The Riddle Of Nde For Embrittlemient Detection

1997 ◽  
Vol 503 ◽  
Author(s):  
M. Blaszkiewicz
Keyword(s):  
Nuclear Reactor ◽  
Hyperfine Interactions ◽  
Pressure Vessels ◽  
Nuclear Industry ◽  
Accurate Information ◽  
Destructive Testing ◽  
Future Directions ◽  
Integrity Assessment ◽  
Nondestructive Techniques ◽  
Surveillance Programs

ABSTRACTThe ability to nondestructively determine the level of irradiation induced degradation in nuclear reactor pressure vessels (RPVs) would enhance the integrity assessment currently used by the nuclear industry. Presently, destructive testing of Charpy specimens from surveillance capsules is used to approximate the RPV upper shelf energy and the ductile-to-brittle transition temperature, and approved models and guidelines are used to determine the state of embrittlement. However, these models and surveillance programs do not always provide enough accurate information to support decisions for premature RPV life termination, life continuation to license expiration, or life renewal and extension by means of annealing. Effective nondestructive techniques would extend the usefulness of the surveillance material by reducing the amount of material used for destructive studies, and ultimately by allowing tests to be performed directly on the RPV. Nondestructive techniques, ranging from electrical resistivity to hyperfine interactions, have been, and continue to be, explored for use in embrittlement assessment. The current states of these various techniques are discussed, and future directions for research are suggested.

scholarly journals Some Useful Mechanical Property Correlations for Nuclear Reactor Pressure Vessel Steels

2018 ◽  
Author(s):  
Randy K. Nanstad ◽  
William L. Server ◽  
Mikhail A. Sokolov ◽  
G. Robert Odette ◽  
Nathan Almirall
Keyword(s):  
Mechanical Property ◽  
Nuclear Reactor ◽  
Pressure Vessels ◽  
Nuclear Industry ◽  
Rigorous Approach ◽  
Pressure Vessel Steels ◽  
Physically Based ◽  
Induced Changes ◽  
The Relationship ◽  
Reactor Pressure

The use of correlations is common in the research and development arena of the nuclear industry with the realization that some applications with direct implications to safety demand a more rigorous approach. Most correlations involve the relationship between two experimental properties, such as that between hardness and tensile strength. There are others that are much more complicated and are often designated models because they incorporate physically-based knowledge; examples of this are predictive correlations for irradiation-induced embrittlement of reactor pressure vessels (RPV). The objective of this paper is to collect and discuss many of the commonly used correlations for applications to nuclear RPVs. This paper identifies and discusses various correlations that relate easily measured properties to properties that are more difficult, more time consuming, or more expensive to measure. In the case of irradiated RPV materials, irradiation-induced changes in easily measured properties are related to the changes in those more difficult to measure. It is noted that recognition and understanding of the uncertainties associated with all correlations is highly important.

Benchmark Campaign on the Uncertainty Treatment Analysis Tools Developed for French Nuclear Reactor Pressure Vessels Integrity Assessment

2012 ◽  
Author(s):  
Emilie Dautreme ◽  
Romain Beaufils ◽  
Thomas Pernot ◽  
Emmanuel Remy ◽  
Eric Meister ◽  
...  
Keyword(s):  
Input Data ◽  
Nuclear Reactor ◽  
Software Tools ◽  
Pressure Vessels ◽  
Integrity Assessment ◽  
Analysis Tools ◽  
Good Accordance ◽  
Rational Combination ◽  
Treatment Analysis ◽  
Reactor Pressure

In France, nuclear reactor pressure vessels (RPV) integrity assessment follows a deterministic methodology and is performed according to the 1999 French regulation requirements with the use of safety coefficients. In the deterministic methodology, the input data are penalized in order to address the uncertainties. EDF and AREVA developed some couplings between integrity assessment models and uncertainty treatment software in order to address explicitly the uncertainties on the input data. In such analysis, the accumulation of conservatisms is replaced by a rational combination of parameters taking into account their uncertainty and variability. This paper presents a benchmark campaign conducted jointly by EDF and AREVA to validate the couplings between the codes and to strengthen the confidence in the calculations. The benchmark campaign was performed with several physical thermo-mechanical codes and uncertainty treatments software tools to evaluate the sensitivity of the analyses results as a function of code effects. The benchmark results are in good accordance between the several codes used. This conclusion gives confidence in the code couplings used for performing the uncertainty treatment analyses.

