Relevant bibliographies by topics / Nuclear industry piping

Contents

  1. Journal articles
  2. Dissertations / Theses
  3. Books
  4. Book chapters
  5. Conference papers

Academic literature on the topic 'Nuclear industry piping'

Author: Grafiati
Published: 4 June 2021
Last updated: 4 February 2022

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Journal articles on the topic "Nuclear industry piping"

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Bernsen, Sidney, Bryan Erler, Dana K. Morton, and Owen Hedden. "The Code Builders." Mechanical Engineering 136, no. 05 (May 1, 2014): 36–41. http://dx.doi.org/10.1115/1.2014-may-2.

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This article elaborates the evolution of code and standards for nuclear power plants. In the 1950s, need was felt for a revised set of design and fabrication rules to facilitate the development of safe, economically competitive water-cooled reactors contained in pressure vessels. These rules were codified in the first edition of the ASME Boiler and Pressure Vessel Code Section III, which was completed in 1963 and published in 1964. From the outset, both regulators and industry realized that the best way to develop many of the needed rules for the design, construction, and operation of nuclear facilities was the national standards consensus process. This process, followed by the American National Standards Institute and other recognized standards-issuing bodies such as ASME, brings together the expertise of individuals from government, industry, academia, and other stakeholders. In the years following the first publication of Section III, the coverage of the Code expanded to incorporate piping requirements, pressure-retaining components for pumps and valves, equipment and piping supports, reactor vessel internal structures, and other features of nuclear power plants.
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Ciechanowski, Ata, T. Walsh, and S. W. Dean. "NSF 14: Shaping the Future of the Plastic Piping Industry." Journal of ASTM International 8, no. 9 (2011): 103879. http://dx.doi.org/10.1520/jai103879.

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Lee, Sang Min, Young Hwan Choi, Hae Dong Chung, Yoon Suk Chang, and Young Jin Kim. "Plastic Limit Analysis of an Elbow with Various Wall-Thinning Geometries." Key Engineering Materials 385-387 (July 2008): 833–36. http://dx.doi.org/10.4028/www.scientific.net/kem.385-387.833.

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A piping system including straight pipes, elbows and tee branches in a nuclear power plant is mostly subjected to severe loading conditions with high temperature and pressure. In particular, the wall-thinning of an elbow due to flow accelerated corrosion is one of safety issues in the nuclear industry. In this respect, it is necessary to investigate the limit loads of an elbow with a wall-thinned part for evaluating integrity. In this paper, three dimensional plastic limit analyses are performed to obtain limit loads of an elbow with different bend angles as well as defect geometries under internal pressure and in-plane/out-of-plane bending moment. The limit loads are also compared with the results from limit load solutions of an uninjured elbow based on the von Mises yield criteria. Finally, the effects of significant factors, bend angle and defect shape, are quantified to estimate the exact load carrying capacity of an elbow during operation.
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Chen, X. L., Z. Yang, A. Nanjundan, N. Chen, and J. J. Janosch. "Achieving manufacturing quality and reliability using thru-process simulation." Journal de Physique IV 120 (December 2004): 793–800. http://dx.doi.org/10.1051/jp4:2004120092.

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Manufacturing of heavy industrial components such as the front linkage, case frame and car body of construction and mining equipments involves multiple processes including thermal cutting, bending/forming and welding. Distortion, welding joint integrity and manufacturing variability pose significant challenges on today's manufacturing. In this paper, the authors presented VFT$^{\tiny TM}$- Virtual Fabrication Technology [1], a thru-process simulation package, developed at the technical center of Caterpillar Inc., Peoria, IL, USA and Battelle Memorial Institute, Columbus Ohio, USA. VFT is a trademark of Battelle Memorial Institute. This work was performed under the support of the US Department of Commerce, National Institute of Standards, Advanced Technology Program. Steel rolling simulation, thermal cutting simulation, bending and welding simulation capabilities in the VFT software have significantly reduced the time and cost of manufacturing process developments as well as to achieve competitive advantages through quality improvement and differentiated products. Besides the Caterpillar applications, VFT software has also found applications in shipbuilding, nuclear piping, and automotive industry.
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Au-Yang, M. K. "Acoustic and Ultrasonic Signals as Diagnostic Tools for Check Valves." Journal of Pressure Vessel Technology 115, no. 2 (May 1, 1993): 135–41. http://dx.doi.org/10.1115/1.2929507.

