scholarly journals Radiation-Tolerance Assessment of a Redundant Wireless Device

2018 ◽  
Vol 170 ◽  
pp. 03004
Author(s):  
Q. Huang ◽  
J. Jiang
Keyword(s):  
Nuclear Power ◽  
Evaluation Method ◽  
Radiation Condition ◽  
Severe Accident ◽  
Radiation Environment ◽  
Radiation Tolerance ◽  
Monitoring Device ◽  
Wireless Monitoring ◽  
Study Results ◽  
Accident Conditions

This paper presents a method to evaluate radiation-tolerance without physical tests for a commercial off-the-shelf (COTS)-based monitoring device for high level radiation fields, such as those found in post-accident conditions in a nuclear power plant (NPP). This paper specifically describes the analysis of radiation environment in a severe accident, radiation damages in electronics, and the redundant solution used to prolong the life of the system, as well as the evaluation method for radiation protection and the analysis method of system reliability. As a case study, a wireless monitoring device with redundant and diversified channels is evaluated by using the developed method. The study results and system assessment data show that, under the given radiation condition, performance of the redundant device is more reliable and more robust than those non-redundant devices. The developed redundant wireless monitoring device is therefore able to apply in those conditions (up to 10 M Rad (Si)) during a severe accident in a NPP.

Research about the Hydrogen Removal Verification Method through Porous Medium of Silver Zeolite in Nuclear Power Station

2020 ◽  
Vol 999 ◽  
pp. 31-38
Author(s):  
Qing Bo Bao ◽  
Jian Hu ◽  
Shao Fei Zhou
Keyword(s):  
Nuclear Power Station ◽  
Nuclear Power ◽  
Evaluation Method ◽  
Natural Circulation ◽  
Catalyst Layer ◽  
Severe Accident ◽  
Test Device ◽  
Power Station ◽  
Verification Method ◽  
Silver Zeolite

After severe accident in the nuclear power station, it is necessary to remove the hydrogen timely for the purpose of preventing the containment integrity from breach. This report has investigated and studied the role of silver zeolite in the reaction of hydrogen and oxygen. According to the catalyst role, the principle test device for hydrogen removal with silver zeolite is provided. The force of natural circulation for principle test device is created by the Chimney Effect, which is the result of different density between the internal and external of the device. Also, this report suggests the flowing capability calculation method of up-thrust about the mixture gas passing through the catalyst layer of silver zeolite. The evaluation method of hydrogen removal efficiency with silver zeolite is described. Finally, this report gives the method of CFX numeric analog and the specific simulating steps for the layer of silver zeolite using for catalytic role.

scholarly journals Steam Oxidation of Silicon Carbide at High Temperatures for the Application as Accident Tolerant Fuel Cladding, an Overview

Thermo ◽  
2021 ◽  
Vol 1 (2) ◽  
pp. 151-167
Author(s):  
Hai V. Pham ◽  
Masaki Kurata ◽  
Martin Steinbrueck
Keyword(s):  
Silicon Carbide ◽  
High Temperatures ◽  
Nuclear Power ◽  
Steam Oxidation ◽  
Severe Accident ◽  
Light Water Reactors ◽  
Severe Accidents ◽  
Potential Benefits ◽  
Water Reactors ◽  
Accident Conditions

Since the nuclear accident at Fukushima Daiichi Nuclear Power Station in 2011, a considerable number of studies have been conducted to develop accident tolerant fuel (ATF) claddings for safety enhancement of light water reactors. Among many potential ATF claddings, silicon carbide is one of the most promising candidates with many superior features suitable for nuclear applications. In spite of many potential benefits of SiC cladding, there are some concerns over the oxidation/corrosion resistance of the cladding, especially at extreme temperatures (up to 2000 °C) in severe accidents. However, the study of SiC steam oxidation in conventional test facilities in water vapor atmospheres at temperatures above 1600 °C is very challenging. In recent years, several efforts have been made to modify existing or to develop new advanced test facilities to perform material oxidation tests in steam environments typical of severe accident conditions. In this article, the authors outline the features of SiC oxidation/corrosion at high temperatures, as well as the developments of advanced test facilities in their laboratories, and, finally, give some of the current advances in understanding based on recent data obtained from those advanced test facilities.

