ASME Section III Subsection NF Linear Current and Future Concepts

2003 ◽  
Author(s):  
Robert J. Masterson
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Steel Structures ◽  
Structural Elements ◽  
Structural Rules ◽  
Nuclear Plants ◽  
Basic Set ◽  
Asme Code ◽  
The Difference ◽  
Future Concepts

Over the past thirty years Subsection NF of the ASME Boiler and Pressure Vessel Code, Section III, has been transformed from a basic set of structural rules to evaluate and qualify component supports to a dynamic Code that recognized the difference between pressure retaining and structural elements. Initially, with the publication of the 1974 Edition of the ASME Code, Subsection NF identified essentially two different types of supports, Plate & Shell and Linear. Plate & Shell supports were evaluated using in effect the same rules in Subsections NB, NC and ND for pressure retaining elements. Linear supports, on the other hand, were an odd group. The ASME Code was a pressure Code providing rules for pressure retaining components such as vessels, piping, valves, pumps, tanks etc. In 1963, when the first Edition of ASME Section III for nuclear plants was published, there wasn’t any place for non-pressure retaining elements such as component supports. This continued for eleven years until supports made their debut in the 1974 Edition of Section III (actually the 1973 Winter Addenda to the 1971 Edition). At this time the AISC Manual for Steel Construction was the premier set of rules for the design and construction of steel structures. Since many supports were comprised of structural steel elements due to the nature of loading (thermal and seismic) at nuclear power plants, the AISC Code seemed to be the ideal set of rules to adopt for the construction of supports. Additional requirements for temperature and seismic considerations were incorporated in Subsection NF to account for supports on nuclear plants.

scholarly journals Application of Explosion-Proof RFID Technology in Nuclear Power Plants

2020 ◽  
Vol 20 (2) ◽  
pp. 127-132
Author(s):  
Namjin Cho ◽  
Dongsu Im ◽  
Jungdon Kwon ◽  
Teayeon Cho ◽  
Junglim Lee
Keyword(s):  
Radio Frequency ◽  
Radio Frequency Identification ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Ignition Source ◽  
Rfid Technology ◽  
Nuclear Plants ◽  
Frequency Identification

Nuclear power plants store and use flammable gases and liquids and consequently risk explosions. Therefore, nuclear plants employ explosion-proof equipment; however, this equipment is not always sufficiently maintained. This lack of maintenance can affect the safety-related equipment intended to shut down the reactor, because the explosion-proof equipment itself can act as an ignition source. Radio-frequency identification (RFID) technology should be explored as a tool to improve both the convenience and efficiency of maintenance. We analyzed and compared explosion-proof RFID technology that can be used in nuclear power plants.

The Need for Strategic Planning for Nuclear Power Plant I&C Upgrades

2002 ◽  
Author(s):  
J. Douglas Hill ◽  
Paul Moore
Keyword(s):  
Strategic Planning ◽  
Control Systems ◽  
Digital Technology ◽  
Nuclear Power ◽  
Power Plants ◽  
Short Term ◽  
Nuclear Plants ◽  
And Control ◽  
Plant Control

Nuclear power plants rely on Instrumentation and Control (I&C) systems for control, monitoring and protection of the plant. The original, analog designs used in most nuclear plants have become or soon will be obsolete, forcing plants to turn to digital technology. Many factors affect the design of replacement equipment, including long-term and short-term economics, regulatory issues, and the way the plant operates on a day-to-day basis. The first step to all modernization projects should involve strategic planning, to ensure that the overall long and short-term goals of the plant are met. Strategic planning starts with a thorough evaluation of the existing plant control systems, the available options, and the benefits and consequences of these options.

