Feasibility Study on the In-Service Testing Program Plan for Safety-Related Dampers of Advanced Power Reactor-1400

2021 ◽  
Vol 7 (3) ◽  
Author(s):  
Shincheul Kang ◽  
Chan Park ◽  
Kyungho Lee ◽  
Sungho Lee
Keyword(s):  
Air Conditioning ◽  
Nuclear Power ◽  
Power Plants ◽  
Power Reactor ◽  
Selection Process ◽  
Hvac Systems ◽  
Air Treatment ◽  
Service Testing ◽  
Operation And Maintenance ◽  
Program Plan

Abstract In-service testing (IST) program including test and examination has been emphasized to assess the operability of the safety-related components in nuclear power plants. Since the issuance of ASME N511 in 2007 concerning IST for air treatment, heating, ventilating, and air conditioning (HVAC) systems, the IST program plan for those systems should be needed to establish prior to setting up the detailed program. The proper plan can contribute to enhance the effectiveness of allotment of the resources inevitable for implementation of the program and ultimately save operation and maintenance costs. In order to achieve the purposes, this paper presents the methodologies to establish IST program plan for safety-related dampers including their selection process of the dampers to be included in the program and representative template made by applying the methodologies. In addition, the method to identify testing types per damper for part of selected dampers is also provided in this paper.

Centralization of Canada’s Spent Nuclear Fuel

2013 ◽  
Author(s):  
Krista Nicholson ◽  
John McDonald ◽  
Shona Draper ◽  
Brian M. Ikeda ◽  
Igor Pioro
Keyword(s):  
Nuclear Fuel ◽  
Site Selection ◽  
Nuclear Power ◽  
Power Plants ◽  
Spent Nuclear Fuel ◽  
Selection Process ◽  
Spent Fuel ◽  
Geologic Repository ◽  
Candu Reactors ◽  
High Level

Currently in Canada, spent fuel produced from Nuclear Power Plants (NPPs) is in the interim storage all across the country. It is Canada’s long-term strategy to have a national geologic repository for the disposal of spent nuclear fuel for CANada Deuterium Uranium (CANDU) reactors. The initial problem is to identify a means to centralize Canada’s spent nuclear fuel. The objective of this paper is to present a solution for the transportation issues that surround centralizing the waste. This paper reviews three major components of managing and the transporting of high-level nuclear waste: 1) site selection, 2) containment and 3) the proposed transportation method. The site has been selected based upon several factors including proximity to railways and highways. These factors play an important role in the site-selection process since the location must be accessible and ideally to be far from communities. For the containment of the spent fuel during transportation, a copper-shell container with a steel structural infrastructure was selected based on good thermal, structural, and corrosion resistance properties has been designed. Rail has been selected as the method of transporting the container due to both the potential to accommodate several containers at once and the extensive railway system in Canada.

Qualification of Pumps and Valves for the Safety Injection Path of Nuclear Power Plants

2014 ◽  
Author(s):  
Ingo Ganzmann ◽  
Holger Schmidt
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Critical Factor ◽  
International Organization ◽  
Insulation Material ◽  
Hydraulic Test ◽  
Iso 17025 ◽  
Operation And Maintenance ◽  
Related Plant ◽  
Construction Operation

The reliability of a nuclear power plant depends on the safe functioning of its components during its lifetime: from design through construction, operation and maintenance. This is valid for new build projects as well as for the current fleet. As plants undergo modifications for increased performance or extended lifetimes, component integrity becomes a critical factor in those efforts, particularly for safety-related plant functions. This paper focuses on the qualification of pumps and valves of the safety-injection path, considering new requirements. Going back to the Barsebäck event in the year 1992, it is known that insulation material may cause clogging. Consequently, the presence of debris material in the water may have an impact on the functioning of pumps and valves. For this purpose, AREVA has built new thermo-hydraulic test loops in its accredited test and inspection body (according to International Organization for Standardization (ISO) 17025 and 17020) to consider this effect as it relates to components qualification (Ref. 1). The main relevant aspects of these tests will be discussed together with corresponding thermal shock tests. Paper published with permission.