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A typical nuclear plant has between 60 and 115 safety-related check valves ranging from 2 to 30 in. The majority of these valves control water flow. Recent studies done by the Institute of Nuclear Power Operations (INPO), Electric Power Research Institute (EPRI) and the US Nuclear Regulatory Commission (NRC) found that many of these safety-related valves were not functioning properly. Typical problems found in these valves included disk flutter, backstop tapping, flow leakage, disk pin and hinge pin wear, or even missing disks. These findings led to INPO’s Significant Operating Experience Report (SOER, 1986), and finally, NRC generic letter 89-04, which requires that all safety-related check valves in a nuclear plant be regularly monitored. In response to this need, the industry has developed various diagnostic equipment to monitor and test check valves, using technologies ranging from acoustics and ultrasonics to magnetic—even radiography has been considered. Of these, systems that depend on a combination of acoustic and ultrasonic techniques (Au-Yang et al., 1991) are among the most promising for two reasons: these two technologies supplement each other, making diagnosis of the check valves much more certain than any single technology, and this approach can be made nonintrusive. The nonintrusive feature allows the check valves to be monitored and diagnosed without being disassembled or removed from the piping system. This paper shows that by carefully studying the acoustic and ultrasonic signatures acquired from a check valve, either individually or in combination, an individual with the proper training and experience in acoustic and ultrasonic signature analyses can deduce the structural integrity of the check valve with good confidence. Most of the conclusions are derived from controlled experiments in the laboratory where the diagnosis can be verified. Other conclusions were based on test data obtained in the field.
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Katsareas, Dimitrios Elias, and Anastasius Youtsos. "Residual Stress Prediction in Dissimilar Metal Weld Pipe Joints Using the Finite Element Method." Materials Science Forum 490-491 (July 2005): 53–61. http://dx.doi.org/10.4028/www.scientific.net/msf.490-491.53.

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Dissimilar metal welds are commonly found in the primary piping of pressurized water nuclear reactor power plants. The safety assessment practice for such welds requires residual stresses to be taken into consideration. In the present paper the finite element method is utilized for the simulation of the welding process and prediction of the residual stress field in a dissimilar metal weld pipe joint. Although it is common practice to develop in-house finite element codes for weld simulation, the ANSYS commercial finite element code is selected. This is mainly due to the fact that industry focuses on commercial software, since residual stress analysis procedures based on them can be readily transferred to industrial applications. A simplified 2-D axi-symmetric model, in which residual stresses are produced due to the thermo-mechanical properties mismatch during cooling of the weld, is compared with a detailed model in which the complete multi-pass welding procedure is simulated. The latter incorporates the “birth & death of elements” technique, temperature dependant material properties and kinematic hardening material behavior. The aim of this comparison is to establish the degree of model detail and complexity, necessary to obtain satisfactory results and consequently to define a golden rule between computational cost and practically accurate predictions. Identifying the specific simulation parameters and variables, that have the highest impact on the accuracy of the computed results, is also important. It is concluded that, a bead-by-bead or lump-by-lump detailed simulation is necessary in order to obtain reasonably accurate residual stresses that cannot be predicted by a simplified model. A general conclusion is that the proposed method, being simple in implementation and cost effective concerning model complexity and analysis time, is a potential weld residual stress prediction tool.
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Sukamta, S., Aldi Rahadian Ilham, and S. Sudarja. "THE INVESTIGATION OF VOID FRACTION OF TWO-PHASE FLOW AIR-WATER AND GLYCERINE (0-30%) IN THE CAPILLARY PIPE WITH SLOPE OF 50 TO HORIZONTAL POSITION." Media Mesin: Majalah Teknik Mesin 20, no. 1 (April 12, 2019): 8–17. http://dx.doi.org/10.23917/mesin.v20i1.7385.