Methodology to Evaluate Nuclear Power Plant Vulnerability to Events in the Transmission Grid

2014 ◽  
Author(s):  
Alberto Del Rosso ◽  
Jean-François Roy ◽  
Frank Rahn ◽  
Alejandro Capara
Keyword(s):  
Risk Assessment ◽  
Nuclear Power ◽  
Power Plants ◽  
Assessment Tool ◽  
Transmission Grid ◽  
Simulation Tools ◽  
Nuclear Plants ◽  
Western Electricity Coordinating Council ◽  
The Impact ◽  
The U.S

This paper presents a general approach to evaluate the risk of trip or Loss of Off-site Power (LOOP) events in nuclear power plants due to contingencies in the power grid. The proposed methodology is based on the Zone of Vulnerability concept for nuclear plants introduced by EPRI in previous work. The proposed methodology is intended to be part of an integrated probabilistic risk assessment tool that is being developed under ongoing EPRI R&D programs. A detailed analysis of many events occurred in actual nuclear plants has been performed in order to identify, classify and characterize the various vulnerability and type of failures that may affect a nuclear plant. Based the outcome of that analysis, a methodology for evaluating the impact of off-site transmission system events on nuclear plants has been outlined. It includes description of the type of contingencies and conditions that need to be included in the analysis, as well as provisions regarding the simulation tools and models that should be used in each case. The methodology is illustrated in a simplified representation of the Western Electricity Coordinating Council (WECC) system in the U.S.

Proposed Improvements of the ASME B&PV Code and Future Directions of CIWG

2013 ◽  
Author(s):  
Jinquan Yan ◽  
Yinbiao He ◽  
Gang Li ◽  
Hao Yu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Third Generation ◽  
Future Directions ◽  
The Past ◽  
Development Direction ◽  
Asme Code ◽  
Set Up ◽  
Near Future ◽  
First Time

The ASME B&PV Code, Section III, is being used as the design acceptance criteria in the construction of China’s third generation AP1000 nuclear power plants. This is the first time that the ASME Code was fully accepted in Chinese nuclear power industry. In the past 6 years, a few improvements of the Code were found to be necessary to satisfy the various requirements originated from these new power plant (NPP) constructions. These improvements are originated from a) the stress-strain curves needed in elastic-plastic analysis, b) the environmental fatigue issue, c) the perplexity generated from the examination requirements after hydrostatic test and d) the safe end welding problems. In this paper, the necessities of these proposed improvements on the ASME B&PV code are further explained and discussed case by case. Hopefully, through these efforts, the near future development direction and assignment of the ASME B&PV-III China International Working Group can be set up.

Analysis of Non-Uniformly Degraded Pipe Sections

2003 ◽  
Author(s):  
Amy J. Smith ◽  
Keshab K. Dwivedy
Keyword(s):  
Wall Thickness ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Operating Cycle ◽  
Moment Capacity ◽  
Asme Code ◽  
Uniform Cross Section ◽  
Uniform Wall ◽  
Set Up

The management of flow assisted corrosion (FAC) has been a part of the maintenance of piping in nuclear power plants for more than 15 years. Programs have been set up to identify vulnerable locations, perform inspections, characterize the degraded configurations, and evaluate the structural integrity of the degraded sections. The section of the pipe is repaired or replaced if the structural integrity cannot be established for the projected degraded section at the next outage. During the past 15 years, significant improvements have been made to every aspect of the program including structural integrity evaluation. Simplified methods and rules are established in ASME Section XI code and in several code cases for verifying structural integrity. The evaluation of structural integrity is performed during the plant outage prior to a decision for repair or replacement. Any improvement in structural integrity evaluation to extend the life of a component by one additional operating cycle can help in performance of repair/replacement of component in a planned manner. Simplified methods and rules provided in the code can be easily used for analysis of pipe sections with degraded area with uniform wall thickness and for non-uniformly degraded sections, provided the degraded portions are modeled with uniform wall thickness equal to the lowest thickness of the section. The representation of a non-uniformly degraded section in this manner is necessarily conservative. The purpose of this paper is to develop methodology to analyze the non-uniformly degraded sections subjected to pressure and moment loading by modeling it in a manner that accounts for the non-uniform cross-section. The formulation developed here is more realistic than the code methodology and is still conservative. The results are presented in form of charts comparing the limit moment capacity of the degraded sections calculated by the formulation in this paper with that using ASME code formulation. The paper concludes that the proposed formulation can be used to supplement the ASME Code method to extend the remaining life of FAC degraded components.