Development and Verification of Severe Accident Simulation Capability for a NPP Simulator

2013 ◽  
Author(s):  
Zhifei Yang ◽  
Xiaofei Xie ◽  
Xing Chen ◽  
Shishun Zhang ◽  
Yehong Liao ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Simulation System ◽  
Severe Accident ◽  
Normal Operation ◽  
Feed Water ◽  
Accident Simulation ◽  
Accident Conditions ◽  
Relap5 Code

It is reflected in the severe accident in Fukushima Daiichi that the emergency capacity of nuclear power plant needs to be enhanced. A nuclear plant simulator that can model the severe accident is the most effective means to promote this capacity. Until now, there is not a simulator which can model the severe accident in China. In order to enhance the emergency capacity in China, we focus on developing a full scope simulator that can model the severe accident and verify it in this study. The development of severe accident simulation system mainly includes three steps. Firstly, the integral severe accident code MELCOR is transplanted to the simulation platform. Secondly, the interface program must be developed to switch calculating code from RELAP5 code to MELCOR code automatically when meeting the severe accident conditions because the RELAP5 code can only simulate the nuclear power plant normal operation state and design basis accident but the severe accident. So RELAP5 code will be stopped when severe accident conditions happen and the current nuclear power plant state parameters of it should be transported to MELCOR code, and MELCOR code will run. Finally, the CPR1000 nuclear power plant MELCOR model is developed to analyze the nuclear power plant behavior in severe accident. After the three steps, the severe accident simulation system is tested by a scenario that is initiated by the station black out with reactor cooling pump seal leakage, HHSI, LHSI and auxiliary feed water system do not work. The simulation result is verified by qualitative analysis and comparison with the results in severe accident analysis report of the same NPP. More severe accident scenarios initiated by LBLOCA, MBLOCA, SBLOCA, SBO, ATWS, SGTR, MSLB will be tested in the future. The results show that the severe accident simulation system can model the severe accident correctly; it meets the demand of emergency capacity promotion.

Evaluation Method of Response Reliability During an Accident and its Applicability to Fast Reactor Plants

2018 ◽  
Author(s):  
Masaaki Suzuki ◽  
Kazuyuki Demachi ◽  
Shigeru Takaya ◽  
Yoshitaka Chikazawa
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Fast Reactor ◽  
Nuclear Power ◽  
Evaluation Method ◽  
Severe Accident ◽  
Evaluation Procedure ◽  
System Safety ◽  
Recovery Capacity ◽  
Safety Function

In this study, we develop new indices that evaluate the response capability — response margin and response reliability — of a nuclear power plant in case of occurrence of a severe accident by considering the ability of the nuclear power plant to recover system safety functions, which may have been affected by the accident, to the minimum safety values. Further, we demonstrate a specific evaluation procedure for our proposed indices. While performing the evaluation, we consider margins with respect to a minimum safety function level and a time constraint along with their temporal changes. Additionally, for a simplified fast reactor (FR) plant model, we conduct a trial assessment of the recovery capacity of the system to ensure system safety based on the concept of new indices. Further, the applicability of the response reliability evaluation method to FR plants is discussed based on the viewpoints of reflecting the characteristics of each reactor type.

Study of Equipment Survivability Under Severe Accident Conditions

2016 ◽  
Author(s):  
Jinquan Yan ◽  
Shanhu Xue ◽  
Lin Tian ◽  
Wei Lu
Keyword(s):  
Nuclear Power ◽  
Severe Accident ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Plant Safety ◽  
New Development ◽  
Nuclear Power Plant Safety ◽  
Accident Conditions ◽  
Accident Sequences

To improve nuclear power plant safety, severe accident prevention and mitigation for both new development and existing plants are generally required by various nuclear safety authorities worldwide. Although great efforts have been made, how to ensure equipment survivability under severe accident conditions is still a concern. This paper depicts an approach to demonstrate the equipment survivability under severe accident conditions by taking passive pressurized water reactor CAP1400 as an instance, including screening of severe accident sequences, determination of bounding environment conditions within containment, equipments identification used for severe accident mitigation and proposed test plan.

