Application of Artificial Intelligence for Automated Detection of Defects in Nuclear Energy Domain

2021 ◽  
Author(s):  
Eleftherios Anagnostopoulos ◽  
Yann Kernin
Keyword(s):  
Artificial Intelligence ◽  
Nuclear Power ◽  
Power Plants ◽  
Data Augmentation ◽  
Primary Circuit ◽  
Destructive Testing ◽  
Blind Test ◽  
Pressurized Water ◽  
Data Set ◽  
Training Time

Abstract Ensuring the integrity of the primary circuit in nuclear power plants is crucial considering the extreme pressures and temperatures while operating Pressurized Water Reactors (PWR). Non-Destructive Testing (NDT) on such harsh environments is a challenging and complex scenario. Automated assistance on acquisition and analysis systems can importantly contribute as supplementary safety barrier by providing real-time alarms for potential existence of defects. In this paper we present the application of Artificial Intelligence in Visual Testing (VT) of Bottom Mounted Nozzles (BMN) of the Reactor Pressure Vessel (RPV). The method that we apply is based on Object Detection using Convolutional Neural Networks (CNN) combined with the Transfer Learning technique in order to limit the necessary training time of the model and the use of Data Augmentation methods for reducing the size of the learning data set. The proposed CNN demonstrates great performances for automatic surface defect detection (cracks) in highly noisy environments with variating illumination conditions. These performances combined with accurate localization and characterization of the defects confirms the interest of advanced CNNs against traditional imaging processing methods for NDT applications. In this study, the results of a comparative blind-test between Human VT analysts are also presented.

Lessons Learned in Reviewing CCF Events

2009 ◽  
Author(s):  
Ashley Mossa ◽  
Rupert Weston
Keyword(s):  
Information Exchange ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Lessons Learned ◽  
Level Of Detail ◽  
Data Sets ◽  
Pressurized Water ◽  
Data Set ◽  
Multiple Data Sets

The U.S. Nuclear Regulatory Commission (NRC) has an ongoing Common Cause Failure (CCF) data analysis program that periodically collects and evaluates information on component failures at U.S. commercial Nuclear Power Plants (NPPs). The primary information sources include the Licensee Event Reports (LER) and records from the Equipment Performance Information Exchange (EPIX) program. Once the information is collected, the failure records are evaluated to identify potential CCF events. CCF events are then coded, reviewed, and loaded into the NRC’s database. Verification of the CCF events is performed with the intended purpose of ensuring that events entered into the CCF database are indeed CCF events and that the event coding is consistent and correct. To ensure technical accuracy and correctness of the events loaded into the CCF database, the NRC requested the Pressurized Water Reactors Owners Group (PWROG) support in reviewing these events. Reviews of multiple data sets of CCF events were conducted on behalf of the PWROG. The data sets included CCF events that have occurred at U.S. commercial nuclear power plants. CCF events that occurred during 2006 through 2007 were included in the most recent data set that was reviewed. The level of information provided for reported CCF events varies from utility-to-utility. Without utility participation or input, the lack of consistency and varying level of detail can lead to incorrect interpretation and classification of a CCF event regarding its Probabilistic Risk Assessment (PRA) impact. This paper offers lessons learned from the reviews that were conducted. Insights for improving the consistency and level of detail related to the PRA information are summarized in this paper. The leading causes of initial misclassification of CCF events and patterns observed in conducting the reviews are discussed. The resolutions of misclassified CCF events are also discussed as part of the evaluation process to enhance the pedigree of the CCF database.

A Study of Ductility Dip Cracking of Inconel 690 Welding Filler Metal: Development of a Refusion Cracking Test

2017 ◽  
Author(s):  
Abel Rapetti ◽  
Patrick Todeschini ◽  
Sofiane Hendili ◽  
Frédéric Christien ◽  
Franck Tancret
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Tensile Tests ◽  
Primary Circuit ◽  
Fast Heating ◽  
Pressurized Water ◽  
Hot Crack ◽  
Inconel 690 ◽  
Crack Susceptibility ◽  
Filler Metals

Inconel® alloy 690 is nowadays commonly used instead of 600 for the manufacturing of certain components of the primary circuit of pressurized water reactor (PWR) nuclear power plants, due to its superior resistance to corrosion and stress corrosion cracking. However 690 alloy, and the corresponding welding filler metals (types 52 and 152), can be sensitive to a solid state hot cracking phenomenon during welding, called “ductility dip cracking” (DDC) associated to grain boundary cracking. This work is undertaken to determine more precisely the thermomechanical conditions of the occurrence of DDC in two types of materials: filler metals 52M and 152. To do this, we designed a simple hot crack susceptibility test. This test is based on multiple welding beads on a cuboidal mockup. This test clearly demonstrates the effect of multiple passes on the occurrence of DDC. In parallel, hot tensile tests following fast heating were performed to determine the DDC temperature range, to try and correlate DDC to the thermomechanical behavior.

