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.

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.

Nuclear Power Plant Fires and Explosions: Part IV — Water Hammer Explosion Mechanisms

2017 ◽  
Author(s):  
Robert A. Leishear
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Reactor ◽  
Water Hammer ◽  
Compression Process ◽  
Flammable Gas ◽  
Piping Systems ◽  
Fukushima Daiichi ◽  
Accident Conditions

Water hammers, or fluid transients, compress flammable gasses to their autognition temperatures in piping systems to cause fires or explosions. While this statement may be true for many industrial systems, the focus of this research are reactor coolant water systems (RCW) in nuclear power plants, which generate flammable gasses during normal operations and during accident conditions, such as loss of coolant accidents (LOCA’s) or reactor meltdowns. When combustion occurs, the gas will either burn (deflagrate) or explode, depending on the system geometry and the quantity of the flammable gas and oxygen. If there is sufficient oxygen inside the pipe during the compression process, an explosion can ignite immediately. If there is insufficient oxygen to initiate combustion inside the pipe, the flammable gas can only ignite if released to air, an oxygen rich environment. This presentation considers the fundamentals of gas compression and causes of ignition in nuclear reactor systems. In addition to these ignition mechanisms, specific applications are briefly considered. Those applications include a hydrogen fire following the Three Mile Island meltdown, hydrogen explosions following Fukushima Daiichi explosions, and on-going fires and explosions in U.S nuclear power plants. Novel conclusions are presented here as follows. 1. A hydrogen fire was ignited by water hammer at Three Mile Island. 2. Hydrogen explosions were ignited by water hammer at Fukushima Daiichi. 3. Piping damages in U.S. commercial nuclear reactor systems have occurred since reactors were first built. These damages were not caused by water hammer alone, but were caused by water hammer compression of flammable hydrogen and resultant deflagration or detonation inside of the piping.

Sensitivity Analyses for a PWR SCC Case Using the Probabilistic Loss-of-Coolant-Accident (Pro-LOCA) Software

2016 ◽  
Author(s):  
Bruce A. Young ◽  
Sang-Min Lee ◽  
Paul M. Scott
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
The United States ◽  
Design Criterion ◽  
Mitigation Strategies ◽  
Sensitivity Analyses ◽  
Base Case ◽  
Piping System ◽  
Loss Of Coolant Accident ◽  
Piping Systems

As a means of demonstrating compliance with the United States Code of Federal Regulations 10CFR50 Appendix A, General Design Criterion 4 (GDC-4) requirement that primary piping systems for nuclear power plants exhibit an extremely low probability of rupture, probabilistic fracture mechanics (PFM) software has become increasingly popular. One of these PFM codes for nuclear piping is Pro-LOCA which has been under development over the last decade. Currently, Pro-LOCA is being enhanced under an international cooperative program entitled PARTRIDGE-II (Probabilistic Analysis as a Regulatory Tool for Risk-Informed Decision GuidancE - Phase II). This paper focuses on the use of a pre-defined set of base-case inputs along with prescribed variation in some of those inputs to determine a comparative set of sensitivity analyses results. The benchmarking case was a circumferential Primary Water Stress Corrosion Crack (PWSCC) in a typical PWR primary piping system. The effects of normal operating loads, temperature, leak detection, inspection frequency and quality, and mitigation strategies on the rupture probability were studied. The results of this study will be compared to the results of other PFM codes using the same base-case and variations in inputs. This study was conducted using Pro-LOCA version 4.1.9.

Effects on Pipe Vibrations of Cavitation in an Orifice and in Globe-Style Valves

2006 ◽  
Author(s):  
Se´bastien Caillaud ◽  
Rene´-Jean Gibert ◽  
Pierre Moussou ◽  
Joe¨l Cohen ◽  
Fabien Millet
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Low Frequency ◽  
Flow Regimes ◽  
Piping System ◽  
Low Frequencies ◽  
Choked Flow ◽  
Single Hole ◽  
Piping Systems ◽  
Large Amplitude Vibrations

A piping system of French nuclear power plants displays large amplitude vibrations in particular flow regimes. These troubles are attributed to cavitation generated by single-hole orifices in depressurized flow regimes. Real scale experiments on high pressure test rigs and on-site tests are then conducted to explain the observed phenomenon and to find a solution to reduce pipe vibrations. The first objective of the present paper is to analyze cavitation-induced vibrations in the single-hole orifice. It is then shown that the orifice operates in choked flow with supercavitation, which is characterized by a large unstable vapor pocket. One way to reduce pipe vibrations consists in suppressing the orifices and in modifying the control valves. Three technologies involving a standard trim and anti-cavitation trims are tested. The second objective of the paper is to analyze cavitation-induced vibrations in globe-style valves. Cavitating valves operate in choked flow as the orifice. Nevertheless, no vapor pocket appears inside the pipe and no unstable phenomenon is observed. The comparison with an anti-cavitation solution shows that cavitation reduction has no impact on low frequency excitation. The effect of cavitation reduction on pipe vibrations, which involve essentially low frequencies, is then limited and the first solution, which is the standard globe-style valve installed on-site, leads to acceptable pipe vibrations. Finally, this case study may have consequences on the design of piping systems. First, cavitation in orifices must be limited. Choked flow in orifices may lead to supercavitation, which is here a damaging and unstable phenomenon. The second conclusion is that the reduction of cavitation in globe-style valve in choked flow does not reduce pipe vibrations. The issue is then to limit cavitation erosion of valve trims.