Alternative RTNDT for Linde 80 Weld Materials

2002 ◽  
Author(s):  
K. K. Yoon ◽  
J. B. Hall
Keyword(s):  
Fracture Toughness ◽  
Transition Temperature ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Pressure Vessels ◽  
Screening Criteria ◽  
Master Curve Method ◽  
Asme Code ◽  
Curve Method

The ASME Boiler and Pressure Vessel Code provides fracture toughness curves of ferritic pressure vessel steels that are indexed by a reference temperature for nil ductility transition (RTNDT). The ASME Code also prescribes how to determine RTNDT. The B&W Owners Group has reactor pressure vessels that were fabricated by Babcock & Wilcox using Linde 80 flux. These vessels have welds called Linde 80 welds. The RTNDT values of the Linde 80 welds are of great interest to the B&W Owners Group. These RTNDT values are used in compliance of the NRC regulations regarding the PTS screening criteria and plant pressure-temperature limits for operation of nuclear power plants. A generic RTNDT value for the Linde 80 welds as a group was established by the NRC, using an average of more than 70 RTNDT values. Emergence of the Master Curve method enabled the industry to revisit the validity issue surrounding RTNDT determination methods. T0 indicates that the dropweight test based TNDT is a better index than Charpy transition temperature based index, at least for the RTNDT of unirradiated Linde 80 welds. An alternative generic RTNDT is presented in this paper using the T0 data obtained by fracture toughness tests in the brittle-to-ductile transition temperature range, in accordance with the ASTM E1921 standard.

Economic Evaluation of Proactive Replacement of Upper Internals Guide Tube Support Pins (Split Pins)

2005 ◽  
Author(s):  
Robert K. Perdue ◽  
G. Gary Elder ◽  
Gregory Gerzen
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Net Present Value ◽  
Time To Failure ◽  
Guide Tube ◽  
Reliability Modeling ◽  
Nuclear Plants ◽  
Zinc Addition ◽  
Primary Water ◽  
The Impact

Certain nuclear power plants have “Rev B” reactor vessel upper internals guide tube support pins, commonly referred to as split pins, made from material with properties similar to Alloy 600 and known to be susceptible to primary water stress corrosion cracking (PWSCC). This paper describes a rigorous probabilistic methodology for evaluating the economics of a preemptive replacement of these split pins, and describes an application at four of Exelon Generation’s nuclear plants. The method uses Bayesian statistical reliability modeling to estimate a Weibull time-to-failure prediction model using limited historical failures, and a Westinghouse proactive aging management simulation tool called PAM to select a split pin replacement date that would maximize the net present value of cash flow to a plant. Also in this study is a sensitivity evaluation of the impact of zinc addition on split pin replacement timing. Plant decisions made based in part on results derived from applying this approach are noted.

Application and Experience Feedback of ASME Code Case N-155-2 for RTR Piping in Nuclear Plants

2013 ◽  
Author(s):  
Nicolas d’Udekem ◽  
Philippe Art ◽  
Jacques Grisel
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Sea Water ◽  
Water System ◽  
Moderate Pressure ◽  
Raw Water ◽  
Cooling Systems ◽  
Piping Systems ◽  
Asme Code

