Slow Crack Growth Resistance of Parent and Joint Materials From PE4710 Piping for Safety-Related Nuclear Power Plant Piping

2011 ◽  
Author(s):  
S. Kalyanam ◽  
D.-J. Shim ◽  
P. Krishnaswamy ◽  
Y. Hioe
Keyword(s):  
Crack Growth ◽  
Nuclear Power ◽  
Power Plants ◽  
Slow Crack Growth ◽  
Nuclear Industry ◽  
Pipe Material ◽  
Service Water ◽  
Hdpe Pipe ◽  
Pipe Materials ◽  
Hdpe Pipes

HDPE pipes are considered by the nuclear industry as a potential replacement option to currently employed metallic piping for service-water applications. The pipes operate under high temperatures and pressures. Hence HDPE pipes are being evaluated from perspective of design, operation, and service life requirements before routine installation in nuclear power plants. Various articles of the ASME Code Case N-755 consider the different aspects related to material performance, design, fabrication, and examination of HDPE materials. Amongst them, the material resistance (part of Article 2000) to the slow crack growth (SCG) from flaws/cracks present in HDPE pipe materials is an important concern. Experimental investigations have revealed that there is a marked difference (almost three orders less) in the time to failure when the notch/flaw is in the butt-fusion joint, as opposed to when the notch/flaw is located in the parent HDPE material. As part of ongoing studies, the material resistance to SCG was investigated earlier for unimodal materials. The current study investigated the SCG in parent and butt-fusion joint materials of bimodal HDPE (PE4710) pipe materials acquired from two different manufacturers. The various stages of the specimen deformation and failure during the creep test are characterized. Detailed photographs of the specimen side-surface were used to monitor the specimen damage accumulation and SCG. The SCG was tested using a large specimen (large creep frame) as well as using a smaller size specimen (PENT frame) and the results were compared. Further, the effect of polymer orientation or microstructure in the bimodal HDPE pipe on the SCG was studied using specimens with axial and circumferential notch orientations in the parent pipe material.

Dependence of Slow Crack Growth Resistance on Pipe Sizes for Scratched Polyethylene Nuclear Water Pipe

2009 ◽  
Author(s):  
Z. Jimmy Zhou ◽  
Dane Chang
Keyword(s):  
Stress Intensity ◽  
Crack Growth ◽  
Nuclear Power ◽  
Power Plants ◽  
Water Distribution ◽  
Slow Crack Growth ◽  
Water Pipe ◽  
Scratch Depth ◽  
Geometric Factors ◽  
Service Water

Polyethylene (PE) material is chemically inert and does not corrode. It is a preferred material choice for many piping applications including the service water pipe in nuclear power plants. With an outstanding field service record for 5 decades for both gas and water distribution, recently a 36″ SDR 9 PE4710 pipe received approval for Release Request from the U.S. Nuclear Regulatory Committee (NRC) and was installed in 2008 by a major US utility to replace the steel pipe for the safety related nuclear water pipe application. The PE pipe industry uses a maximum 10% scratch as a rule of thumb for scratched pipes. As the pipe diameter increases, the stress intensity increases for the same 10% of scratch depth. It is a concern for the regulators how the increased stress intensity affects the long term performance of the scratched pipe against slow crack growth (SCG) failure. The SCG resistance depends on the resin material, stress intensity and testing temperature. The stress intensity is controlled by pipe geometric factors, the applied stress, and the notch depth. In order to achieve the equivalent SCG performance for all pipe sizes, the following actions may be taken: (1) use a resin material that has higher SCG resistance to compensate impact from the increased stress intensity; (2) reduce the scratch depth and (3) reduce the pipe hoop stress so that the increased stress intensity is decreased. The minimum SCG resistance requirement that eliminates the need for action (2) and (3) needs to be determined. ASTM F1473 has addressed the equivalent pipe sizes from 1″ SDR 11 to 8″ SDR 11 pipe. This paper investigates the dependence of SCG resistance on the geometric factors for the pipe diameters from 2″ to 44″. The recent PE4710 nuclear water pipe installation is also discussed.

