Precautions at Fukushima That Would Have Suppressed the Accident Severity

2018 ◽  
Vol 4 (3) ◽  
Author(s):  
Kenji Iino ◽  
Ritsuo Yoshioka ◽  
Masao Fuchigami ◽  
Masayuki Nakao
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Water Injection ◽  
Reactor Core ◽  
Radioactive Material ◽  
Injection System ◽  
Tsunami Waves ◽  
Ac Power ◽  
Pressure Water ◽  
High Pressure Coolant

Abstract The Great East Japan Earthquake on Mar. 11, 2011 triggered huge tsunami waves that attacked Fukushima Daiichi Nuclear Power Plant (Fukushima-1). Units 1, 3, and 4 had hydrogen explosions. Units 1–3 had core meltdowns and released a large amount of radioactive material. Published investigation reports did not explain how the severity of the accident could have been prevented. We formed a study group to find: (A) Was the earthquake-induced huge tsunami predictable at Fukushima-1? (B) If it was predictable, what preparations at Fukushima-1 could have avoided the severity of the accident? Our conclusions were: (a) The tsunami that hit Fukushima-1 was predictable, and (b) the severity could have been avoided if the plant had prepared a set of equipment, and most of all, had exercised actions to take against such tsunami. Necessary preparation included: (1) a number of direct current (DC) batteries, (2) portable underwater pumps, (3) portable alternating current (AC) generators with sufficient gasoline supply, (4) high voltage AC power trucks, and (5) drills against extended loss of all electric power and seawater pumps. This set applied only to this specific accident. A thorough preparation would have added (6) portable compressors, (7) watertight modification to reactor core isolation cooling system (RCIC) and high pressure coolant injection system (HPCI) control and instrumentation, and (8) fire engines for alternate low pressure water injection. Item (5), i.e., to study plans and carry out exercises against the tsunami would have identified all other necessary preparations.

What Could Have Saved Fukushima From Its Severe Accident

2016 ◽  
Author(s):  
Kenji Iino ◽  
Ritsuo Yoshioka ◽  
Masao Fuchigami ◽  
Masayuki Nakao
Keyword(s):  
Electric Power ◽  
Nuclear Power ◽  
Water Injection ◽  
Radioactive Material ◽  
Severe Accident ◽  
Energy Agency ◽  
Event Scale ◽  
Tsunami Waves ◽  
Ac Power ◽  
Utility Companies

The Great East Japan Earthquake on March 11, 2011 triggered huge tsunami waves that devastated the northeast region of Japan along the Pacific coastline. The Tokyo Electric Power Company (TEPCO) owned Fukushima Daiichi Nuclear Power Plant (Fukushima-1) survived the earthquake, however, not the tsunami that followed. Four of the 6 reactor units underwent Station Blackout. Unit 5 lost all its own AC power, however, it shared AC power with Unit 6. Units 1, 3, and 4 had hydrogen explosions that destroyed their reactor buildings, and even worse, 1, 2, and 3 had core meltdowns to release a large amount of radioactive material to their surroundings. The accident was rated Level 7 on the International Nuclear Event Scale, the worst level defined by International Atomic Energy Agency (IAEA). Reports and papers have been published by a number of entities including the Japanese Diet, Government, TEPCO, IAEA, and more. They give detail explanation of how the accident developed into a nuclear disaster explaining the direct and background causes and faults made after the accident broke out. Finding the accident process, i.e., how it happened, and its causes of why it happened, are the most important first steps in accident analysis. Figuring out how to prevent similar events in the future, or even if it is possible to do so, however, is equally important for our future. We started our study in 2014 to find what actions TEPCO could have taken before the accident for preventing it from growing into a catastrophe. Then in February 2015, we set the goal of our study group to find answers to the following two questions: A. Was the huge tsunami, induced by a huge earthquake, predictable at Fukushima-1? B. If it was predictable, what preparations at Fukushima-1 could have reduced the severity of the accident? In response to our invitation to experts in the nuclear field, active and retired people gathered from academia, manufacturers, utility companies, and even regulators. After a series of tense discussions, we reached the conclusions that: Aa. Tsunami of the level that hit Fukushima-1 in 2011 was well predictable, and, Ba. The accident would have been much less severe if the plant had prepared a set of equipment, and most of all, had exercised actions against such tsunami. Preparation at the plant to prevent the severe accident consisted of the following items 1 through 7, and drills in 8: 1. A number of 125Vdc and 250Vdc batteries, 2. Portable underwater pumps, 3. Portable AC generators with sufficient gasoline supply to run the pumps, and 4. High voltage AC power truck This set applied only to this specific accident. For preparing against many other situations that could have taken place at Fukushima-1, we recommend having, in addition, the following equipment and modifications. 5. Portable compressor to drive air-operated valves for venting, 6. Watertight modification to RCIC and HPCI control and instrumentation, 7. Fire engines for alternate low pressure water injection after vent (Fukushima-1 had three). Just making these preparations would not have been sufficient. Activating valves with DC batteries, for example, takes disengaging the regular power supply lines and hooking up the batteries. 8. Drills against extended loss of all electric power and seawater pump This item 8, on and off-site drills was the most important preparation that should had been made. All other necessary preparations to save the plant in such cases would have followed logically.

