ASME Section XI Appendix L Flaw Tolerance Evaluation of Pressurized Water Reactor Piping Systems to Support Second License Renewal (80-Years Operation)

2018 ◽  
Author(s):  
Anees Udyawar ◽  
Charles Tomes ◽  
Alexandria Carolan ◽  
Steve Marlette ◽  
Thomas Meikle ◽  
...  
Keyword(s):  
Crack Growth ◽  
Structural Integrity ◽  
Operating Conditions ◽  
Pressurized Water Reactor ◽  
Safe Operation ◽  
Pressurized Water ◽  
Water Reactor ◽  
Flaw Tolerance ◽  
Piping Systems ◽  
Operating Plant

One of the goals of ASME Section XI is to ensure that systems and components remain in safe operation throughout the service life, which can include plant license extensions and renewals. This goal is maintained through requirements on periodic inspections and operating plant criteria as contained in Section XI IWB-2500 and IWB-3700, respectively. Operating plant fatigue concerns can be caused from operating conditions or specific transients not considered in the original design transients. ASME Section XI IWB-3740, Operating Plant Fatigue Assessments, provides guidance on analytical evaluation procedures that can be used when the calculated fatigue usage exceeds the fatigue usage limit defined in the original Construction Code. One of the options provided in Section XI Appendix L is through the use of a flaw tolerance analysis. The flaw tolerance evaluation involves postulation of a flaw and predicting its future growth, and thereby establishing the period of service for which it would remain acceptable to the structural integrity requirements of Section XI. The flaw tolerance approach has the advantage of not requiring knowledge of the cyclic service history, tracking future cycles, or installing systems to monitor transients and cycles. Furthermore, the flaw tolerance can also justify an inservice inspection period of 10 years, which would match a plant’s typical Section XI in-service inspection interval. The goal of this paper is to demonstrate a flaw tolerance evaluation based on ASME Section XI Appendix L for several auxiliary piping systems for a typical PWR (Pressurized Water Reactor) nuclear power plant. The flaw tolerance evaluation considers the applicable piping geometry, materials, loadings, crack growth mechanism, such as fatigue crack growth, and the inspection detection capabilities. The purpose of the Section XI Appendix L evaluation is to demonstrate that a reactor coolant piping system continues to maintain its structural integrity and ensures safe operation of the plant.

Crack Growth Behavior of Nickel Alloy Welds in a PWR Environment

2007 ◽  
Author(s):  
B. Alexandreanu ◽  
O. K. Chopra ◽  
W. J. Shack
Keyword(s):  
Crack Growth ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Crack Growth Behavior ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Cracking Behavior ◽  
Water Reactor ◽  
Ni Alloy ◽  
The Individual

A program is under way at Argonne National Laboratory to evaluate the resistance of Ni alloys and their welds to environmentally assisted cracking in simulated Light Water Reactor (LWR) coolant environments. This paper focuses on the cracking behavior of Ni-alloy welds in simulated pressurized water reactor (PWR) environment at 290–350°C. Crack growth tests have been conducted on both field- and laboratory-produced welds. The results are compared with the existing crack-growth-rate (CGR) data for Ni-alloy welds to determine the relative susceptibility of specific Ni-alloy welds to environmentally enhanced cracking. To analyze the CGRs, a superposition model was used to establish the individual contributions of mechanical fatigue, corrosion fatigue, and stress corrosion cracking.

Environmental Crack Growth Predictions for Piping Components

2007 ◽  
Author(s):  
Sam Ranganath ◽  
Guy DeBoo
Keyword(s):  
Stainless Steel ◽  
Crack Growth ◽  
Test Data ◽  
Structural Integrity ◽  
Reference Curve ◽  
Pressurized Water ◽  
Water Reactor ◽  
Integrity Assessment ◽  
Asme Code ◽  
Reference Curves

Structural integrity assessment of reactor components requires consideration of crack growth. A key input to this is the development of reference stress corrosion crack (SCC) growth rate curves for use in the structural evaluation. The ASME Section XI Task Group on SCC Reference Curve is looking into available SCC data for stainless steel and nickel based alloys and associated weldment in both pressurized water reactor (PWR) and boiling water reactor (BWR) environments. The test data show significant data scatter in crack growth rates (CGR). The conservative approach is to develop reference curves that bound all available data so that upper bound crack growth predictions. While this approach may be conservative, it may lead to excessive estimates of crack growth and result in unrealistic (and often meaningless) structural margin predictions. Selection of the appropriate SCC reference curves requires realistic interpretation of test data so that the predictions are consistent with field behavior and provide reasonable, but conservative assessment. This paper describes crack growth assessment for stainless steel piping and Alloy 600 safe end components with Alloy 182/82 welds in BWR environment. The results from the crack growth analysis for piping can be used to determine whether a proposed reference curve provides reasonable results. The objective is to use the piping and safe end crack growth predictions to develop optimal SCC Reference Curves for use in ASME Code evaluations.