Dynamic Testing of High Density Polyethylene Vent and Drain Configurations

2012 ◽  
Author(s):  
Douglas Munson ◽  
Timothy M. Adams ◽  
Shawn Nickholds
Keyword(s):  
Carbon Steel ◽  
High Density Polyethylene ◽  
Nuclear Power ◽  
Power Plants ◽  
Dynamic Testing ◽  
High Density ◽  
Steel Pipe ◽  
Labor Costs ◽  
Service Water ◽  
Hdpe Pipe

For corroded piping in low temperature systems, such as service water systems in nuclear power plants, replacement of carbon steel pipe with high density polyethylene (HDPE) pipe is a cost-effective solution. HDPE pipe can be installed at much lower labor costs than carbon steel pipe, and HDPE pipe has a much greater resistance to corrosion. This paper presents the results of the seismic testing of selected vent and drain configurations. This testing was conducted to provide proof of the conceptual design of HDPE vent and drain valve configurations. A total of eight representative models of HDPE vent and drain assemblies were designed. The models were subjected to seismic SQURTS spectral acceleration up to maximum shake table limits. The test configurations were then checked for leakage and operability of the valves. The results for these tests, along with the test configurations, are presented. Also presented are the acceleration data observed at various points on the test specimens.

Modern Bimodal High Density Polyethylene for the Nuclear Power Plant Piping System

2009 ◽  
Author(s):  
Adel N. Haddad
Keyword(s):  
High Density Polyethylene ◽  
Nuclear Power ◽  
Large Diameter ◽  
Water System ◽  
High Density ◽  
Nuclear Industry ◽  
Piping System ◽  
Service Water ◽  
Hdpe Pipe ◽  
Related Service

Originally introduced in the 1990s, bimodal HDPE, pipe resins are still finding new niches today, including even nuclear power plants. HDPE pipe grades are used to make strong, corrosion resistant and durable pipes. High density polyethylene, PE 4710, is the material of choice of the nuclear industry for the Safety Related Service Water System. This grade of polymer is characterized by a Hydrostatic Design Basis (HDB) of 1600 psi at 73 °F and 1000 psi at 140 °F. Additionally bimodal high density PE 4710 grades display >2000 hours slow crack growth resistance, or PENT. HD PE 4710 grades are easy to extrude into large diameter pipes; fabricate into fitting and mitered elbows and install in industrial settings. The scope of this paper is to describe the bimodal technology which produces HDPE pipe grade polymer; the USA practices of post reactor melt blending of natural resin compound with black masterbatch; and the attributes of such compound and its conformance to the nuclear industry’s Safety Related Service Water System.

Determination of Creep Properties of PE 4710 Cell Classification 445574C High Density Polyethylene

2014 ◽  
Author(s):  
Timothy M. Adams ◽  
Shawn Nickholds ◽  
Douglas Munson ◽  
Jeffery Andrasik
Keyword(s):  
Carbon Steel ◽  
Pressure Vessel ◽  
High Density Polyethylene ◽  
Nuclear Power ◽  
Power Plants ◽  
High Density ◽  
Labor Costs ◽  
Cell Classification ◽  
Service Water ◽  
Hdpe Pipe

For corroded piping in low temperature systems, such as service water systems in nuclear power plants, replacement of carbon steel piping with high density polyethylene (HDPE) is a cost-effective solution. Polyethylene pipe can be installed at much lower labor costs than carbon steel pipe and HDPE pipe has a much greater resistance to corrosion. The ASME Boiler and Pressure Vessel Code, Section III, Division 1 currently permits the use of non-metallic piping in buried safety Class 3 piping systems. Additionally, HDPE pipe has been successfully used in non-safety-related systems in nuclear power facilities and is commonly used in other industries such as water mains and natural gas pipelines. This paper presents the results of creep testing of PE 4710 cell classification 445574C pipe compliant with ASME Boiler and Pressure Vessel Code material requirements. This information was developed to support and provide a strong technical basis for material properties of HDPE pipe for use in ASME Boiler and Pressure Vessel Code, Section III New Construction and Section XI repair or replacement activities. The data may also be useful for applications of HDPE pipe in commercial electric power generation facilities and chemical, process and waste water plants via its possible use in the B31 series piping codes. The report provides long term creep and modulus data, as well as an analysis of the stress dependency of both.