What Could Have Saved Fukushima From Its Severe Accident

2018 ◽  
Author(s):  
Kenji Iino ◽  
Ritsuo Yoshioka ◽  
Masao Fuchigami ◽  
Masayuki Nakao
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
High Voltage ◽  
Nuclear Power ◽  
Radioactive Material ◽  
Severe Accident ◽  
Study Group ◽  
Tsunami Waves ◽  
Ac Power ◽  
Fukushima Daiichi

The Great East Japan Earthquake on March 11, 2011 triggered huge tsunami waves that attacked Fukushima Daiichi Nuclear Power Plant (Fukushima-1). Units 1, 3, and 4 had hydrogen explosions. Units 1, 2, and 3 had core meltdowns and released a large amount of radioactive material. Published investigation reports did not explain how the severity of the accident could have been prevented. We formed a study group to find what preparations at Fukushima-1 could have avoided the severity of the accident. We concluded that the severity could have been avoided if the plant had prepared a set of equipment, and had exercised actions to take against such tsunami. Necessary preparation included (1) A number of DC batteries, (2) Portable underwater pumps, (3) Portable AC generators with sufficient gasoline supply, (4) High voltage AC power trucks, and (5) Drills against extended loss of all electric power and seawater pumps. The most important preparation was item (5), i.e., to study plans and carry out exercises against huge tsunami. That alone would have identified all other necessary preparations.

Analysis of Accident Progression of Fukushima Daiichi NPPs With SAMPSON Code

2013 ◽  
Author(s):  
Masanori Naitoh ◽  
Marco Pellegrini ◽  
Hideo Mizouchi ◽  
Hiroaki Suzuki ◽  
Hidetoshi Okada
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Reactor Core ◽  
Pressure Vessels ◽  
Severe Accident ◽  
Injection System ◽  
Part Load ◽  
Calculation Results ◽  
Fukushima Daiichi ◽  
Measured Pressure

The Fukushima Daiichi Nuclear Power Plant units 1, 2, and 3 had serious damages due to the huge earthquake and tsunami which occurred on March 11th 2011. Pressure transients in the reactor pressure vessels (RPVs) of the units 1, 2, and 3 were analyzed with the severe accident analysis code, SAMPSON for a few days from the scram until occurrence of depressurization. Since preliminary analysis results with the original SAMPSON showed difference from the measured data, the following phenomena were newly considered in the current analyses. For unit 1: Damage of a source range monitor, which is one of in-core monitors. For unit 2: Part load operation of the reactor core isolation cooling system. For unit 3: Part load operation of the high pressure coolant injection system. The calculation results showed fairly good agreements with the measured pressure data and showed RPV bottom damage for all the units resulting in falling of debris in the core region into the pedestal of the drywell.