Aging Management Strategies for Pressurized Water Reactor Vessel Internals

2004 ◽  
Author(s):  
Timothy J. Griesbach ◽  
Robert E. Nickell ◽  
H. T. Tang ◽  
Jeff D. Gilreath
Keyword(s):  
Reactor Vessel ◽  
Performance Criteria ◽  
Pressurized Water Reactor ◽  
Aging Effects ◽  
Pressurized Water ◽  
Water Reactor ◽  
Flaw Tolerance ◽  
License Renewal ◽  
Technical Basis ◽  
Reactor Internals

Management of materials aging effects, such as loss of material, reduction in fracture toughness, or cracking, depends upon the demonstrated capability to detect, evaluate, and potentially correct conditions that could affect function of the internals during the license renewal term. License renewal applicants in their submittals to NRC have identified the general elements of aging management programs for Pressurized Water Reactor (PWR) internals, including the use of inservice inspection and monitoring with the possibility of enhancement or augmentation if a relevant condition is discovered. As plants near the license renewal term, plant-specific aging management programs will be implemented focusing on those regions most susceptible to aging degradation. A framework for the implementation of an aging management program is proposed in this paper. This proposed framework is based on current available research results and state of knowledge and utilizes inspections and flaw tolerance evaluations to manage the degradation issues. The important elements of this framework include: • The screening of components for susceptibility to the aging mechanisms, • Performing functionality analyses of the components with representative material toughness properties under PWR conditions, • Evaluating flaw tolerance of lead components or regions of greatest susceptibility to cracking, loss of toughness, or swelling, and • Using focused inspections to demonstrate no loss of integrity in the lead components or regions of the vessel internals. The EPRI Material Reliability Program (MRP) Reactor Internals Issue Task Group (RI-ITG) is actively working to develop the data and methods to quantify an understanding of aging and potential degradation of reactor vessel internals, to develop materials/components performance criteria, and to provide utilities tools for extending plant operations. Under this MRP Program, the technical basis for the framework will be documented. Then, based on that technical basis, PWR internals inspection and flaw evaluation guidelines will be developed for plants to manage reactor internals aging and associated potential degradation.

Effect of Uneven Crack Front on Crack Tip Mechanics and the Implication to Stress Corrosion Crack Growth

2009 ◽  
Author(s):  
He Xue ◽  
Zhanpeng Lu ◽  
Hiroyoshi Murakami ◽  
Tetsuo Shoji
Keyword(s):  
Growth Rate ◽  
Crack Growth Rate ◽  
Stress Corrosion ◽  
Crack Growth ◽  
Crack Front ◽  
Pressurized Water Reactor ◽  
Element Analysis ◽  
Pressurized Water ◽  
Water Reactor ◽  
Nickel Base

Uneven crack fronts have been observed in laboratory stress corrosion cracking tests. For example, cracking fronts of nickel-base alloys tested in simulated boiling water reactor (BWR) and pressurized water reactor (PWR) environments could exhibit uneven crack front. Analyzing the effect of an uneven crack front on further crack growth is important for quantification of crack growth. Finite-Element analysis shows that the local KI distribution can be significantly affected by the shape and size of the uneven crack front. Stress intensity factor at the locally extended crack front can be significantly reduced. Since generally there is a nonlinear CGR versus KI relationship, it is expected that crack growth rate at the locally extended crack front can be significantly different from those in the neighboring areas. There could be several patterns for the growth of an uneven crack front. For example, once initiated, the crack growth rate in areas other than the locally protruded front would become higher and then the whole crack front would tend to become uniform. On the other hand, if the crack growth in other areas is still low, there is a possibility that the crack growth rate at the front tip would slow down.