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Two-phase flow is a flow that is composed of two combined substances, i.e., liquid and gas. Two-phase flow phenomena can be a lot encountered in nature, for example, exhaust gas smoke, fog, rain, cloud, snow, and so forth. Moreover, in the practice of company or industry environment, the flow of the two phases can be encountered on the equipment in the form of heat exchangers, boilers, nuclear reactors, piping systems, geothermal, natural gas liquefaction, and others. Data collection in this research was carried out in 4 stages (from working fluid), namely: air-water, air-water+0% glycerin, air-water+10% glycerin, air-water+20% glycerin, and air-water+30% glycerin. The research aims to determine the void fraction by using the method of digital image processing with the software MATLAB R2014a. Based on the research results, it was obtained that when superficial gas velocity high then, the value of the void fraction obtained will increase, otherwise the higher speed of the superficial liquid then, the value of void fraction that is retrieved will decrease. The viscosity of the fluid very influenced the flow pattern bubbly and plug. It is because the higher the viscosity of fluid flow pattern speed then, the bubbly and the plug will increasingly decline, instead of getting down viscosity makes bubbly flow pattern speed, and the plug will progressively increase. The length of the bubbly and plug pattern is affected by the increasing value of homogeneous (β). As a consequence, the length of the pattern increases. In the frequency when the bubbly and plug occurred, it was obtained a high enough rate that led to the value of the void fraction generated considerable increases.Keyword: two-phase, fluid, void fraction, viscosity, flow pattern.
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Gupta, Abhinav, Ankit R. Dubey, and Sung-Gook Cho. "Seismic Fragility of Piping Nozzles in Nuclear Power Plants: A Case for Updating the Current State-of-Practice." Journal of Pressure Vessel Technology 143, no. 6 (July 9, 2021). http://dx.doi.org/10.1115/1.4051136.

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Abstract Nuclear industry spends enormous time and resources on designing and managing piping nozzles in a plant. Nozzle locations are considered as a potential location for possible failure that can lead to loss of coolant accident. Industry spends enormous time in condition monitoring and margin management at nozzle locations. Margins against seismic loads play a significant role in the overall margin management. Available margins against thermal loads are highly dependent upon seismic margins. In recent years, significant international collaboration has been undertaken through Nuclear Energy Agency (NEA) to study the seismic margin in piping systems and nozzles through experimental and analytical studies. It has been observed that piping nozzles are highly overdesigned and the margins against seismic loads are quite high. While this brings a perspective of sufficient safety, such excessively high margins compete with available margins against thermal loads particularly during the life extension and subsequent license renewal (SLR) studies being conducted by many plants around the world. This paper focuses on identifying and illustrating two key reasons that lead to excessively conservative estimates of nozzle fragilities. First, it compares fragilities based on conventional seismic analysis that ignores piping–equipment–structure interaction on nozzle fragility with the corresponding assessment by considering such interactions. Then, it presents a case that the uncertainties considered in various parameters for calculating nozzle fragility are excessively high. The paper identifies a need to study the various uncertainties in order to achieve a more realistic quantification based on recent developments in our understanding of the seismic behavior of piping systems.
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Leishear, Robert A. "A Hydrogen Ignition Mechanism for Explosions in Nuclear Facility Piping Systems." Journal of Pressure Vessel Technology 135, no. 5 (September 18, 2013). http://dx.doi.org/10.1115/1.4024639.

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Hydrogen explosions may occur simultaneously with fluid transients' accidents in nuclear facilities, and a theoretical mechanism to relate fluid transients to hydrogen deflagrations and explosions is presented herein. Hydrogen and oxygen generation due to the radiolysis of water is a recognized hazard in piping systems used in the nuclear industry, where the accumulation of hydrogen and oxygen at high points in the piping system is expected, and explosive conditions may occur. Pipe ruptures in nuclear reactor cooling systems were attributed to hydrogen explosions inside pipelines, i.e., Hamaoka, Nuclear Power Station in Japan, and Brunsbuettel in Germany (Fig. 1Fig. 1Hydrogen explosion damage in nuclear facilities Antaki, et al. [9,10–12] (ASME, Task Group on Impulsively Loaded Vessels, 2009, Bob Nickell)). Prior to these accidents, an ignition source for hydrogen was not clearly demonstrated, but these accidents demonstrated that a mechanism was, in fact, available to initiate combustion and explosion. A new theory to identify an ignition source and explosion cause is presented here, and further research is recommended to fully understand this explosion mechanism. In fact, this explosion mechanism may be pertinent to explosions in major nuclear accidents, and a similar explosion mechanism is also possible in oil pipelines during off-shore drilling.
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Krishnaswamy, P., D. J. Shim, and S. Kalyanam. "Comparison of Parent and Butt-Fusion Material Properties of Unimodal High-Density Polyethylene." Journal of Pressure Vessel Technology 139, no. 4 (June 1, 2017). http://dx.doi.org/10.1115/1.4036658.