Use of Polyethylene (PE) Pipe in Safety-Related, Class 3, Service-Water Piping

2008 ◽  
Author(s):  
Prabhat Krishnaswamy ◽  
Eric M. Focht ◽  
Do-Jun Shim ◽  
Tao Zhang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Elevated Temperatures ◽  
Slow Crack Growth ◽  
Large Diameter ◽  
Modes Of Failure ◽  
Service Water ◽  
Division 1 ◽  
Piping Systems ◽  
Technical Issues

The ASME Boiler and Pressure Vessel Code Committee (BPVC) has recently published Code Case N-755 that describes the requirements for the use of Polyethylene (PE) pipe for the construction of Section III, Division 1 Class 3 buried piping systems for service water applications in nuclear power plants. The code case was developed by Special Working Group–Polyethylene Pipe (SWG-PP) within Section III (Design) of the BPVC. This paper provides a critical review of the design requirements described in CC N-755 from pressure boundary integrity considerations. The various technical issues that need to be addressed for safety-critical PE piping are discussed in this paper. Specifically, the premise of allowing defects in pipe that are 10% of the wall thickness has been reviewed especially for cases involving large diameter piping [> 304.8 mm (12 inches)] that is to be operated at elevated temperatures as high as 60°C (140°F). One of the common modes of failure in PE piping under sustained internal pressure is due to slow crack growth (SCG) from manufacturing or installation defects in the pipe wall. The effect of pipe diameter and stresses on the crack driving force for a 10% deep flaw is calculated for comparison with the material resistance to SCG at elevated temperatures.

Analysis of ASME Class 3 Buried HDPE Piping Systems Related to Code Case N-755-1

2013 ◽  
Author(s):  
Young Seok Kim ◽  
Jung Kwang Yoon ◽  
Young Ho Kim
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Design Procedure ◽  
Piping System ◽  
Service Water ◽  
Hdpe Pipe ◽  
Division 1 ◽  
Piping Systems ◽  
Asme Code ◽  
Essential Service

This paper proposes an analysis method for Section III, Division 1, Class 3 buried High Density polyethylene (HDPE) piping system in the nuclear power plants (NPP). Although HDPE pipe would yield at high temperature (limited to 140°F), it may be suitable for the areas prone to earthquakes; owing to its comparable ductility and flexibility. Thus, the buried HDPE piping may be applicable for the safety related Essential Service Water (ESW) system in the NPPs. Despite some limitations to buried HDPE piping, the piping could be designed based on ASME Code Case [1]. Generally, codes and standards including ASME Code Case [1] do not provide load combinations for the design of both buried steel piping and HDPE piping. Meanwhile, EPRI Report [4] provides load combinations including thermal expansion effects and seismic loads with detailed seismic criteria for polyethylene pipe. In this paper, load cases and load combinations for buried HDPE piping are suggested for implementation of reference documents and a buried HDPE piping system is analyzed referring to EPRI Report [4] to evaluate stress, force, and moment using a piping stress analysis program. Additionally, this paper will recommend the design procedure in accordance with ASME Code Case [1] using an example of buried HDPE piping analysis. An investigation of soil spring coefficients and the design considerations for hydrostatic tests are suggested for the enhanced analysis of buried HDPE piping.

Ultrasonic Inspection of Electrofusion Joints of Large Polyethylene Pipes in Nuclear Power Plants

2016 ◽  
Vol 138 (6) ◽  
Author(s):  
Jinyang Zheng ◽  
Dongsheng Hou ◽  
Weican Guo ◽  
Xiaoming Miao ◽  
Yaoda Zhou ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Ultrasonic Inspection ◽  
Cooling Water ◽  
Operating Pressure ◽  
Hdpe Pipe ◽  
Hdpe Pipes

High-density polyethylene (HDPE) pipe has many advantages such as good flexibility, corrosion resistance, and long service life. It has been introduced into nuclear power plants for transportation of cooling water in U.S. and Europe. Recently, four HDPE pipelines (PE4710) were used in essential cooling water system with operating pressure of 0.6 MPa and operating temperature of no more than 60 °C in a newly established AP1000 nuclear power plant in Zhejiang, China. The outside diameter and thickness are 30 in. and 3.3 in., respectively, which are much larger and thicker than traditional HDPE pipe for natural gas. This brought forward a challenge for nondestructive testing (NDT) and safety assessment of such pipes. In this paper, a solution for inspecting electrofusion (EF) joints of thick-walled HDPE pipes is presented, and the results of an on-site inspection of the nuclear power plant are revealed. To expand the thickness up-limit of previously developed ultrasonic-phased array instrument, an optimization method was proposed by calculating weighing effects of different testing parameters and introducing the concept of overall performance according to practical requirement, by comprehensively considering sensitivity, penetration, signal-to-noise ratio (SNR), resolution, and accuracy. Typical defects were found in field inspection. The result shows that the presented technique is capable of inspecting EF joints for connecting large-size HDPE pipes used in nuclear power plants.

scholarly journals An Approach to Analyze Diagnosis Errors in Advanced Main Control Room Operations Using the Cause-Based Decision Tree Method

