, we propose a method that is robust against radiation noise. We assume that the waterdrops drip vertically in indoor environments without wind, such as in the FDNPP. Hence, the orientation of the stereo camera can be adjusted by the rotating base such that the vertical lines in the three-dimensional space are also projected as vertical lines in the image planes. Thereafter, the columns of pixels in the images are treated as image features and used to extract and match the waterdrop trajectories. We demonstrated the effectiveness of our leakage position estimation method in a simulated environment of the FDNPP with gamma-ray image noise.

Estimation About Amounts of Radioactive Wastes From Fukushima Daiichi Nuclear Power Station

2016 ◽  
Author(s):  
Taichi Sakai ◽  
Shunichi Suzuki ◽  
Koji Okamoto
Keyword(s):  
Migration Rate ◽  
Nuclear Power ◽  
Cooling Water ◽  
Radioactive Wastes ◽  
Power Station ◽  
Current Standard ◽  
Migration Rates ◽  
Elapsed Time ◽  
Fukushima Daiichi ◽  
Almost All

In this research, the amounts of the radioactive wastes generated by the secondary wastewater treatment was focused on. In this paper, at first, cooling water circulating in the plant and the facilities which are removing radionuclides were modeled and how the radionuclides migrate with elapsed time was formulated. Then, the analytic value was fitted to the measured value to compute undetermined coefficients of the formula by setting a migration rate of these nuclides leaching from fuel debris as the parameter for this calculation. Secondly, based on the above calculated inventory, the concentration of those nuclides per vessel was estimated and classified, supposing that the current standard of waste classification for disposal in the normal decommissioning can be applied even for this accident. In this case, the stabilization processing of wastes inside the vessels was not taken into consideration, and these were presumed to be restored directly. By this fitting, the results show that the migration rates of Cs-137 and Cs-134 are 24.4 TBq / day and 10.5 TBq / day. And the classification results indicate that the concentrations of almost all vessels are classified as the equivalent of pit disposal or trench disposal group.

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