scholarly journals Unitization for portability of emergency response surveillance robot system: experiences and lessons learned from the deployment of the JAEA-3 emergency response robot at the Fukushima Daiichi Nuclear Power Plants

2017 ◽  
Vol 4 (1) ◽  
Author(s):  
Shinji Kawatsuma ◽  
Ryuji Mimura ◽  
Hajime Asama
Keyword(s):  
Emergency Response ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Lessons Learned ◽  
Robot System ◽  
Fukushima Daiichi

Emergency response to the nuclear accident at the Fukushima Daiichi Nuclear Power Plants using mobile rescue robots

2012 ◽  
Vol 30 (1) ◽  
pp. 44-63 ◽  
Author(s):  
Keiji Nagatani ◽  
Seiga Kiribayashi ◽  
Yoshito Okada ◽  
Kazuki Otake ◽  
Kazuya Yoshida ◽  
...  
Keyword(s):  
Emergency Response ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Accident ◽  
Fukushima Daiichi ◽  
Rescue Robots

A Remote Operated Quadruped Robot System for Investigation of Reactor Building

2014 ◽  
Author(s):  
Naotaka Suganuma ◽  
Takuya Uehara ◽  
Kenji Matsuzaki ◽  
Makoto Ochiai ◽  
Fujio Terai ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Control Method ◽  
Quadruped Robot ◽  
Polluted Environment ◽  
Robot System ◽  
Operation System ◽  
Reactor Building ◽  
Fukushima Daiichi

A remote operated quadruped robot has been developed for disaster site which can move on stairs, slopes, and uneven floor under the radiation-polluted environment, such as TEPCO Fukushima Daiichi nuclear power plants [1][2]. In particular, the control method for stable walking and the remote operation system have been developed to move on stairs in the reactor building. We applied this robot to investigation of suspicious water leakage points in reactor building at Fukushima Daiichi nuclear power plants unit2[3]. In this investigation, a small vehicle equipped with camera and a manipulator which is connected the vehicle with cable were mounted on the robot and were carried to near the target by the quadruped robot and the investigation was carried out with the small vehicle.

Identification of Non-Safety Systems to Withstand Beyond Design Basis Events

2012 ◽  
Author(s):  
James Nygaard ◽  
Ping Wan ◽  
Desmond Chan ◽  
Sara Barrientos
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Systems Approach ◽  
Lessons Learned ◽  
Initial Assessment ◽  
Stress Tests ◽  
Multiple Systems ◽  
Design Basis ◽  
Fukushima Daiichi

As an aftermath of the natural disasters affecting the Fukushima Daiichi nuclear power plants in Japan, there has been great attention to provide assurance of safety of nuclear power plants around the world. Accordingly, many countries are requiring “stress tests” for their plants to assess the ability to withstand disaster scenarios for which they were not originally designed. Additional efforts are underway to capture lessons learned related to the operation of critical or major systems. Each operator and each country’s regulatory authority may be at different levels of completion for these activities. However, effects on non-safety related or peripheral systems have not been specifically addressed as standalone items or in an integrated systems approach. This paper seeks to produce an initial assessment of vulnerable systems, structures or components of non-safety related areas that may become critical to the safe operation of a nuclear plant or to the first steps to maintain stability of the plant during a postulated beyond design basis event. The same assessment is valid for events of significant magnitude, or for events affecting the entire site or region, even if a plant’s design basis is not exceeded. The initial assessment is based on widespread events, such as at the Fukushima Daiichi station, with focus on large nuclear power reactors. Certain peripheral plant systems support plant operators and staff or emergency responders such as by affording communication or physical access to plant areas. Other peripheral systems support plant operation or recovery, for example provision of diverse power supply or cooling means. Passive components common to multiple systems such as cables and piping are also assessed. Once vulnerable systems, structures or components are identified, various modifications or mitigation approaches will be discussed.