Validation of Ultrasonic Inspections for Sizewell ‘B’ Pressurized Water Reactor

1989 ◽  
Vol 203 (1) ◽  
pp. 59-64
Author(s):  
D K Cartwright
Keyword(s):  
United Kingdom ◽  
Structural Integrity ◽  
High Standard ◽  
Atomic Energy ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Atomic Energy Authority ◽  
Circuit Components

To assist in ensuring a high standard of structural integrity of the Sizewell ‘B’ pressure circuit, components are inspected both during fabrication and in service. The more critical inspections are to be validated. This paper describes the validations of ultrasonic inspections to be carried out at the Inspection Validation Centre, United Kingdom Atomic Energy Authority, Risley.

scholarly journals SOURCE TERM ASSESSMENT FOR 100 MWE PRESSURIZED WATER REACTOR

2020 ◽  
Vol 22 (2) ◽  
pp. 61
Author(s):  
Pande Made Udiyani ◽  
Muhammad Budi Setiawan
Keyword(s):  
Power Plants ◽  
Public Perception ◽  
Source Term ◽  
Fission Products ◽  
Operating Conditions ◽  
Normal Operation ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Starting Point

One of the barriers on the implementation of nuclear energy in Indonesia is public perception towards the safety of nuclear power plants (NPPs). Therefore, it is necessary to perform a study about the radiation impact of normal and abnormal operations of an NPP. In accordance to the program of Ministry of Research and Technology period 2020-2024, concerning the plan to build a small modular reactor (SMR)-type NPP, a radiation safety study has been performed for the 100 MWe Pressurized Water Reactor (PWR-100MWe). Source term release of radioactive substances into the environment from PWR-100MWe is a starting point in the study of the radiological consequences of reactor operation. Therefore, this paper will examine the PWR-100MWe source term under normal and abnormal operating conditions, according to the design and the design basis accident (DBA). The initial trigger of the DBA is Lost of Coolant Accident (LOCA) such as Small LOCA and Large LOCA.  Due to the limitations of available SMR data, the study of PWR-100MWe source term refers to the assumption of the release fraction of fission products per subsystem in a larger 1000MWe PWR. It is expected from this assumption that pessimistic source term will be obtained. The study begins with calculation of PWR-100MWe core inventory using ORIGEN2 code based on PWR-100MWe reactor parameters. Through the mechanistic source term model and PWR-1000MWe release parameters, source terms will be obtained for normal operation and abnormal conditions i.e. DBA. Normal source term is used to calculate the consequences of normal operation, which will be used for environmental monitoring and environmental safety analysis of the site. Whereas accident source term is the basis for calculating the radiological consequences of accidents used for SAR documents and nuclear preparedness.Keywords: SMR, PWR-100MWe, normal operation, source term, accident

Simulated Environmental Tests for Selected ATF Cladding Solutions

2021 ◽  
Author(s):  
Sami Penttila ◽  
Juha-Matti Autio ◽  
Jari Lydman ◽  
Aki Toivonen ◽  
Seppo Peltonen ◽  
...  
Keyword(s):  
High Temperature ◽  
Base Material ◽  
Operating Conditions ◽  
Advanced Technology ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
High Temperature Furnace ◽  
Normal Light ◽  
C 30

Abstract Current development on advanced technology fuel (ATF) claddings is aiming at improved high temperature integrity of new cladding solutions that are based on the existing zirconium claddings. To assess their performance for commercial use, their thorough characterization is essential. The primary requirement for the cladding materials is the ability to tolerate loss of cooling for a significant period without failing. The tests in this work were performed on different types of coated Zr-alloys in a high temperature furnace in flowing steam conditions at 1100 °C/ 60 min, 1200 °C/ 30 min and 1300 °C / 5 min. In addition, exposures were performed in pressurized water reactor (PWR) water chemistry to confirm the material viability in normal light water reactor (LWR) operating conditions. After PWR and steam tests, the exposed specimens were studied using a Zeiss Crossbeam 540 field emission gunscanning electron microscope (FEG-SEM) equipped with a semi-quantitative energy dispersive X-ray spectrometer (EDS). Most of the tested specimens indicated detached coating layer. Varying amounts of cracking in the coatings were present. Some of the cracks extended into the base material. Based on this study, further development is needed.

Sign in / Sign up

Export Citation Format

Share Document