The Application of an Integrated Head Assembly for Advanced Power Reactor 1400

2016 ◽  
Author(s):  
Tae Kyo Kang ◽  
Won Ho Jo ◽  
Yeon Ho Cho ◽  
Sang Gyoon Chang ◽  
Dae Hee Lee
Keyword(s):  
United Arab Emirates ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Power Reactor ◽  
Reactor Vessel ◽  
Head Region ◽  
The Usa ◽  
Head Assembly ◽  
Storage Area

The reactor vessel head region consists of a number of components and systems including reactor vessel head, CEDMs with their cables, cooling air system with ducts and fans, missile shield, seismic supports, head lift rig and cable supports. Prior to refueling operation, those components must be dismantled separately, and moved to the designated storage area. It was a very complicated and time consuming process. As a result, the integrated head assembly (IHA) was introduced to simplify those disassembling procedures, reduce refueling outage period, and improve safety in the containment building as those components are combined into a single system. To reduce refueling outage duration and radiation exposures to the workers by integrating the complicated reactor head region structures, KEPCO E&C has developed the IHA concept in the Korean Next Generation Reactor (KNGR) project [1]. The first application was implemented for the Optimized Power Reactor 1000 (OPR1000) at Shin-Kori units 1&2 and Shin-Wolsong units 1&2. With the past experience, the IHA was upgraded to be applied to the Advanced Power Reactor 1400 (APR1400). The design was patented in Korea [2], China, EU and the USA as modular reactor head area assembly. The IHA was applied for APR1400 nuclear power plants at Shin-Kori and Shin-Hanul, Korea. The design was also supplied to Barakah Nuclear Power Plants in the United Arab Emirates. This paper presents the design features and a variety of analysis which have been used for the APR1400 IHA.

Development of RI-ISI Program for Krško NPP

2013 ◽  
Author(s):  
Scott Kulat ◽  
Robin Graybeal ◽  
Benjamin Montgomery ◽  
Ivan Vrbanić ◽  
Ivica Bašić ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Selection Process ◽  
Degradation Mechanism ◽  
Electrical Power ◽  
Implementation Program ◽  
Class 1 ◽  
Process Risk ◽  
Almost All ◽  
The U.S

Risk-informed methodologies for inservice inspections of safety related piping in nuclear power plants were formally established in mid-1990s in the U.S. Since then, they have been adopted and applied by almost all of the U.S. plants. Nowadays, risk-informed inservice inspection (RI-ISI) is considered to be a standard for the operating plants in the U.S. It was not long before the RI-ISI practice started to be “exported” from the U.S. to other countries. By now, RI-ISI had found its way to a number of European and other countries. Among the recent examples is the Krško Nuclear Power Plant (Krško NPP), a two-loop Westinghouse-designed PWR located in Slovenia and owned by Slovenian and Croatian utilities. Krško NPP finished its third inservice inspection (ISI) interval in July 2012 and initiated implementation of the RI-ISI program at the start of the fourth interval. The process used to develop the RI-ISI program conformed to the methodology described in Electrical Power Research Institute (EPRI) Topical Report TR-112657 and included a degradation mechanism evaluation, consequence analysis and risk characterization for ASME Class 1 and Class 2 piping, as well as an element/examination selection process, risk impact assessment and inspection implementation program development. This paper describes the development of the Krško NPP RI-ISI program and the results of its RI-ISI application. A discussion is, also, provided on some aspects relevant for application of RI-ISI approaches developed in the U.S to plants outside of the U.S.

ASME Operation and Maintenance (OM) Code Overview

2016 ◽  
Author(s):  
Ronald C. Lippy
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Reactor Power ◽  
Regulatory Requirements ◽  
General Overview ◽  
Operation And Maintenance ◽  
American Society ◽  
General Organization

The purpose of this paper is to provide a general overview of the organization and content of the American Society of Mechanical Engineers (ASME) Operation and Maintenance of Nuclear Power Plants (OM) Code. This will involve a brief description of the regulatory requirements associated with Inservice Testing (IST) as well as a brief overview of the OM Code scope and requirements. This paper will discuss, in general, the regulations requiring IST as well as a brief discussion on when Preservice Testing (PST) and IST become required. A general organization of the ASME OM Code will be provided as well as general topics associated with how to determine when testing and examination intervals are established; what documentation is required; and general discussion regarding the various subsections of the OM Code and the components associated with the OM Code. Alternatives to the OM Code requirements and how to obtain these alternatives will also be provided as well as how the edition applicability of the ASME OM Code is determined. There is also discussion regarding a few general issues associated with the OM Code regarding existing reactor power plants as well as the “new builds” and advanced reactor plants and designs.

