Thermal Loading Failure Analysis of Molten Pool in the Reactor Vessel

2010 ◽  
Author(s):  
Xiao Yang ◽  
Yanhua Yang ◽  
Bo Kuang ◽  
Xiaoliang Fu
Keyword(s):  
Heat Flux ◽  
Molten Pool ◽  
Nuclear Power ◽  
Power Plants ◽  
Thermal Loading ◽  
Success Probability ◽  
Local Heat ◽  
Reactor Vessel ◽  
Transfer Model ◽  
Severe Accidents

Uncertainties are addressed in the special context of assessing and managing risks from rare, severe-consequence hazards. Risk Oriented Accident Analysis Methodology (ROAAM) is used to analyze uncertainties during severe accidents analysis in nuclear power plants. In-vessel Retention (IVR) is one of the mitigations for severe accidents which will cause core damage. By external reactor vessel cooling (ERVC), the integrity of the reactor vessel is preserved. The success criterion for IVR is the local heat flux on the wall of lower head is less than the critical heat flux (CHF). This paper analyzes the uncertain parameters which decide the mitigation to be successful or fail. Two bounding structures and 4 molten pool steady states are defined. And the success probability of IVR is evaluated with a molten pool heat transfer model. Then the effectiveness of IVR-ERVC under the two bounding structures is evaluated.

scholarly journals Development of Instrument Transmitter Protecting Device against High-Temperature Condition during Severe Accidents

2014 ◽  
Vol 2014 ◽  
pp. 1-7 ◽  
Author(s):  
Min Yoo ◽  
Sung Min Shin ◽  
Hyun Gook Kang
Keyword(s):  
Heat Transfer ◽  
High Temperature ◽  
Nuclear Power ◽  
Power Plants ◽  
Severe Accident ◽  
Transfer Model ◽  
Potential Threat ◽  
Severe Accidents ◽  
Development Of Instrument ◽  
Novel Concept

Reliable information through instrumentation systems is essential in mitigating severe accidents such as the one that occurred at the Fukushima Daiichi nuclear power plant. There are five elements which might pose a potential threat to the reliability of parameter detection at nuclear power plants during a severe accident: high temperature, high pressure, high humidity, high radiation, and missiles generated during the evolution of a severe accident. Of these, high temperature apparently poses the most serious threat, since thin shielding can get rid of pressure, humidity, radiation (specifically, alpha and beta radiations), and missile effects. In view of this fact, our study focused on designing an instrument transmitter protecting device that can eliminate the high-temperature effect on transmitters to maintain their functional integrity. We present herein a novel concept for designing such a device in terms of heat transfer model that takes into account various heat transfer mechanisms associated with the device.

Design analysis of the molten core confinement within the reactor vessel in the case of severe accidents at nuclear power plants equipped with a reactor of the VVER type

2013 ◽  
Vol 76 (14) ◽  
pp. 1688-1699
Author(s):  
Yu. A. Zvonaryov ◽  
M. A. Budaev ◽  
A. M. Volchek ◽  
V. A. Gorbaev ◽  
V. N. Zagryazkin ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Design Analysis ◽  
Reactor Vessel ◽  
Severe Accidents

scholarly journals Reactor Vessel Internals Segmentation Experience using Mechanical Cutting Tools

2013 ◽  
Vol 10 (2) ◽  
pp. 6-10 ◽  
Author(s):  
Petr Pospíšil
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cutting Tools ◽  
Reactor Vessel ◽  
Boiling Water Reactors ◽  
Abrasive Waterjet ◽  
Pressurized Water ◽  
Water Reactors ◽  
Mechanical Cutting ◽  
Reactor Internals

Abstract Some commercial nuclear power plants have been permanently shut down to date and decommissioned using dismantling methods. Other operating plants have decided to undergo an upgrade process that includes replacement of reactor internals. In both cases, there is a need to perform a segmentation of the reactor vessel internals with proven methods for long term waste disposal. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive waterjet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Mechanical cutting has been used by Westinghouse since 1999 for both Pressurized Water Reactors (PWR’s) and Boiling Water Reactors (BWR’s) and its process has been continuously improved over the years. The complexity of the work requires well designed and reliable tools. Different band saws, disc saws, tube cutters and shearing tools have been developed to cut the reactor internals. All of those equipments are hydraulically driven which is very suitable for submerged applications. Westinghouse experience in mechanical cutting has demonstrated that it is an excellent technique for segmentation of internals. In summary, the purpose of this paper will be to provide an overview of the Westinghouse mechanical segmentation process, based on actual experience from the work that has been completed to date.