Energies ◽  
2021 ◽  
Vol 14 (13) ◽  
pp. 3832
Author(s):  
Awwal Mohammed Arigi ◽  
Gayoung Park ◽  
Jonghyun Kim
Keyword(s):  
Decision Tree ◽  
Nuclear Power ◽  
Power Plants ◽  
Human Error ◽  
Cognitive Activity ◽  
Emergency Operation ◽  
Nuclear Industry ◽  
Working Environment ◽  
Control Rooms ◽  
Diagnosis Errors

Advancements in the nuclear industry have led to the development of fully digitized main control rooms (MCRs)—often termed advanced MCRs—for newly built nuclear power plants (NPPs). Diagnosis is a major part of the cognitive activity in NPP MCRs. Advanced MCRs are expected to improve the working environment and reduce human error, especially during the diagnosis of unexpected scenarios. However, with the introduction of new types of tasks and errors by digital MCRs, a new method to analyze the diagnosis errors in these new types of MCRs is required. Task analysis for operator diagnosis in an advanced MCR based on emergency operation was performed to determine the error modes. The cause-based decision tree (CBDT) method—originally developed for analog control rooms—was then revised to a modified CBDT (MCBDT) based on the error mode categorizations. This work examines the possible adoption of the MCBDT method for the evaluation of diagnosis errors in advanced MCRs. We have also provided examples of the application of the proposed method to some common human failure events in emergency operations. The results show that with some modifications of the CBDT method, the human reliability in advanced MCRs can be reasonably estimated.

Inservice Testing Program Improvements for New Reactors Small Modular Reactors (SMRs)

2014 ◽  
Author(s):  
Ronald C. Lippy
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Nuclear Industry ◽  
Small Modular Reactors ◽  
American Society ◽  
Construction Permit ◽  
Regulatory Commission

The nuclear industry is preparing for the licensing and construction of new nuclear power plants in the United States. Several new designs have been developed and approved, including the “traditional” reactor designs, the passive safe shutdown designs and the small modular reactors (SMRs). The American Society of Mechanical Engineers (ASME) provides specific Codes used to perform preservice inspection/testing and inservice inspection/testing for many of the components used in the new reactor designs. The U.S. Nuclear Regulatory Commission (NRC) reviews information provided by applicants related to inservice testing (IST) programs for Design Certifications and Combined Licenses (COLs) under Part 52, “Licenses, Certifications, and Approvals for Nuclear Power Plants,” in Title 10 of the Code of Federal Regulations (10 CFR Part 52) (Reference 1). The 2012 Edition of the ASME OM Code defines a post-2000 plant as a nuclear power plant that was issued (or will be issued) its construction permit, or combined license for construction and operation, by the applicable regulatory authority on or following January 1, 2000. The New Reactors OM Code (NROMC) Task Group (TG) of the ASME Code for Operation and Maintenance of Nuclear Power Plants (NROMC TG) is assigned the task of ensuring that the preservice testing (PST) and IST provisions in the ASME OM Code to address pumps, valves, and dynamic restraints (snubbers) in post-2000 nuclear power plants are adequate to provide reasonable assurance that the components will operate as needed when called upon. Currently, the NROMC TG is preparing proposed guidance for the treatment of active pumps, valves, and dynamic restraints with high safety significance in non-safety systems in passive post-2000 reactors including SMRs.

Architecting “Maintenance Program” for Piping and Pipe Support

2013 ◽  
Author(s):  
Koichi Tsumori ◽  
Yoshizumi Fukuhara ◽  
Hiroyuki Terunuma ◽  
Koji Yamamoto ◽  
Satoshi Momiyama
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Industry ◽  
Maintenance Management ◽  
3D Cad ◽  
Cad System ◽  
Nuclear Disaster ◽  
Safety Regulations ◽  
Maintenance Program

A new inspection standard that enhanced quality of operating /maintenance management of the nuclear power plant was introduced in 2009. After the Fukushima Daiichi nuclear disaster (Mar. 11th 2011), the situation surrounding the nuclear industry has dramatically changed, and the requirement for maintenance management of nuclear power plants is pushed for more stringent nuclear safety regulations. The new inspection standard requires enhancing equipment maintenance. It is necessary to enhance maintenance of not only equipment but also piping and pipe support. In this paper, we built the methodology for enhancing maintenance plan by rationalizing and visualizing of piping and pipe support based on the “Maintenance Program” in cooperating with 3D-CAD system.

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