Nowadays, the usefulness of RTR (Reinforced Thermosetting Resin) for pressure retaining equipment does not need further proof: they are lightweight, strong, with low thermal elongation and highly corrosion resistant. The use of RTR piping makes all sense for piping systems circulating raw water such as sea water at moderate pressure and temperature for plants cooling. However, this material is rarely used for safety related cooling systems in nuclear power plants. In Belgium, Electrabel and Tractebel have chosen to replace the existing carbon steel pipes of the raw water system by GRE (Glassfiber Reinforced Epoxy) pipes, in accordance with the Authorized Inspection Agency, applying the ASME Code Case (CC) N-155-2 defining the specifications and requirements for the use of RTR pipes, fittings and flanges. After a challenging qualification process, Class 3 GRE pipes are now installed and operating for raw water cooling systems in two Belgian nuclear units and will soon be installed in a third one. The paper will address the followed qualification processes and the implementation steps applied by Electrabel/Tractebel and relate the overcome obstacles encountered during manufacturing, erection and commissioning of Class 3 GRE piping in order to ensure quality, reliability and traceability required for safety equipment in nuclear power plants.

Conceptual Design of a 4 × 300 MW Modular Nuclear Plant

2011 ◽  
Author(s):  
Asko Vuorinen
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Plant ◽  
Medium Size ◽  
Modular Construction ◽  
Construction Costs ◽  
University Of Technology ◽  
Nuclear Plants ◽  
Original Estimate ◽  
Under Construction

The Finnish companies have built four medium size nuclear power plants. In addition they have constructed two nuclear icebreakers and several floating power plants. The latest 1650 MWe nuclear power plant under construction Olkiluoto-3 has had many problems, which have raised the costs of the plant to €3500/kWe from its original estimate of €2000/kWe and constriction schedule from four to eight years. It is possible to keep the costs down and schedule short by making the plant in shipyard and transport it to site by sea. The plant could be then lifted to its place by pumping seawater into the channel. This kind of concept was developed by the author in 1991, when he was making his thesis of modular gas fired power plants in Helsinki University of Technology. The modular construction of nuclear plants has made in a form of two nuclear icebreakers, which Wa¨rtsila¨ Marine has built in Helsinki Shipyard. The latest modular nuclear plant was launched in 2010 in St Petersburg shipyard. One of the benefits of modular construction is a possibility to locate the plant under rock by making the transportation channels in tunnels. This will give the plant external protection for aircraft crash and make the outer containment unnecessary. The water channels could also be used as pressure suppression pools in case of venting steam from the containment. This could reduce the radioactive releases in case of possible reactor accidents. The two 440 MW VVER plants build in Finland had construction costs of €1600 /kWe at 2011 money. The author believes that a 1200 MW nuclear plant with four 300 MW units can be constructed in five years and with €3300/kW costs, where the first plant could be generating power within 40 months and next units with 6 month intervals.

Proposal for Evaluation With Strain-Based Criterion of Impact Phenomena Between Steel Structure and Tornado Missile

2016 ◽  
Author(s):  
Yuko Sakamoto ◽  
Koji Shirai ◽  
Toshiko Udagawa ◽  
Shunsuke Kondo
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Impact Analysis ◽  
Structural Integrity ◽  
Steel Structures ◽  
Severe Accident ◽  
Numerical Code ◽  
Pressurized Water ◽  
Impact Phenomena ◽  
The Impact

In Japan, nuclear power plants must be protected from tornado missiles that are prescribed by Nuclear Regular Authority (NRA). When evaluating the structural integrity of steel structures in the plant with impact analysis by numerical code, strain-based criteria are appropriate because the tornado missiles have huge impact energy and may cause large deformation of the structures. As one of the strain-based criteria, the Japan Society of Mechanical Engineers (JSME) prescribes limiting triaxial strain for severe accident of Pressurized Water Reactor (PWR) steel containment. To confirm whether or not this criterion is appropriate to the evaluation of the impact phenomena between the steel structures and the tornado missiles, a free drop impact experiment to steel plates (carbon steel and austenitic stainless steel) was carried out with heavy weights imitated on one of the tornado missiles, followed by an impact analysis of the experiment with AUTODYN code and the JSME strain-based criterion. Consequently, it was confirmed that the strain-based criterion of JSME standard was for evaluating the fracture of steel structures caused by tornado missiles.

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