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.

Analysis of Accident Progression of Fukushima Daiichi NPP by the SAMPSON Code

2014 ◽  
Author(s):  
Masanori Naitoh ◽  
Marco Pellegrini ◽  
Hiroaki Suzuki ◽  
Hideo Mizouchi ◽  
Hidetoshi Okada
Keyword(s):  
Nuclear Power ◽  
Water Injection ◽  
Heat Removal ◽  
Fission Products ◽  
Severe Accident ◽  
Core Material ◽  
Decay Heat ◽  
The Core ◽  
Fukushima Daiichi ◽  
Fire Engine

This paper describes analysis results of the early phase accident progression of the Fukushima Daiichi Nuclear Power Plant (NPP) Unit 1 by the severe accident analysis code SAMPSON. The isolation condensers were the only devices for decay heat removal at Unit 1, but they stopped after the loss of AC and DC powers. Since there were no decay heat removal for about 14 hours after their termination until the start of alternative water injection into the core by the fire engine, the core melt and the reactor pressure vessel (RPV) bottom failure occurred resulting in large amount of fission products release into the environment. The original SAMPSON was improved by adding new modellings for the phenomena which have been deemed specific to the Fukushima Daiichi NPP: (1) deterioration of SRV gaskets and (2) buckling of in-core-monitor housings which caused the early steam leakage from the core into the drywell, and (3) melt of the in-core-monitor housings in the lower plenum of the RPV. The analysis results showed that (1) 55.3% of UO2 of the initial loading and 66.1% of the core material including UO2, zircaloy, steel and control materials had melted down into the pedestal of the drywell, (2) the amount of Hydrogen generated by Zr-H2O reaction was 686 kg, (3) amount of Cs element released from fuels was 61 kg which was 72% of the total Cs element which was included in fuels at the initiation of the accident, and (4) 18.3% of the corium which fell into the pedestal was one large lump and the 81.7% was particulate corium.

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.

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.

scholarly journals Investigation on Main Radiation Source at Operation Floor of Fukushima Daiichi Nuclear Power Station Unit 4

2017 ◽  
Vol 153 ◽  
pp. 08011
Author(s):  
Hideo Hirayama ◽  
Kenjiro Kondo ◽  
Seishiro Suzuki ◽  
Shimpei Hamamoto ◽  
Kohei Iwanaga
Keyword(s):  
Nuclear Power Station ◽  
Radiation Source ◽  
Nuclear Power ◽  
Power Station ◽  
Fukushima Daiichi ◽  
Station Unit
Sign in / Sign up

Export Citation Format

Share Document