scholarly journals Spreading of Excellence in SARNET Network on Severe Accidents: The Education and Training Programme

2012 ◽  
Vol 2012 ◽  
pp. 1-9 ◽  
Author(s):  
Sandro Paci ◽  
Jean-Pierre Van Dorsselaere
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Training Programme ◽  
Education And Training ◽  
Severe Accident ◽  
Research Network ◽  
Safety Issues ◽  
Severe Accidents ◽  
Management Procedures ◽  
And Training

The SARNET2 (severe accidents Research NETwork of Excellence) project started in April 2009 for 4 years in the 7th Framework Programme (FP7) of the European Commission (EC), following a similar first project in FP6. Forty-seven organisations from 24 countries network their capacities of research in the severe accident (SA) field inside SARNET to resolve the most important remaining uncertainties and safety issues on SA in water-cooled nuclear power plants (NPPs). The network includes a large majority of the European actors involved in SA research plus a few non-European relevant ones. The “Education and Training” programme in SARNET is a series of actions foreseen in this network for the “spreading of excellence.” It is focused on raising the competence level of Master and Ph.D. students and young researchers engaged in SA research and on organizing information/training courses for NPP staff or regulatory authorities (but also for researchers) interested in SA management procedures.

scholarly journals Smart support system for diagnosing severe accidents in nuclear power plants

2018 ◽  
Vol 50 (4) ◽  
pp. 562-569 ◽  
Author(s):  
Kwae Hwan Yoo ◽  
Ju Hyun Back ◽  
Man Gyun Na ◽  
Seop Hur ◽  
Hyeonmin Kim
Keyword(s):  
Support System ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Severe Accidents

Analysis of Focusing Effect of Light Metallic Layer in Stratified Molten Pool Under IVR-ERVC Condition

2015 ◽  
Vol 1 (2) ◽  
Author(s):  
Xi Wang ◽  
Xu Cheng
Keyword(s):  
Heat Flux ◽  
Molten Pool ◽  
Fluid Dynamic ◽  
Metal Layer ◽  
Local Heat ◽  
Metallic Layer ◽  
Parameter Method ◽  
Focusing Effect ◽  
Local Heat Flux ◽  
Reactor Pressure

The main failure mechanism of in-vessel corium retention through external reactor vessel cooling (IVR-ERVC) happens when the local heat flux through reactor pressure vessel (PRV) wall exceeds the critical heat flux (CHF). High local heat flux in the molten pool is usually caused by the metallic layer focusing effect due to stratification. In this paper, depending on experimental correlations, both the lump parameter method and computational fluid dynamic method are used to investigate the mechanism of focusing effect. The concentration factor varying with the height of metallic layer is studied. The results show that the lump parameter method probably overestimates the wall heat flux of metal layer.

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.

Analysis of Transient Corium Pool Structure in the Lower Plenum of Reactor Vessel

2020 ◽  
Author(s):  
Liang Chen ◽  
Hua Pang ◽  
Ximing Xie ◽  
Lei Zhong ◽  
Rong Cai
Keyword(s):  
Heavy Metal ◽  
Heat Flux ◽  
Nuclear Power ◽  
Metal Layer ◽  
Reactor Vessel ◽  
Loss Of Coolant Accident ◽  
Density Evaluation ◽  
Station Blackout ◽  
Molten Materials ◽  
The Moment

Abstract The transient stratification of the corium in the lower plenum and its impact on the heat flux distribution on the outside of reactor vessel is analyzed in this work. A method for predicting the kinetic corium pool structure is proposed, which takes into account both thermo-chemical equilibrium and density evaluation of the corium. The transient stratification of the corium pool formed after a large loss of coolant accident (LLOCA) and a station blackout (SBO) accident of ACP1000 nuclear power plant in China was analyzed by this method. The transient structure of the corium pool was calculated at the moment when the amount of molten materials in the corium pool increased obviously. The results shown that the formation of a three-layer pool is highly possible when a two-layer pool is formed in the previous moment with a heavy metal layer on the bottom and the density of the heavy metal layer at the bottom is greater than the density of the newly added molten material at the next moment. The heat flux on the outside of the vessel wall faced the thin top metal layer and the vessel failure probability of the vessel here are high if a three-layer pool occurred.

Sign in / Sign up

Export Citation Format

Share Document