Expectations for Inservice Testing Programs at New Nuclear Power Plants

2017 ◽  
Author(s):  
Thomas G. Scarbrough
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Reactor Core ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Mechanical Equipment ◽  
American Society ◽  
Regulatory Commission ◽  
Safety Systems

In a series of Commission papers, the U.S. Nuclear Regulatory Commission (NRC) described its policy for inservice testing (IST) programs to be developed and implemented at nuclear power plants licensed under 10 CFR Part 52. This paper discusses the expectations for IST programs based on those Commission policy papers as applied in the NRC staff review of combined license (COL) applications for new reactors. For example, the design and qualification of pumps, valves, and dynamic restraints through implementation of American Society of Mechanical Engineers (ASME) Standard QME-1-2007, “Qualification of Active Mechanical Equipment Used in Nuclear Power Plants,” as accepted in NRC Regulatory Guide (RG) 1.100 (Revision 3), “Seismic Qualification of Electrical and Active Mechanical Equipment and Functional Qualification of Active Mechanical Equipment for Nuclear Power Plants,” will enable IST activities to assess the operational readiness of those components to perform their intended functions. ASME has updated the Operation and Maintenance of Nuclear Power Plants (OM Code) to improve the IST provisions for pumps, valves, and dynamic restraints that are incorporated by reference in the NRC regulations with applicable conditions. In addition, lessons learned from performance experience and testing of motor-operated valves (MOVs) will be implemented as part of the IST programs together with application of those lessons learned to other power-operated valves (POVs). Licensee programs for the Regulatory Treatment of Non-Safety Systems (RTNSS) will be implemented for components in active nonsafety-related systems that are the first line of defense in new reactors that rely on passive systems to provide reactor core and containment cooling in the event of a plant transient. This paper also discusses the overlapping testing provisions specified in ASME Standard QME-1-2007; plant-specific inspections, tests, analyses, and acceptance criteria; the applicable ASME OM Code as incorporated by reference in the NRC regulations; specific license conditions; and Initial Test Programs as described in the final safety analysis report and applicable RGs. Paper published with permission.

Nuclear Power Plant Fires and Explosions: Part IV — Water Hammer Explosion Mechanisms

2017 ◽  
Author(s):  
Robert A. Leishear
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Water Hammer ◽  
Compression Process ◽  
Flammable Gas ◽  
Piping Systems ◽  
Fukushima Daiichi ◽  
Accident Conditions

Water hammers, or fluid transients, compress flammable gasses to their autognition temperatures in piping systems to cause fires or explosions. While this statement may be true for many industrial systems, the focus of this research are reactor coolant water systems (RCW) in nuclear power plants, which generate flammable gasses during normal operations and during accident conditions, such as loss of coolant accidents (LOCA’s) or reactor meltdowns. When combustion occurs, the gas will either burn (deflagrate) or explode, depending on the system geometry and the quantity of the flammable gas and oxygen. If there is sufficient oxygen inside the pipe during the compression process, an explosion can ignite immediately. If there is insufficient oxygen to initiate combustion inside the pipe, the flammable gas can only ignite if released to air, an oxygen rich environment. This presentation considers the fundamentals of gas compression and causes of ignition in nuclear reactor systems. In addition to these ignition mechanisms, specific applications are briefly considered. Those applications include a hydrogen fire following the Three Mile Island meltdown, hydrogen explosions following Fukushima Daiichi explosions, and on-going fires and explosions in U.S nuclear power plants. Novel conclusions are presented here as follows. 1. A hydrogen fire was ignited by water hammer at Three Mile Island. 2. Hydrogen explosions were ignited by water hammer at Fukushima Daiichi. 3. Piping damages in U.S. commercial nuclear reactor systems have occurred since reactors were first built. These damages were not caused by water hammer alone, but were caused by water hammer compression of flammable hydrogen and resultant deflagration or detonation inside of the piping.

scholarly journals Reactor performance, system reliability, instrumentation and control

