An Investigation Into the Feasibility of Environmental Fatigue Design for APR 1400 Primary Components and Piping With a 60 Year Design Life

2003 ◽  
Author(s):  
J.-S. Park ◽  
J.-M. Kim ◽  
G.-S. Kim ◽  
T.-S. Choi
Keyword(s):  
Design Methodology ◽  
Low Alloy Steel ◽  
Fatigue Design ◽  
Design Life ◽  
Division 1 ◽  
Asme Code ◽  
Alternative Approaches ◽  
Surge Line ◽  
Coolant System ◽  
Reactor Pressure

This paper investigates the feasibility of environmental fatigue design for the APR1400 with the 60 year design life, which is based on sample evaluations of fatigue lives for the component and piping designed to the ASME Code Section III, Division 1. The materials sampled from the reactor coolant system (RCS) components are the low alloy steel for reactor pressure vessel outlet nozzles and the austenitic stainless steel for pressurizer surge line piping. Environmental fatigue evaluations of the component materials are performed employing the environment factor approach. Based on the evaluation results it is concluded the environmental fatigue design of the RCS components and piping for the APR1400 is not feasible as the Code requirement, and alternative approaches need be developed removing various conservatisms in the fatigue design methodology.

Environmental Fatigue Evaluation for Primary Components of APR1400 Reactor Coolant System

2010 ◽  
Author(s):  
J. M. Kim ◽  
K. W. Kim ◽  
K. S. Yoon ◽  
S. H. Park ◽  
I. Y. Kim ◽  
...  
Keyword(s):  
Environmental Effects ◽  
Integral Method ◽  
Power Reactor ◽  
Light Water Reactor ◽  
Outlet Nozzle ◽  
Design Life ◽  
Class 1 ◽  
Asme Code ◽  
Fatigue Evaluation ◽  
Coolant System

USNRC Regulatory Guide (RG) 1.207 provides a guideline for evaluating fatigue analyses due to the environmental effects on the new light water reactor (LWR). The environmental correction factor (Fen) is used to incorporate the LWR environmental effect into fatigue analyses of ASME Class 1 components. In this paper, the environmental fatigue evaluation is applied to some primary components with 60 year design life of Advanced Power Reactor (APR1400). The materials sampled from Class 1 components are the low alloy steel for the reactor vessel (RV) outlet nozzle and the carbon steel for the hot leg which are attached to the outlet nozzle. The simplified method, time-based integral method and strain-based integral method are used to compute the Fen values. The calculated fatigue usage factors including the environmental effects are compared with those obtained using the current ASME Code rules. As the calculated cumulative fatigue usage factor considering environmental effects (CUFen) is below 1.0, there is no concern for the RV outlet nozzle to implement design for environmental fatigue effects.

Basis of the Upcoming Changes to the Piping Inelastic Fatigue Design Rules in the ASME Boiler and Pressure Vessel Code Section III, Division 1, Article NB-3600

2015 ◽  
Author(s):  
Timothy M. Adams
Keyword(s):  
Pressure Vessel ◽  
Service Level ◽  
Fatigue Analysis ◽  
Elastic Plastic ◽  
Fatigue Design ◽  
Division 1 ◽  
Class 1 ◽  
Asme Code ◽  
Code Changes ◽  
Near Future

In conducting a Class 1 piping analysis per the simplified rules of the ASME Boiler and Pressure Vessel Code, Section III, Division 1, Article NB-3600, a fatigue analysis is required per paragraph NB-3653 for both Service Level A and Service Level B. The fatigue analysis provides two options. The options are dependent on Equation 10 of subparagraph NB-3653.1. If this equation is met for a given load set pair under consideration, then the analysis proceeds directly to subparagraphs NB-3653.2 through NB-3653.5. If however, Equation 10 is exceeded, the Code allows the use of a simplified Elastic Plastic Analysis as delineated in subparagraph NB-3653.6. The first requirement of NB-3653.6 is that both Equation 12 and Equation 13 must be met. The changes in the seismic design in the last 25+ years have not been appropriately reflected in the subparagraph NB-3653.6(b) Equation 13. Also, the Code provides no clear guidance on seismic anchor motions in paragraph NB-3650. In 2012 ASME Code Committees undertook an action to address these issues. This paper provides the background and basis for Code changes that are anticipated will be implemented in the near future in paragraph NB-3653.6 of the ASME Boiler and Pressure Vessel Code, Section III, Division 1 that will address both of these issues. This implementation will make the Elastic Plastic Fatigue rules of NB-3653.6 consistent with the design by analysis approach of NB-3228.5.