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The U.S. nuclear power industry is seeking U.S. Nuclear Regulatory Commission (USNRC) approval to use high-density polyethylene (HDPE) in safety-related applications. The USNRC had granted approval for the use of HDPE for safety-related service water applications, with limitations, to Catawba (Duke Energy Corp., Catawba, SC) and Callaway (Union Electric Co., Callaway, MO) based on separate relief requests submitted by the licensees. The nuclear industry continues to show increasing interest in utilizing HDPE in safety-related piping systems. In order to evaluate and maintain the structural integrity of HDPE pipes, the material properties and the fracture resistance behavior must be fully characterized. Although there has been extensive work on material property development of HDPE, most of the investigations have been focused on the parent (base) material. Hence, the material property and fracture resistance behavior of the butt-fusion region have not been comprehensively investigated. In this paper, tensile, dynamic mechanical analysis (DMA), and slow crack growth (SCG) tests were performed for unimodal PE 4710 HDPE material. Specimens were machined from both parent piping material and butt-fusion regions. The test results indicate that the tensile and DMA properties show no significant differences between parent and butt-fusion joint materials. However, in terms of SCG resistance, the time to failure for butt-fusion joint material was an order of magnitude lower than that of the parent material.
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Dissertations / Theses on the topic "Nuclear industry piping"

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Zhao, Yong. "Random vibration for seismic analysis of multiply supported nuclear piping." Case Western Reserve University School of Graduate Studies / OhioLINK, 1994. http://rave.ohiolink.edu/etdc/view?acc_num=case1061299266.

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Books on the topic "Nuclear industry piping"

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Regulation, U. S. Nuclear Regulatory Commission Office of Nuclear Reactor. Standard review plan for trial use for the review of risk-informed inservice inspection of piping. [Washington, DC]: U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, 1998.

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Shah, Vikram N. Review of industry efforts to manage pressurized water reactor feedwater nozzle, piping, and feedring cracking and wall thinning. Washington, DC: U.S. Nuclear Regulatory Commission, 1997.

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Workshop on Debris Impact on Emergency Coolant Recirculation (2004 Albuquerque, N.M.). Debris impact on emergency coolant recirculation: Workshop proceedings, Albuquerque, NM, United States, 25-27 February 2004 : a joint report. [Paris]: Nuclear Energy Agency, Organisation for Economic Co-operation and Development, 2004.

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D, Todd M., U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Engineering., and Oak Ridge National Laboratory, eds. A characterization of check valve degradation and failure experience in the nuclear power industry. Washington, DC: Division of Engineering, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1993.

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C, Watkins John, Bramwell D, U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Engineering Technology., and Idaho National Engineering Laboratory, eds. Motor-operated valve (MOV) actuator motor and gearbox testing. Washington, DC: Division of Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1997.

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C, Watkins John, Bramwell D, U.S. Nuclear Regulatory Commission. Office of Nuclear Regulatory Research. Division of Engineering Technology., and Idaho National Engineering Laboratory, eds. Motor-operated valve (MOV) actuator motor and gearbox testing. Washington, DC: Division of Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1997.

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Motor-operated valve (MOV) actuator motor and gearbox testing. Washington, DC: Division of Engineering Technology, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 1997.

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Agency, OECD Nuclear Energy. Debris Impact On Emergency Coolant Recirculation (Nuclear Safety). OECD, 2004.

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Book chapters on the topic "Nuclear industry piping"

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Wellein, R. "Applications of PFM in the nuclear industry to reactor pressure vessel, main coolant piping and steel containment." In Probabilistic fracture mechanics and reliability, 325–50. Dordrecht: Springer Netherlands, 1987. http://dx.doi.org/10.1007/978-94-017-2764-8_7.

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Conference papers on the topic "Nuclear industry piping"

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Ezekoye, L. Ike. "Failure Analysis Prediction Versus Industry Failure Experience of Air Operated Valves in the Nuclear Industry." In ASME 2016 Pressure Vessels and Piping Conference. American Society of Mechanical Engineers, 2016. http://dx.doi.org/10.1115/pvp2016-63852.