Application of TRACE and FRAPTRAN in the Spent Fuel Pool of Chinshan Nuclear Power Plant

2013 ◽  
Vol 479-480 ◽  
pp. 543-547
Author(s):  
Jong Rong Wang ◽  
Hao Tzu Lin ◽  
Wan Yun Li ◽  
Shao Wen Chen ◽  
Chun Kuan Shih
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Safety Issue ◽  
Water Injection ◽  
Safety Analysis ◽  
Spent Fuel ◽  
Spent Fuel Pool ◽  
Study Case

In the nuclear power plant (NPP) safety, the safety analysis of the NPP is very important work. In Fukushima NPP event, due to the earthquake and tsunami, the cooling system of the spent fuel pool failed and the safety issue of the spent fuel pool generated. In this study, the safety analysis of the Chinshan NPP spent fuel pool was performed by using TRACE and FRAPTRAN, which also assumed the cooling system of the spent fuel pool failed. There are two cases considered in this study. Case 1 is the no fire water injection in the spent fuel pool. Case 2 is the fire water injection while the water level of the spent fuel pool uncover the length of fuel rods over 1/3 full length. The analysis results of the case 1 show that the failure of cladding occurs in about 3.6 day. However, the results of case 2 indicate that the integrity of cladding is kept after the fire water injection.

Passive Safety Systems of Advanced Nuclear Power Plant: AP1000

2010 ◽  
Author(s):  
Guohua Yan ◽  
Chen Ye
Keyword(s):  
Nuclear Power ◽  
Human Error ◽  
Cooling System ◽  
Heat Removal ◽  
Injection System ◽  
Passive Safety ◽  
Human Errors ◽  
Power Sources ◽  
Passive Safety Systems ◽  
Safety Systems

In the entire history of commercial nuclear power so far, only two major accidents leading to damage of reactor core have taken place. One is Three Mile Island (TMT) accident (1979), which is caused by a series of human error, and the other is Chernobyl accident (1986), which is due to the combined reason of design defects and human errors. After TMI and Chernobyl accidents, in order to reduce manpower in operation and maintenance and influence of human errors on reactor safety, consideration is given to utilization of passive safety systems. According to the IAEA definition, passive safety systems are based on natural forces, such as convection and gravity, and stored energy, making safety functions less dependent on active systems and operators’ action. Recently, the technology of passive safety has been adopted in many reactor designs, such as AP1000, developed by Westinghouse and EP1000 developed by European vendor, and so on. AP1000 as the first so-called Generation III+ has received the final design approval from US NRC in September 2004, and now being under construction in Sanmen, China. In this paper, the major passive safety systems of AP1000, including passive safety injection system, automatic depressurization system passive residual heat removal system and passive containment cooling system, are described and their responses to a break loss-of-coolant accident (LOCA) are given. Just due to these passive systems’ adoption, the nuclear plant can be able to require no operator action and offsite or onsite AC power sources for at least 72h when one accident occurs, and the core melt and large release frequencies are significantly below the requirement of operating plants and the NRC safety goals.

Self-Propelling Cooling Systems: Back-Fitting Passive Cooling Functions to Existing Nuclear Power Plants

2012 ◽  
Author(s):  
Dominik von Lavante ◽  
Dietmar Kuhn ◽  
Ernst von Lavante
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cooling System ◽  
Electrical Power ◽  
Reactor Core ◽  
Passive Cooling ◽  
Spent Fuel ◽  
Cooling Systems ◽  
Spent Fuel Pool