Development of a Fatigue Design Curve for Austenitic Stainless Steels in LWR Environments: A Review

2002 ◽  
Author(s):  
Omesh K. Chopra
Keyword(s):  
Stainless Steels ◽  
Pressure Vessel ◽  
Type Strain ◽  
Nuclear Power ◽  
Austenitic Stainless Steels ◽  
Environmental Parameters ◽  
Light Water Reactor ◽  
Code Design ◽  
Fatigue Design ◽  
Asme Code

The ASME Boiler and Pressure Vessel Code provides rules for the construction of nuclear power plant components and specifies fatigue design curves for structural materials. However, the effects of light water reactor (LWR) coolant environments are not explicitly addressed by the Code design curves. Existing fatigue strain–vs.–life (ε–N) data illustrate potentially significant effects of LWR coolant environments on the fatigue resistance of pressure vessel and piping steels. This paper reviews the existing fatigue ε–N data for austenitic stainless steels in LWR coolant environments. The effects of key material, loading, and environmental parameters, such as steel type, strain amplitude, strain rate, temperature, dissolved oxygen level in water, and flow rate, on the fatigue lives of these steels are summarized. Statistical models are presented for estimating the fatigue ε–N curves for austenitic stainless steels as a function of the material, loading, and environmental parameters. Two methods for incorporating environmental effects into the ASME Code fatigue evaluations are presented. Data available in the literature have been reviewed to evaluate the conservatism in the existing ASME Code fatigue design curves.

Suggested Improvements to Appendix G of ASME Section XI Code

2008 ◽  
Author(s):  
Hardayal S. Mehta ◽  
Timothy J. Griesbach ◽  
Gary L. Stevens
Keyword(s):  
Thermal Gradient ◽  
Research Council ◽  
Ductile Failure ◽  
Calculation Methods ◽  
Technical Literature ◽  
Corner Crack ◽  
The Past ◽  
Asme Code ◽  
Reactor Pressure ◽  
Calculation Procedures

This paper reviews some of the original basis documents for ASME Section XI Nonmandatory Appendix G for calculating pressure-temperature (P-T) limits and recommends areas for improvement. The original Appendix G in Section XI of ASME Code was mainly based on Welding Research Council (WRC) Bulletin 175 (WRC-175). Changes have been made to Appendix G over the past 20 years such as the use of the KIC reference toughness curve instead of KIR. However, aspects of the Appendix G method still refer back to WRC Bulletin 175. The published technical literature since the development of WRC 175 could be used to enhance the Appendix in a number of areas. One such area is stress intensity factor (K) calculation procedures for thermal gradient loading at a nozzle corner. This paper will review and evaluate the available K calculation methods for a nozzle corner crack, and develop closed-form expressions for incorporation into Appendix G. Also, the following areas will be reviewed: (1) treatment of operating stresses exceeding the material yield stress, and (2) fracture toughness criteria typically used for other than reactor pressure vessel (RPV) and piping for protection against non-ductile failure. This paper will also identify areas for future improvements in Appendix G.

Investigation of Nozzle on Knuckle Region of Dished Head

2021 ◽  
Author(s):  
Sujay S. Pathre ◽  
Ameya M. Mathkar ◽  
Shyam Gopalakrishnan
Keyword(s):  
Stress Pattern ◽  
Design Specification ◽  
Specific Location ◽  
Bending Effect ◽  
Division 1 ◽  
User Design ◽  
Asme Code ◽  
Section Viii ◽  
Design Requirements ◽  
Induced Stresses