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Safety related valves in the nuclear industry are designed to meet the requirements of the design specifications for the systems in which they are to be installed. In developing valve specifications, systems and valve engineers collaborate to craft the essential requirements needed to support the procurement of the valves that meet the design requirements, and thereby provide reliable service during plant life. The specification requirements, together with the ASME Boiler and Pressure Vessel Code and Standards, provide a strong basis for assuring both structural integrity and functionality of the valve assemblies. The functional requirements cover the duties of the valves. As these valves are safety related, they are generally subjected to preoperational testing and possibly additional qualification testing during manufacture, to ensure that the valves can perform their safety related functions in service. The nuclear experience of engineered products such as valves shows that considerable amount of analysis and documentation of component stresses are performed to ensure compliance with the ASME code and specification requirements. The ASME Code requirements, together with the normal controls applied during manufacture of safety related valves, enhance the reliability of the valves. However, valve failures still occur during plant operation. In this paper, the failures of air operated valves (AOVs) used in nuclear applications were reviewed and the data compared against the failures predicted by valve suppliers based on weak link analysis of the valves. The study shows that there are significant differences between what the suppliers consider structurally likely to fail, what the purchaser expects to fail, and what really fails from field experience. The study shows that field failures are complex. They can be initiated by many factors, most of which are not obvious and cannot be controlled by the valve designer. The complexity of field failures of air operated valves is discussed in this paper.
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Angell, Peter, Sheng-Hui Wang, Phil Simon, Hank Kleinfelder, Kevin Garrity, and Elizabeth Sisk. "Addressing the Challenge of Inspecting Buried Nuclear Piping in Nuclear Power Plants." In ASME 2012 Pressure Vessels and Piping Conference. American Society of Mechanical Engineers, 2012. http://dx.doi.org/10.1115/pvp2012-78794.

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Similar to many other industries, nuclear power plants (NPPs) have many kilometres of buried pipe that is not readily accessible for direct inspection. Given the nature of the systems, the nuclear industry experiences additional challenges as many services run in the same area, leading to what is described as a “spaghetti bowl” of piping. As a result, the traditional indirect, over the line, inspection practices developed for the oil and gas industry have not always been successfully applied at nuclear power plants. To address this issue, a collaborative Electric Power Research Institute/CANDU Owners Group (EPRI/COG) research project was established with Mears Group Inc. and Atomic Energy of Canada Limited (AECL) Nuclear Laboratories. In this program, initial testing of four methods was carried out at the Mears Group Inc, test site that had been modified with additional grounding systems to simulate a nuclear power plant. This was followed by testing of the same methods at the AECL Nuclear Laboratories Chalk River site. This paper will discuss the results of those studies and present some of the findings that were made that can help to overcome the challenges faced by Nuclear Power Plants.
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McCracken, Steven L., and Richard E. Smith. "Alternative Approach for Qualification of Temperbead Welding in the Nuclear Industry." In ASME 2012 Pressure Vessels and Piping Conference. American Society of Mechanical Engineers, 2012. http://dx.doi.org/10.1115/pvp2012-78571.

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Temperbead welding is common practice in the nuclear power industry for in-situ repair of quenched and tempered low alloy steels where post weld heat treatment is impractical. The temperbead process controls the heat input such that the weld heat-affected-zone (HAZ) in the low alloy steel is tempered by the welding heat of subsequent layers. This tempering eliminates the need for post weld heat treatment (PWHT). Unfortunately, repair organizations in the nuclear power industry are experiencing difficulty when attempting to qualify temperbead welding procedures on new quenched and tempered low alloy steel base materials manufactured to modern melting and deoxidation practices. The current ASME Code methodology and protocol for verification of adequate fracture toughness in materials was developed in the early 1970s. This paper reviews typical temperbead qualification results for vintage heats of quenched and tempered low alloy steels and compares them to similar test results obtained with modern materials of the same specification exhibiting superior fracture toughness.
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Ezekoye, L. Ike, Mark Rain, and Theodore E. Thygesen. "Check Valves in Nuclear Power Plants: ASME OM Code Requirements and Industry Compliance." In ASME 2020 Pressure Vessels & Piping Conference. American Society of Mechanical Engineers, 2020. http://dx.doi.org/10.1115/pvp2020-21315.