IRSN R&D Actions on FP Behaviour for RCS, Containment and FCVS in Severe Accident Conditions

2016 ◽  
Author(s):  
Laurent Cantrel ◽  
Thierry Albiol ◽  
Loïc Bosland ◽  
Juliette Colombani ◽  
Frédéric Cousin ◽  
...  
Keyword(s):  
Porous Materials ◽  
Nuclear Power ◽  
Source Term ◽  
Cooling System ◽  
Fission Products ◽  
Simulation Software ◽  
Severe Accident ◽  
Status Report ◽  
Diagnosis And Prognosis ◽  
Accident Conditions

This paper deals with near past, ongoing and planned R&D works on fission products (FPs) behaviour in Reactor Cooling System (RCS), containment building and in Filtered Containment Venting Systems (FCVS) for severe accident (SA) conditions. For the last topic, in link with the Fukushima post-accident management and possible improvement of mitigation actions for such SA, the FCVS topic is again on the agenda (see Status Report on Filtered Containment Venting, OECD/NEA/CSNI, Report NEA/CSNI/R(2014)7, 2014.) with a large interest at the international scale. All the researches are collaborative works; the overall objective is to develop confident models to be implemented in ASTEC SA simulation software. After being initiated in the International Source Term Program (ISTP), researches devoted to the understanding of iodine transport through the RCS are still ongoing in the frame of a bilateral agreement between IRSN and EDF with promising results. In 2017, a synthesis report of the last 10 years of researches, which have combined experimental and fundamental works based on the use of theoretical chemistry tools, will be issued. For containment, the last advances are linked to the Source Term Evaluation and Mitigation (STEM) OECD/NEA project operated by IRSN. The objective of the STEM project was to improve the evaluation of Source Term (ST) for a SA on a nuclear power plant and to reduce uncertainties on specific phenomena dealing with the chemistry of two major fission products: iodine and ruthenium. More precisely, the STEM project provided additional knowledge and improvements for calculation tools in order to allow a more robust diagnosis and prognosis of radioactive releases in a SA. STEM data will be completed by a follow-up, called STEM2, to further the knowledge concerning some remaining issues and be closer to reactor conditions. Two additional programmes deal with FCVS issues: the MItigation of outside Releases in the Environment (MIRE) (2013–2019) French National Research Agency (NRA) programme and the Passive and Active Systems on Severe Accident source term Mitigation (PASSAM) (2013–2016) European project. For FCVS works, the efficiencies for trapping iodine with various FCVS, covering scrubbers and dry filters, are examined to get a clear view of their abilities in SA conditions. Another part, performed in collaboration with French universities (Lorraine and Lille 1), is focused on the enhancement of the performance of these filters with specific porous materials able to trap volatile iodine. For that, influence of zeolites materials parameters (nature of the counter-ions, structure, Si/Al ratio …) will be tested. New kind of porous materials constituted by Metal organic Frameworks (MOF) will also be looked at because they can constitute a promising way of trapping.

Experimental Investigation Into Buckling Failure of Slender Metal Plates in Severe Accident Conditions

2014 ◽  
Author(s):  
Byeongnam Jo ◽  
Wataru Sagawa ◽  
Koji Okamoto
Keyword(s):  
High Speed ◽  
Nuclear Power ◽  
Power Plants ◽  
Failure Load ◽  
Severe Accident ◽  
Steel Columns ◽  
Wide Range ◽  
Buckling Failure ◽  
The Difference ◽  
Accident Conditions

Buckling failure load of stainless steel columns under compressive stress was experimentally measured in severe accident conditions, which addresses the accidents in Fukushima Daiichi nuclear power plants. Firstly, buckling failure load defined as load which causes failure of the column (plastic collapse) was measured in a wide range of temperatures from 25 °C up to 1200 °C. The load values measured in this study were compared to numerical estimations by eigenvalue simulations (for an ideal column) and by nonlinear simulations (for a column with initial bending). Two different methods for measurement of the buckling failure load were employed to examine the effect of thermal history on buckling failure. Different load values were obtained from two methods in high temperature conditions over 800 °C. The difference in the buckling failure load between two methods increased with temperature, which was explained by the effect of creep at high temperatures. Moreover, the influence of asymmetric temperature profiles along a plate column was also explored with regard to the failure mode and the buckling failure load. In present study, all of the buckling processes were visualized by a high speed camera.

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