The present paper describes a back-fit solution proposed by RWE Technology GmbH for adding passive cooling functions to existing nuclear power plants. The Fukushima accidents have high-lighted the need for managing station black-out events and coping with the complete loss of the ultimate heat sink for long time durations, combined with the unavailability of adequate off-site supplies and adequate emergency personnel for days. In an ideal world, a nuclear power plant should be able to sustain its essential cooling functions, i.e. preventing degradation of core and spent fuel pool inventories, following a reactor trip in complete autarchy for a nearly indefinite amount of time. RWE Technology is currently investigating a back-fit solution involving “self-propelling” cooling systems that deliver exactly this long term autarchy. The cooling system utilizes the temperature difference between the hotter reactor core or spent fuel pond with the surrounding ultimate heat sink (ambient air) to drive its coolant like a classical heat machine. The cooling loop itself is the heat machine, but its sole purpose is to merely achieve sufficient thermal efficiency to drive itself and to establish convective cooling (∼2% thermal efficiency). This is realized by the use of a Joule/Brayton Cycle employing supercritical CO2. The special properties of supercritical CO2 are essential for this system to be practicable. Above a temperature of 30.97°C and a pressure of 73.7bar CO2 becomes a super dense gas with densities similar to that of a typical liquid (∼400kg/m3), viscosities similar tothat of a gas (∼3×105Pas) and gas like compressibility. This allows for an extremely compact cooling system that can drive itself on very small temperature differences. The presented parametric studies show that a back-fitable system for long-term spent fuel pool cooling is viable to deliver excess electrical power for emergency systems of approximately 100kW. In temperate climates with peak air temperatures of up to 35°C, the system can power itself and its air coolers at spent fuel pool temperatures of 85°C, although with little excess electrical power left. Different back-fit strategies for PWR and BWR reactor core decay heat removal are discussed and the size of piping, heat exchangers and turbo-machinery are briefly evaluated. It was found that depending on the strategy, a cooling system capable of removing all decay heat from a reactor core would employ piping diameters between 100–150mm and the investigated compact and sealed turbine-alternator-compressor unit would be sufficiently small to be integrated into the piping.

scholarly journals A pre-test analysis of ATLAS SBO with RCP seal leakage using MARS code

2021 ◽  
Vol 7 (4) ◽  
pp. 26-33
Author(s):  
Quang Huy Pham ◽  
Sang Yong Lee ◽  
Seung Jong Oh
Keyword(s):  
Coping Strategies ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Cooling System ◽  
Water Injection ◽  
Cooling Water ◽  
Feed Water ◽  
Test Loop ◽  
Station Blackout

The accident in Fukushima Daiichi nuclear power plants shows the important of developing coping strategies for extended station blackout (SBO) scenarios of the nuclear power plants (NPPs). Many NPPs in United State of America are applying FLEX approach as main coping strategies for extended station blackout (SBO) scenarios. In FLEX strategies, outside water injection to reactor cooling system (RCS) and steam generators (SGs) is considered as an effective method to remove residual heat and maintain the inventory of the systems during the accident. This study presents a pretest calculation using MARS code for the Advanced Thermal-hydraulic Test Loop for Accident Simulation (ATLAS) SBO experiment with RCP seal leakage scenario. In the calculation, the turbinedriven auxiliary feed water pumps (TDAFPs) are firstly used after SBO initiation. Then, the outside cooling water injection method is used for long term cooling. In order to minimize operator actions and satisfy requirements of APR1400 emergency operation procedure (EOP), the SGs Atmospheric Dump Valve (ADV) opening ratio, auxiliary feed water (AFW) and outside cooling water injection flow rates were investigated to have suitable values. The analysis results would be useful for performing the experiment to verify the APR 1400 extended SBO optimum mitigation strategy using outside cooling water injection.

scholarly journals Early quantification of reliability for a safety critical and control system: a case study of reactor core cooling system

2018 ◽  
Vol 7 (2.12) ◽  
pp. 248
Author(s):  
Vinay Kumar ◽  
Suraj Gupta ◽  
Anil Kumar Tripathi
Keyword(s):  
Control System ◽  
Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Cooling System ◽  
Reactor Core ◽  
Safety Critical ◽  
Core Cooling ◽  
And Control