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Abstract Nuclear power plants use lots of valves in different applications. Each valve type has specific tasks to perform. The vast majority of safety related valves are monitored during plant operation to ensure they are operable and can be relied upon to do their design functions. Unfortunately, check valves do not fall in the category of valves that can be readily monitored because of how they are designed; usually with all operating components internal to the pressure boundary. Since they are not easily monitored externally, plant operators typically disassemble and inspect them to assess the internal conditions and determine what actions to take, which can be either repair, replace, or use as found. However, not all check valves get inspected for a number of reasons resulting in some valves running to failure. In this paper, the authors present actions the nuclear industry has used to comply with ASME OM Code requirements (8) on check valves and US NRC regulatory activities in this area. The paper also covers the role the nuclear industry and other organizations have played to improve check valve reliability.
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Murakami, Hiroyoshi. "First Publication of the Nuclear Materials Code for Nuclear Facilities in Japan." In ASME/JSME 2004 Pressure Vessels and Piping Conference. ASMEDC, 2004. http://dx.doi.org/10.1115/pvp2004-2693.

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The Subgroup on Nuclear Materials of JSME (Japan Society of Mechanical Engineers) has been developed a new nuclear materials code for nuclear facilities. The current nuclear materials code was published as a part of the Ministry of Economy, Trade and Industry Code (METI) Notification No. 501 [1]. The new materials code shall come out as a separate publication. It shall publish in 2004. The new nuclear materials code contains 58 ferrous material specifications of Japanese Industry Standards (JIS), 15 nonferrous material specifications of JIS and 14 nuclear specified materials. The code has widely consistent with JIS such as JIS B8265 [2] and B8266 [3], JSME code such as the Rules on Design and Construction for Thermal Power Generation Facilities [4], and ASME Boiler and Pressure Vessel Section II [5]. This paper introduces outlines on the Rules on Nuclear Materials for Nuclear Facilities.
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Server, William, and Bruce Craig. "Some Comparisons and Potential Structural Integrity Methodology Transfer Between the Petroleum Industry and the Nuclear Power Industry." In ASME 2006 Pressure Vessels and Piping/ICPVT-11 Conference. ASMEDC, 2006. http://dx.doi.org/10.1115/pvp2006-icpvt-11-93679.

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Many different large component industries develop structural integrity methodologies independently of each other even though the guiding principles of fracture mechanics and flaw tolerance concepts are followed. This paper presents a comparison of some key aspects of flaw tolerance approaches used in the petroleum and nuclear power industries. Differences in the types of components and structures are acknowledged as well as material degradation and failure mechanisms. Potential transfer of structural integrity technology for some aspects of the methodologies will be presented and discussed. The hope is that some cross-fertilization between the industries can be developed and maintained.
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Gosselin, S. R., F. A. Simonen, P. G. Heasler, S. R. Doctor, and F. L. Becker. "NDE Performance (POD) Curves for Fatigue Cracks in Piping Based on Industry Performance Demonstration Data." In 10th International Conference on Nuclear Engineering. ASMEDC, 2002. http://dx.doi.org/10.1115/icone10-22418.

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This paper evaluates non-destructive examination (NDE) detection capabilities of fatigue cracks in piping. Industry performance demonstration initiative (PDI) data for fatigue crack detection were used to develop a matrix of statistically based probability of detection (POD) curves that consider various NDE performance factors. Seven primary performance factors were identified — Material, Crack Geometry/Type, NDE Examination Access, NDE Procedure, Examiner Qualification, Pipe Diameter, and Pipe Wall Thickness. A database of 16,181 NDE performance observations, with 18 fields associated with each observation, was created and used to develop statistically based POD curves for 42 stainless steel and 14 carbon steel performance cases. Subsequent comparisons of the POD fits for each of the cases showed that excellent NDE performance for fatigue cracks can be expected for ferritic materials. Very little difference was observed between the POD curves for the 14 carbon steel performance cases considered in this study and NDE performance could therefore be represented by a single POD curve. For stainless steel, very good performance can also be expected for circumferential cracks located on the same side of the weld from which the NDE examination is made. POD depended primarily on component thickness. Three POD curves for stainless steel were prepared. Best estimate and the associated 95% confidence bounds POD verses through-wall depth logistic regression digital data are provided. Probabilistic fracture mechanics (PFM) calculations were performed to compare best estimate leak probabilities obtained from both the new performance-based POD curves and previous PFM models. This work was performed under joint funding by EPRI and the U.S. Department of Energy (DOE), Office of Nuclear Energy Science and Technology’s Nuclear Energy Plant Optimization (NEPO) program.
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Sanwarwalla, Mansoor H. "Regulatory and Industry Guidance on Use of Risk Technology in Nuclear Power Plant Applications." In ASME 2007 Pressure Vessels and Piping Conference. ASMEDC, 2007. http://dx.doi.org/10.1115/pvp2007-26271.