2020 ◽  
Vol 6 ◽  
pp. 43
Author(s):  
Andreas Schumm ◽  
Madalina Rabung ◽  
Gregory Marque ◽  
Jary Hamalainen
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Lessons Learned ◽  
Software Systems ◽  
Reactor Performance ◽  
Horizon 2020 ◽  
Term Operation ◽  
In The Beginning ◽  
Electrical Cables

We present a cross-cutting review of three on-going Horizon 2020 projects (ADVISE, NOMAD, TEAM CABLES) and one already finished FP7 project (HARMONICS), which address the reliability of safety-relevant components and systems in nuclear power plants, with a scope ranging from the pressure vessel and primary loop to safety-critical software systems and electrical cables. The paper discusses scientific challenges faced in the beginning and achievements made throughout the projects, including the industrial impact and lessons learned. Two particular aspects highlighted concern the way the projects sought contact with end users, and the balance between industrial and academic partners. The paper concludes with an outlook on follow-up issues related to the long term operation of nuclear power plants.

Post-Fukushima Research and Development Strategies and Priorities for Water Cooled Reactor Technology Development

2016 ◽  
Author(s):  
Katsumi Yamada ◽  
Abdallah Amri ◽  
Lyndon Bevington ◽  
Pal Vincze
Keyword(s):  
Research And Development ◽  
International Organizations ◽  
Nuclear Power ◽  
Power Plants ◽  
Development Strategies ◽  
Lessons Learned ◽  
Member States ◽  
North East ◽  
Severe Accidents ◽  
Fukushima Daiichi

The Great East Japan Earthquake and the subsequent tsunami on 11 March 2011 initiated accident conditions at several nuclear power plants (NPPs) on the north-east coast of Japan and developed into a severe accident at the Fukushima Daiichi NPP, which highlighted a number of nuclear safety issues. After the Fukushima Daiichi accident, new research and development (R&D) activities have been undertaken by many countries and international organizations relating to severe accidents at NPPs. The IAEA held, in cooperation with the OECD/NEA, the International Experts’ Meeting (IEM) on “Strengthening Research and Development Effectiveness in the Light of the Accident at the Fukushima Daiichi Nuclear Power Plant” at IAEA Headquarters in Vienna, Austria, 16–20 February 2015. The objective of the IEM was to facilitate the exchange of information on these R&D activities and to further strengthen international collaboration among Member States and international organizations. One of the main conclusions of the IEM was that the Fukushima Daiichi accident had not identified completely new phenomena to be addressed, but that the existing strategies and priorities for R&D should be reconsidered. Significant R&D activities had been already performed regarding severe accidents of water cooled reactors (WCRs) before the accident, and the information was very useful for predicting and understanding the accident progression. However, the Fukushima Daiichi accident highlighted several challenges that should be addressed by reconsidering R&D strategies and priorities. Following this IEM, the IAEA invited several consultants to IAEA Headquarters, Vienna, Austria, 12–14 May 2015, and held a meeting in order to discuss proposals on possible IAEA activities to facilitate international R&D collaboration in relation to severe accidents and how to effectively disseminate the information obtained at the IEM. The IAEA also held Technical Meeting (TM) on “Post-Fukushima Research and Development Strategies and Priorities” at IAEA Headquarters, Vienna, Austria, 15–18 December 2015. The objective of this meeting was to provide a platform for experts from Member States and international organizations to exchange perspectives and information on strategies and priorities for R&D regarding the Fukushima Daiichi accident and severe accidents in general. The experts discussed R&D topic areas that need further attention and the benefits of possible international cooperation. This paper discusses lessons learned from the Fukushima Daiichi accident based on the presentations and discussions at the meetings mentioned above, and identifies the needs for further R&D activities to develop WCR technologies to cope with Fukushima Daiichi-type accidents.

Sign in / Sign up

Export Citation Format

Share Document