Pressure compensation systems in the primary circuit of a nuclear power plant with a pressurized water reactor

2020 ◽  
Author(s):  
S. Gorbatov ◽  
V. Polunichev
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Design Feature ◽  
Primary Circuit ◽  
Pressurized Water Reactor ◽  
Technical Documentation ◽  
Pressurized Water ◽  
Compensation Systems ◽  
Water Reactors ◽  
Pressure Compensation

A pressure compensator is a technical pressure vessel with a special design that compensates for changes in the volume of water in a closed loop when it is heated. It is a design feature of two-circuit reactors with pressurized water as a coolant (including heavy water reactors) used at nuclear power plants, nuclear submarines and ships and is usually considered as part of a technological system that maintains the pressure in the primary circuit in stationary modes and pressure deviations in transient and emergency modes of the reactor plant. The pressure compensator is at the same time a system for providing the required pressure and compensating for changes in the volume of the coolant in the primary circuit, therefore it has a double name - in technical documentation and literature it can be called both a pressure compensator and a volume compensator.

Codes Comparison and Analysis of Weld Cladding Ultrasonic Testing in Nuclear Power Plants

2013 ◽  
Author(s):  
Zhimin Zhong ◽  
Jian Min ◽  
Kai Li
Keyword(s):  
Ultrasonic Testing ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Pressurized Water Reactor ◽  
Destructive Testing ◽  
Pressurized Water ◽  
The Difference ◽  
Weld Cladding ◽  
Non Destructive

This paper briefly introduces the weld cladding structure, its common defects during the manufacture and operation stage and its application in pressurized water reactor nuclear power plants. Some ultrasonic testing codes or standards for nuclear power plant pressure vessel or piping, such as ASME BPVC volume V & III & XI, Germany KTA 3201.3 and 3201.4 code, France RCC-M and RSE-M code, and Russia code of light water nuclear power plants were discussed. The difference of those codes and some feed backs have been analyzed and discussed. Furthermore, these works really benefit the compiling of NB/T 20003.2-2010, Non-destructive Testing for Mechanical Components in Nuclear Island of Nuclear Power Plants-Part 2: Ultrasonic Testing, as China building more and more nuclear power plants. It was concluded that we shall pay more attention to the inspection of cladding, not only at manufacture stage but in operation outage stage. One of important work is periodically updating the inspection standard revision. It was believed that improving the cladding defects inspection reliability and effectiveness is very important to the safety of nuclear power plants operation in China and in the world.

The Failure Analysis and Processing of Digital Reactor Protection System

2013 ◽  
Author(s):  
Jun Zhao ◽  
Xing Zhou ◽  
Jin Hu ◽  
Yanling Yu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Phase 1 ◽  
Protection System ◽  
Pressurized Water Reactor ◽  
Communication Failure ◽  
Pressurized Water ◽  
Power Failure ◽  
Self Test ◽  
Reactor Protection System

The Qinshan Nuclear Power Plant phase 1 unit (QNPP-1) has a power rating of 320 MWe generated by a pressurized water reactor that was designed and constructed by China National Nuclear Corporation (CNNC). The TELEPERM XS I&C system (TXS) is to be implemented to transform analog reactor protection system (RPS) in QNPP-1. The paper mainly describes the function, structure and characteristic of RPS in QNPP-1. It focuses on the outstanding features of digital I&C, such as strong online self-test capability, the degradation of the voting logic processing, interface improvements and CPU security. There are some typical failures during the operation of reactor protection system in QNPP-1. The way to analyze and process the failures is different from analog I&C. The paper summarizes typical failures of the digital RPS in the following types: CPU failure, communication failure, power failure, Input and output (IO) failure. It discusses the cause, risk and mainly processing points of typical failure, especially CPU and communication failures of the digital RPS. It is helpful for the maintenance of the system. The paper covers measures to improve the reliability of related components which has been put forward effective in Digital reactor protection system in QNPP-1. It will be valuable in nuclear community to improve the reliability of important components of nuclear power plants.

Development of Long Rotating Blade Applied in the New Generation Nuclear Steam Turbine

2012 ◽  
Author(s):  
Kai Cheng ◽  
Zeying Peng ◽  
Gongyi Wang ◽  
Xiaoming Wu ◽  
Deqi Yu
Keyword(s):  
Steam Turbine ◽  
Nuclear Power ◽  
Power Plants ◽  
Structure Design ◽  
Good Reliability ◽  
Pressurized Water ◽  
Height Range ◽  
Water Reactor ◽  
Rotating Blade ◽  
Last Stage

In order to meet the high economic requirement of the 3rd generation Pressurized Water Reactor (PWR) or Boiling Water Reactor (BWR) applied in currently developing nuclear power plants, a series of half-speed extra-long last stage rotating blades with 26 ∼ 30 m2 nominal exhaust annular area is proposed, which covers a blade-height range from 1600 mm to 1900 mm. It is well known that developing an extra long blade is a tough job involving some special coordinated sub-process. This paper is dedicated to describe the progress of creating a long rotating blade for a large scaled steam turbine involved in the 3rd generation nuclear power project. At first the strategy of how to determine the appropriate height for the last-stage-rotating-blade for the steam turbine is provided. Then the quasi-3D flow field quick design method for the last three stages in LP casing is discussed as well as the airfoil optimization method. Furthermore a sophisticated blade structure design and analyzing system for a long blade is introduced to obtain the detail dimension of the blade focusing on the good reliability during the service period. Thus, except for CAD and experiment process, the whole pre-design phase of the extra-long turbine blade is presented which is regarded as an assurance of the operation efficiency and reliability.