Long Term Performance of PE4710 Materials in Disinfectant Treated Nuclear Raw Water Systems

2013 ◽  
Author(s):  
Alexander M. Summe ◽  
Douglas P. Munson ◽  
Kenneth Oliphant ◽  
Sarah Chung
Keyword(s):  
Power Plants ◽  
Slow Crack Growth ◽  
Life Time ◽  
Water Systems ◽  
Superior Performance ◽  
Mechanical Degradation ◽  
Service Water ◽  
Long Term Performance ◽  
Term Performance

Degradation of service water systems is a major issue facing nuclear power plants and many plants will require repair or replacement of existing carbon steel piping components. High-density polyethylene (HDPE) has been used in non-safety service water systems for over ten years and has demonstrated superior performance. However, there still exist knowledge gaps around material properties, inspectability, and long-term performance. Specifically, there is a lack of insight on the aging of HDPE piping in disinfectant treated service water systems. This paper summarizes the methodology and results of predicting the expected life time of HDPE piping exposed to oxidizing biocides in numerous end-use scenarios. The aging mechanism of concern is Stage III Chemical-Mechanical degradation, where the polymer is oxidized by biocides and then experiences slow crack growth (SCG). An Aging Model is used to provide general predictions of pipe service life. The results were analyzed for trends and limiting or sensitive operating parameters were identified. For most applications, the specific resin used in the model demonstrated good performance for lifetimes of well over 40 years.

Development of Code Case N-755: Use of Polyethylene (PE) Plastic Pipe for Section III, Division 1, Construction and Section XI Repair/Replacement Activities

2008 ◽  
Author(s):  
Frank J. Schaaf
Keyword(s):  
Nuclear Power ◽  
Nuclear Regulatory Commission ◽  
Regulatory Review ◽  
Plastic Pipe ◽  
Service Water ◽  
Division 1 ◽  
Minimum Requirements ◽  
American Society ◽  
Regulatory Commission ◽  
Safety Applications

With the increasing failures of metallic pipe in nuclear Service Water Systems, a new material needed to be found. One option is polyethylene (PE) pipe. PE pipe can be used in non-safety applications at a nuclear plant using the American Society of Mechanical Engineers (ASME) B31, Standards of Pressure Piping with no regulatory review. However, the use of PE material in safety applications, which are regulated by the Nuclear Regulatory Commission (NRC), necessitates a new Standard with special requirements. At the request of the Duke Power Corporation, a new ASME Standard was written by a special Project Team. This standard is found in the form of a Code Case under the control of the ASME Boiler & Pressure Vessel Code (B&PVC). The Code Case utilizes Sections of the B&PVC as its foundation and includes the design, procurement, installation, fusing, examination and testing requirements for the use of PE pipe within safety systems. The first version of the Code Case contained only the minimum requirements needed to support Duke Power Corporation’s first phase of PE piping installation into a safety system within a nuclear power plant. The Code Case developed is titled, N-755, Use of Polyethylene (PE) Plastic Pipe for Section III, Division 1, Construction and Section XI Repair/Replacement Activities. The first version of this case is limited to buried piping using only the following components; straight PE pipe, PE mitered elbows, and transition flanges. The Code Case will be revised as data for material and components becomes available at the completion of testing.

Seismic Fragility of Piping Nozzles in Nuclear Power Plants: A Case for Updating the Current State-of-Practice

2021 ◽  
Author(s):  
Abhinav Gupta ◽  
Ankit Dubey ◽  
Sunggook Cho
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Seismic Analysis ◽  
Nuclear Industry ◽  
Life Extension ◽  
Seismic Loads ◽  
Thermal Loads ◽  
Loss Of Coolant Accident ◽  
Piping Systems ◽  
Recent Developments

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 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 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.

Computational Study of Conjugate Heat Transfer in T-Junctions

2009 ◽  
Author(s):  
Simon Kuhn ◽  
Bojan Nicˇeno ◽  
Horst-Michael Prasser
Keyword(s):  
Thermal Fatigue ◽  
Nuclear Power ◽  
Power Plants ◽  
Cold Water ◽  
Computational Study ◽  
Solid Wall ◽  
Turbulent Viscosity ◽  
Life Time ◽  
Scale Model ◽  
Piping Systems

Thermal fatigue is a relevant problem in the context of life-time extension of nuclear power plants (NPP). In many piping systems in NPPs hot and cold water is mixed, which leads to high temperature fluctuations in the region close to the solid wall and resulting thermal loads on the pipe walls that can cause fatigue. One of the relevant geometric test cases for thermal fatigue is the mixing in T-junctions. In this study we apply large–eddy simulations (LES) to the mixing of hot and cold water in a T-junction. We perform a set of simulations by using different formulations of the LES subgrid scale model, i.e. standard Smagorinsky and dynamic procedure, to identify the influence of the modelled subgrid scales on the simulation results. The results exhibit a large difference between the models, which is caused by the use of turbulent viscosity wall–damping functions when applying the standard model.

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