Abstract ASME Code Section VIII Division 1 [1] provides rules for the shape of openings, size of openings, strength and design of openings, however, the existing rules do not provide any restrictions on the specific location of the nozzle on the dished head knuckle region. Many corporate guidelines/ user design requirements meant for pressure vessel design and specification suggest avoiding placement of any type of nozzle in the knuckle area of a dished head and generally state in their design specification to limit the placement of a nozzle including its reinforcement within the crown area. This applies to Torispherical and Ellipsoidal dished heads. Code [1] rule UG-37(a) provides the benefit in reinforcement by reducing the required thickness (tr) of the dished head when the nozzle is in the spherical portion of the dished head for the Ellipsoidal and Torispherical dished head. High stresses occur in the knuckle region of the dished head due to the edge bending effect caused as the cylinder and head try to deform in different directions. For various reasons the user design requirements insist on placing the nozzle in the knuckle region, further compounding the complexity of the stress pattern in the knuckle area. The work carried out in this paper was an attempt to check whether it is safe to locate a nozzle in the knuckle region of the dished head since the knuckle portion is generally subjected to higher stresses in comparison to the crown portion of a dished head and the Code [1] and [2] does not impose any restrictions for the placement of nozzles in the knuckle region. Also, in this paper an attempt was made to evaluate the induced stresses when equivalent sizes of nozzles are placed in the crown as well as the knuckle portion of the dished head.

Protecting children from internet pornography? A critical assessment of statutory age verification and its enforcement in the UK

2019 ◽  
Vol 43 (1) ◽  
pp. 183-197
Author(s):  
Majid Yar
Keyword(s):  
Design Methodology ◽  
Regulatory Regime ◽  
Content Type ◽  
Critical Response ◽  
Online Pornography ◽  
Policy And Strategy ◽  
Online Behaviour ◽  
Alternative Approaches ◽  
Age Verification ◽  
The Uk

Purpose The purpose of this paper is to critically assess the newly created regulatory and policing regime for age-restricting access to pornography in the UK. Design/methodology/approach It examines the pivotal legislation, policy and strategy documents, consultation submissions and interventions from a range of stakeholders such as children’s charities, content providers and privacy advocates. Findings Even before its implementation, the regulatory regime betrays serious flaws and shortcomings in its framing and configuration. These difficulties include its inability to significantly curtail minors’ access to online pornography and risks of privacy violations and associated harms to legitimate users’ interests. Research limitations/implications Remedial measures are available so as to address some of the problems identified. However, it is argued that ultimately the attempt to prohibit minors from accessing such content is set to fail, and that alternative approaches – such as better equipping children through education to cope with explicit materials online – need to be given greater prominence. Originality/value This paper provides the first criminological policy analysis of this latest attempt to regulate and police online behaviour, and offers an important critical response to such efforts.

Component Reliability Considerations for New Designs and Extended Operation of Boiling Water Reactor (BWRs)

2008 ◽  
Author(s):  
E. Kiss
Keyword(s):  
Stainless Steel ◽  
Alloy Steel ◽  
High Reliability ◽  
Pressure Vessels ◽  
Operational Experience ◽  
Low Alloy Steel ◽  
Irradiation Embrittlement ◽  
Component Reliability ◽  
Extended Operation ◽  
Reactor Pressure

To achieve high reliability for new designs and extended operation of Reactor Pressure Vessels and Internals it is mandatory to apply the technical knowledge gained during operation of the existing Plants to assure that sufficient “Margin” is built into the new design. This paper discusses the importance of four key structural degradation mechanisms that have been shown by operational experience to affect the reliability of the BWR. These are: 1) Stress Corrosion Cracking (IGSCC) of Stainless Steel and Nickel-based Alloys; 2) Irradiation Assisted SCC (IASCC) of Stainless Steel and Nickel-based Alloys; 3) Irradiation Embrittlement of RPV low alloy Steel; 4) Corrosion Assisted Fatigue of Carbon and Low Alloy Steel. While the focus of this paper is the BWR, the mechanisms discussed are equally applicable to the PWR, although the water chemistry effects and mitigations will be different.