Using Probabilistic Reliability analysis for Quantifying reliability of a system is already a common practice in Reliability Engineering community. This method plays an important role in analyzing reliability of nuclear plants and its various components. In Nuclear Power Plants Reactor Core Cooling System is a component of prime importance as its breakdown can disrupt Cooling System of power plant. In this paper, we present a framework for early quantification of Reliability and illustrated with a Safety Critical and Control System as case study which runs in a Nuclear Power Plant.  

Fukushima Unit 2 Accident Simulation With MELCOR 2.1

2016 ◽  
Author(s):  
Rafael Bocanegra ◽  
Valentino Di Marcello ◽  
Victor Sanchez ◽  
Gonzalo Jiménez
Keyword(s):  
Failure Modes ◽  
Cooling System ◽  
Water Injection ◽  
Reactor Core ◽  
Research Centre ◽  
Two Phase ◽  
Final State ◽  
Fukushima Daiichi ◽  
Alternative Water ◽  
Improved Model

A Fukushima Daiichi Unit 3 MELCOR model from Technical Research Centre of Finland (VTT) was modified to simulate the Fukushima Daiichi Unit 2 accident. Several simulations were performed using three different modeling approaches. The base model (1F2 v1) includes only the basic modifications to reproduce the accident. The intermediate model (1F2 v2) includes an improved model of the Wet Well. In the advanced model (1F2 v3), the reactor core isolation cooling system logic was modified to avoid the use of tabular functions for the mass flow inlet and outlet. As a result of this analysis it is concluded that there is a strong dependency on parameters which still have many uncertainties, such as the reactor core isolation cooling system two-phase flow operation, the alternative water injection, the suppression pool behavior, the rupture disk behavior and the containment failure modes which affect the final state of the reactor core.

Development of a Water Injection System Fusible Plug for Severe Accidents at Nuclear Power Plants

2020 ◽  
Author(s):  
Sei Hirano ◽  
Daisuke Hirasawa ◽  
Yoshihisa Kiyotoki ◽  
Keisuke Sakemura ◽  
Keiji Sasaki ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
High Reliability ◽  
Water Injection ◽  
Severe Accident ◽  
Injection System ◽  
Normal Operation ◽  
Autonomous Operation ◽  
Performance Requirements

Abstract Background: When terminal stage of Severe Accident (SA) with no coolant injection at a nuclear power plant, the equipment that has cooled and solidified through water injection to a molten core that has ex-vessel and fallen outside of the pressure vessel will then be required to operate autonomously by heat detection, without external signals or power (e.g. electricity, air). The fusible plug operation is triggered by fusible alloy which receives heat from molten core and will melt. Because the fusible plug is also the boundary of Suppression Pool (S/P), high reliability is required for sealing performance. It is for that reason that Hitachi GE Nuclear Energy Ltd. (Hitachi-GE) has developed a fusible plug to serve as a device necessary to operate this system. Features of the Fusible Plug: The autonomous operation of the fusible plug is triggered by the melting of a fusible alloy, which is part of the fusible plug. However, the fusible alloy has a remarkably low mechanical strength and therefore is not suitable as a strength member. As such, it is necessary to ensure reliable plug sealing without applying a load to the fusible alloy so as to prevent the fusible plug from malfunctioning during normal operation. Therefore, to reduce the load to be applied to the fusible alloy, Hitachi-GE has developed a fusible plug structure that operates autonomously by detecting the ambient temperature without using the fusible alloy as a strength member. We have performed a verification test using this fusible plug and confirmed that it satisfies the predetermined performance requirements. Future Actions: Hitachi-GE is holding discussions on using the fusible plug at nuclear power plants in Japan. In the future, we plan to expand to the overseas.

Sign in / Sign up

Export Citation Format

Share Document