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Since the United States Nuclear Regulatory Commission (USNRC) published its landmark “Reactor Safety Study — An Assessment of Accident Risks in U. S. Commercial Nuclear Power Plants” in late 1975, commercial nuclear power industry, encouraged by the USNRC, have since then been applying Probabilistic Risk Assessment (PRAs) in their nuclear power units in areas of in-service testing, in-service inspection, quality assurance, technical specifications, maintenance, etc. To guide and regulate the industry in use of PRAs, Regulatory Guides and Standards have been written and are being revised continuously by the USNRC, American Society of Mechanical Engineers (ASME) and American Nuclear Society (ANS). The current use of PRA takes credit for single failure criterion based on applicability of codes and standards. The proposed new USNRC regulation 10 CFR Part 53 applicable for all reactor technologies is silent on the applicability of current standards endorsed by the regulatory body. The impact of the proposed new rule to both new and the current application needs to be studied. This paper will review the application of the various guidance documents for their use in commercial nuclear power plants with emphasis on the new generation nuclear power plants.
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Alley, David. "US Nuclear Regulatory Commission Perspective on Buried and Underground Piping Issues at Nuclear Power Plants in the United States." In ASME 2014 Pressure Vessels and Piping Conference. American Society of Mechanical Engineers, 2014. http://dx.doi.org/10.1115/pvp2014-28249.

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This paper provides a historical perspective on the need for, and development of, buried and underground piping tanks programs at nuclear power plants. Nuclear power plant license renewal activities, Nuclear Regulatory Commission Buried Piping Action Plan, and the rationale for addressing the issue of buried pipe through an industry initiative as opposed to regulation are discussed. The paper also addresses current NRC activities including the results of Nuclear Regulatory Commission inspections of buried piping programs at nuclear power plants as well as Nuclear Regulatory Commission involvement in industry and standards development organizations. Finally, the paper outlines the Nuclear Regulatory Commission’s future plans concerning the issue of buried piping at US nuclear power plants.
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Matthews, Dale E., Ralph S. Hill, and Charles W. Bruny. "2025 Nuclear Code: The Vision for the Future of ASME Nuclear Codes and Standards." In ASME 2018 Pressure Vessels and Piping Conference. American Society of Mechanical Engineers, 2018. http://dx.doi.org/10.1115/pvp2018-84031.

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ASME Nuclear Codes and Standards are used worldwide in the construction, inspection, and repair of commercial nuclear power plants. As the industry looks to the future of nuclear power and some of the new plant designs under development, there will be some significant departures from the current light water reactor (LWR) technology. Some examples are gas-cooled and liquid metal-cooled high temperature reactors (HTRs), small modular reactors (SMRs), and fusion energy devices that are currently under development. Many of these designs will have different safety challenges from the current LWR fleet. Variations of the current LWR technology are also expected to remain in use for the foreseeable future. Worldwide, many LWRs are planned or are already under construction. However, technology for construction of these plants has advanced considerably since most of the current construction codes were written. As a result, many modern design and fabrication methods available today, which provide both safety and economic benefits, cannot be fully utilized since they are not addressed by Code rules. For ASME Nuclear Codes and Standards to maintain and enhance their position as the worldwide leader in the nuclear power industry, they will need to be modernized to address these items. Accordingly, the ASME Nuclear Codes and Standards organizations have initiated the “2025 Nuclear Code” initiative. The purpose of this initiative is to modernize all aspects of ASME’s Nuclear Codes and Standards to adopt new technologies in plant design, construction, and life cycle management. Examples include modernized finite element analysis and fatigue rules, and incorporation of probabilistic and risk-informed methodology. This paper will present the vision for the 2025 ASME Nuclear Codes and Standards and will discuss some of the key elements that are being considered.
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