Comparative Study on Fatigue Damage Assessment of a Structure Member in a Bulk Carrier Using Various Environmental Conditions

2019 ◽  
Author(s):  
Fredhi Agung Prasetyo ◽  
Naoki Osawa ◽  
Mohammad Arif Kurniawan ◽  
Siti Komariyah
Keyword(s):  
Fatigue Damage ◽  
Environmental Conditions ◽  
Damage Assessment ◽  
North Atlantic Ocean ◽  
Environmental Condition ◽  
Scatter Diagram ◽  
Sea State ◽  
Fatigue Design ◽  
Design Life

Abstract Specific design life could be identified by using fatigue damage assessment in the structure engineering field as well as in the maritime sector. Fatigue assessment is one of the assessments to be conducted during review of ship structure design. Fatigue assessment of ship structural member is mainly conducted based on specific environmental condition. In general, specific environmental condition, which is provided by Classification Society rules, is a long term sea-state data of North Atlantic Ocean. The wave scatter diagram presents the tabulation of a long term data of sea state history in the specific ocean. Therefore, a realistic encounter of wave scatter diagram is essential to simulate the variation of wave loadings applied on the ship structure in determination of fatigue design life. Since the application of North Atlantic ocean environmental condition is commonly used by major Classification societies, this condition might give the substantial deterioration on the fatigue design life of the ship that specially operate only in specific ocean area, i.e. South East Asia area. In this work, the wave scatter diagram of various environmental conditions is chosen and the statistical characteristic is compared. The wave load sequence that is used on the fatigue damage assessment are generated by using the concept of storm model, so that the changing nature of sea state could be emulated as in real ocean. Fatigue damage of a structure member of 220 meter Bulk Carriers is calculated based on various environmental conditions.

Consideration of Environmental Fatigue in the ASME Code for Carbon and Low-Alloy Steel Components

2003 ◽  
Author(s):  
Stan T. Rosinski ◽  
Arthur F. Deardorff ◽  
Robert E. Nickell
Keyword(s):  
Fatigue Life ◽  
Alloy Steel ◽  
Water Environment ◽  
Operating Experience ◽  
Nuclear Industry ◽  
Operating Conditions ◽  
Low Alloy Steel ◽  
Reactor Water ◽  
Asme Code ◽  
The Impact

The potential impact of reactor water environment on reducing the fatigue life of light water reactor (LWR) piping components has been an area of extensive research. While available data suggest a reduction in fatigue life when laboratory samples are tested under simulated reactor water environments, reconciliation of this data with plant operating experience, plant-specific operating conditions, and established ASME Code design processes is necessary before a conclusion can be reached regarding the need for explicit consideration of reactor water environment in component integrity evaluations. U.S. nuclear industry efforts to better understand this issue and ascertain the impact, if any, on existing ASME Code guidance have been performed through the EPRI Materials Reliability Program (MRP). Based on the MRP activities completed to date there is no need for explicit incorporation of reactor water environmental effects for carbon and low-alloy steel components in the ASME Code. This paper summarizes ongoing MRP activities and presents the technical arguments for resolution of the environmental fatigue issue for carbon and low-alloy steel locations.

Study on Flaw Acceptance Standard of ASME Code Sec. XI Based on Failure Probability

2004 ◽  
Author(s):  
Katsuyuki Shibata ◽  
Kunio Onizawa ◽  
YinSheng Li ◽  
Yasuhiro Kanto ◽  
Shinobu Yoshimura
Keyword(s):  
Case Studies ◽  
Failure Probability ◽  
Pressure Vessels ◽  
Aspect Ratios ◽  
Pressurized Thermal Shock ◽  
Asme Code ◽  
Order Of Magnitude ◽  
The Difference ◽  
Reactor Pressure

Based on the failure probability, the flaw acceptance standard of ASME Code Sec. XI is examined with some concerns weather the failure probability is uniform for flaws with various aspect ratios and failure frequencies are small enough. In this paper, the results of preliminary case studies are described on the failure probability of reactor pressure vessels (RPVs) with a surface flaw specified in Sec. XI. PFM code PASCAL was used for case studies. A PTS (Pressurized Thermal Shock) transient prescribed by NRC/EPRI PTS Benchmark Study was used as an applied load. Analysis results showed that the conditional failure probability of a RPV with an initial flaw of acceptable depth depends on the aspect ratio. In the case flaw shapes are close to semi-circular, the failure probability are higher than that of the cases aspect ration are less than 0.6 by one order of magnitude due to the difference of fracture behavior at the surface point. A case study for determining the acceptable flaws based on failure probability was also carried out.

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