Digital I&C Operating Experience in the US

2008 ◽  
Author(s):  
Bruce Geddes ◽  
Ray Torok
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Failure Modes ◽  
Operating Experience ◽  
Data Set ◽  
Common Cause ◽  
Hardware Failure ◽  
Common Cause Failure ◽  
Defensive Measures

The Electric Power Research Institute (EPRI) is conducting research in cooperation with the Nuclear Energy Institute (NEI) regarding Operating Experience of digital Instrumentation and Control (I&C) systems in US nuclear power plants. The primary objective of this work is to extract insights from US nuclear power plant Operating Experience (OE) reports that can be applied to improve Diversity and Defense in Depth (D3) evaluations and methods for protecting nuclear plants against I&C related Common Cause Failures (CCF) that could disable safety functions and thereby degrade plant safety. Between 1987 and 2007, over 500 OE events involving digital equipment in US nuclear power plants were reported through various channels. OE reports for 324 of these events were found in databases maintained by the Nuclear Regulatory Commission (NRC) and the Institute of Nuclear Power Operations (INPO). A database was prepared for capturing the characteristics of each of the 324 events in terms of when, where, how, and why the event occurred, what steps were taken to correct the deficiency that caused the event, and what defensive measures could have been employed to prevent recurrence of these events. The database also captures the plant system type, its safety classification, and whether or not the event involved a common cause failure. This work has revealed the following results and insights: - 82 of the 324 “digital” events did not actually involve a digital failure. Of these 82 non-digital events, 34 might have been prevented by making full use of digital system fault tolerance features. - 242 of the 324 events did involve failures in digital systems. The leading contributors to the 242 digital failures were hardware failure modes. Software change appears as a corrective action twice as often as it appears as an event root cause. This suggests that software features are being added to avoid recurrence of hardware failures, and that adequately designed software is a strong defensive measure against hardware failure modes, preventing them from propagating into system failures and ultimately plant events. 54 of the 242 digital failures involved a Common Cause Failure (CCF). - 13 of the 54 CCF events affected safety (1E) systems, and only 2 of those were due to Inadequate Software Design. This finding suggests that software related CCFs on 1E systems are no more prevalent than other CCF mechanisms for which adherence to various regulations and standards is considered to provide adequate protection against CCF. This research provides an extensive data set that is being used to investigate many different questions related to failure modes, causes, corrective actions, and other event attributes that can be compared and contrasted to reveal useful insights. Specific considerations in this study included comparison of 1E vs. non-1E systems, active vs. potential CCFs, and possible defensive measures to prevent these events. This paper documents the dominant attributes of the evaluated events and the associated insights that can be used to improve methods for protecting against digital I&C related CCFs, applying a test of reasonable assurance.

scholarly journals Calculation of Potential Radiation Doses Associated with Predisposal Management of Dismantled Steam Generators from Nuclear Power Plants

2020 ◽  
Vol 12 (12) ◽  
pp. 5149
Author(s):  
Ga Hyun Chun ◽  
Jin-ho Park ◽  
Jae Hak Cheong
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Radiation Doses ◽  
Steam Generators ◽  
Pressurized Water ◽  
Water Reactor ◽  
General Pathway ◽  
Heavy Water Reactor ◽  
Past Experiences

Although the generation of large components from nuclear power plants is expected to gradually increase in the future, comprehensive studies on the radiological risks of the predisposal management of large components have been rarely reported in open literature. With a view to generalizing the assessment framework for the radiological risks of the processing and transport of a representative large component—a steam generator—12 scenarios were modeled in this study based on past experiences and practices. In addition, the general pathway dose factors normalized to the unit activity concentration of radionuclides for processing and transportation were derived. Using the general pathway dose factors, as derived using the approach established in this study, a specific assessment was conducted for steam generators from a pressurized water reactor (PWR) or a pressurized heavy water reactor (PHWR) in Korea. In order to demonstrate the applicability of the developed approach, radiation doses reported from actual experiences and studies are compared to the calculated values in this study. The applicability of special arrangement transportation of steam generators assumed in this study is evaluated in accordance with international guidance. The generalized approach to assessing the radiation doses can be used to support optimizing the predisposal management of large components in terms of radiological risk.

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