The Role of Human Information Processing Models in System Development

1989 ◽  
Vol 33 (16) ◽  
pp. 1059-1063 ◽  
Author(s):  
Barry H. Kantowitz
Keyword(s):  
Information Processing ◽  
Nuclear Power ◽  
Power Plants ◽  
System Development ◽  
Theoretical Models ◽  
Cognitive Engineering ◽  
Human Cognition ◽  
Human Information Processing ◽  
Operation And Maintenance

Humans are complicated devices. Thus, systems in which people are embedded necessarily are complex. In order to better develop such systems, a means to organize and understand human complexity is required. Theoretical models of human information processing are one cognitive-engineering tool to help system development. This paper discusses the kinds of models that might be effective in solving practical problems. Suggestions are given for selecting a useful model from the plethora of available theoretical models. These issues are illustrated in the context of current research aimed at providing a general model of human cognition and action for application to the development, operation, and maintenance of nuclear power plants in Japan.

scholarly journals ASME in-service inspection and in-service testing programs for Angra 1

2021 ◽  
Vol 8 (3A) ◽  
Author(s):  
ANA ROSA BALIZA MAIA ◽  
Youssef Morghi ◽  
AMIR ZACARIAS MESQUITA
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Functional Requirements ◽  
Term Operation ◽  
Plant Safety ◽  
Service Testing ◽  
Design And Manufacturing ◽  
License Renewal ◽  
Service Inspection

The in-service inspection program of the Angra 1 plant is updated every 10 years, according to applicable standards - designer (American Project - are followed NRC requirements) and Cnen. NRC approves the use of ASME Section XI (In-service Inspection of Nuclear Power Plant Components). The object of in-service inspection of components in nuclear power plants is to provide a continuing assurance that they are safe. To provide this assurance for those components that are subject to the requirements of the ASME Boiler and Pressure Vessel Code, a set of rules has been formulated to provide assurance that the functional requirements of the components are available when required. The rules have been arranged to provide appropriate levels of assurance according to the importance of the component in its relationship to plant safety. The classifications that are established during design and manufacturing have been adopted to provide the levels of importance for the components. The types of components typically found in the various classifications have then been identified and rules formulated for each type. For each type of component in each classification, the functions have been considered and methods of inspecting, testing, or monitoring each component is specified. These rules include methods of determining the limits of acceptance of the results. Should it be necessary to take corrective action to repair various components, rules have been provided to establish acceptable methods of repair or replacement. Angra 1 started the Renewal License and Long-term Operation project and there are three important Aging Management Programs (AMP) that are based on ASME section XI. This article will discuss the ASME section XI subsections that are important for the License Renewal and Long-term Operation for Angra 1.  

Managing Check Valve Flange Leaks: Technical Solutions and Challenges to Overcome

2012 ◽  
Author(s):  
L. Ike Ezekoye ◽  
William E. Densmore ◽  
William M. Turkowski ◽  
Robert E. Becse
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Check Valve ◽  
Technical Solution ◽  
Valve Body ◽  
Service Testing ◽  
Piping Systems ◽  
Maintenance Requirements ◽  
Technical Solutions ◽  
Plant Shutdown

Check valves are the simplest valves in power plants. Their simplicity and passive nature, combined with their relatively low maintenance requirements, often mask their relative importance in piping systems. Compared to power operated valves (POVs), such as motor operated valves or air operated valves, check valves have very few parts. The more parts a valve has, the more likely failures will occur. As such, power operated valves tend to have more stringent requirements that cover periodic verification of operability, in-service testing (IST), and scheduled preventive maintenance to assure functionality. Check valves, on the other hand, do not require nearly the same amount of rigor to assure operability. The passive nature of check valves often leads the user not to expect failures. Consequently, lacking of attention often results to inadvertent failures. One failure that has received significant attention from both the industry and the regulator is check valve body-to-bonnet joint leakage. In nuclear power plants this leakage can contaminate the general area where the valve is located, can lead to a plant shutdown, and pose personnel hazards. In this paper, the technical solutions that can be used to manage check valve body-to-bonnet joint leakage will be presented. The merits of each technical solution and the associated challenges will be discussed. Also, as some of the leakage containment solutions are appurtenances to the valve, the paper will address the interface between the appurtenances and the valve.

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