FAVOR Version 16.1: A Computer Code for Fracture Mechanics Analyses of Nuclear Reactor Pressure Vessels

2017 ◽  
Author(s):  
B. Richard Bass ◽  
Paul T. Williams ◽  
Terry L. Dickson ◽  
Hilda B. Klasky
Keyword(s):  
Fracture Mechanics ◽  
Nuclear Reactor ◽  
Computer Code ◽  
The United States ◽  
Pressure Vessels ◽  
Nuclear Industry ◽  
Current Status ◽  
Pressurized Water ◽  
Oak Ridge ◽  
Reactor Pressure

This paper describes the current status of the Fracture Analysis of Vessels, Oak Ridge (FAVOR) computer code which has been under development at Oak Ridge National Laboratory (ORNL), with funding by the United States Nuclear Regulatory Commission (NRC), for over twenty-five years. Including this most recent release, v16.1, FAVOR has been applied by analysts from the nuclear industry and regulators at the NRC to perform deterministic and probabilistic fracture mechanics analyses to review / assess / update regulations designed to insure that the structural integrity of aging, and increasingly embrittled, nuclear reactor pressure vessels (RPVs) is maintained throughout the vessel’s operational service life. Early releases of FAVOR were developed primarily to address the pressurized thermal shock (PTS) issue; therefore, they were limited to applications involving pressurized water reactors (PWRs) subjected to cool-down transients with thermal and pressure loading applied to the inner surface of the RPV wall. These early versions of FAVOR were applied in the PTS Re-evaluation Project to successfully establish a technical foundation that served to better inform the basis of the then-existent PTS regulations to the original PTS Rule (Title 10 of the Code of Federal Regulations, Chapter I, Part 50, Section 50.61, 10CFR 50.61). A later version of FAVOR resulting from this project (version 06.1 - released in 2006) played a major role in the development of the Alternative PTS Rule (10 CFR 50.61.a). This paper describes recent ORNL developments of the FAVOR code; a brief history of verification studies of the code is also included. The 12.1 version (released in 2012) of FAVOR represented a significant generalization over previous releases insofar as it included the ability to encompass a broader range of transients (heat-up and cool-down) and vessel geometries, addressing both PWR and boiling water reactor (BWR) RPVs. The most recent public release of FAVOR, v16.1, includes improvements in the consistency and accuracy of the calculation of fracture mechanics stress-intensity factors for internal surface-breaking flaws; special attention was given to the analysis of shallow flaws. Those improvements were realized in part through implementation of the ASME Section XI, Appendix A, A-3000 curve fits into FAVOR; an overview of the implementation of those ASME curve fits is provided herein. Recent results from an extensive verification benchmarking project are presented that focus on comparisons of solutions from FAVOR versions 16.1 and 12.1 referenced to baseline solutions generated with the commercial ABAQUS code. The verifications studies presented herein indicate that solutions from FAVOR v16.1 exhibit an improvement in predictive accuracy relative to FAVOR v12.1, particularly for shallow flaws.

Defect Tolerance Assessment for ABWR Nozzle Crotch Corner

2016 ◽  
Author(s):  
Fumihito Hirokawa ◽  
Masaaki Hayashi ◽  
Minoru Masuda ◽  
Yasuhiro Mabuchi ◽  
Yukinori Yamamoto ◽  
...  
Keyword(s):  
Structural Integrity ◽  
Pressure Vessels ◽  
Defect Size ◽  
Nuclear Industry ◽  
Evaluation Procedure ◽  
Defect Tolerance ◽  
Element Analysis ◽  
Integrity Assessment ◽  
Bending Stresses ◽  
Main Steam

In the nuclear industry, demands on the structural integrity reliability of metal components are always increasing. The quantification of allowable defects in pressure vessels should therefore draw on advanced structural integrity assessment procedures. In the UK, R6 [1] is the main procedures used for defect tolerance assessment (DTA). In this paper, the overall evaluation procedure of DTA using R6 applied to the Main Steam (MS) nozzle crotch corner of the Advanced Boiling Water Reactor (ABWR) is presented. At the nozzle crotch corner region, high stresses, including through-wall bending stresses from the local structural discontinuity, were present. These bending stresses have been categorised as secondary. R6 conservatively implies such bending stresses may need to be categorised as primary, to allow for the possibility of elastic follow-up. To support application as a secondary stress, an elastic-plastic finite element analysis has been performed to evaluate the J-integral for the nozzle crotch corner. The resulting values of J, when compared to the stress intensity factor and collapse solutions used for the assessment, showed that treating the bending stress as secondary maintained sufficient margin, indicating conservatism. Finally, the DTA results of the nozzle crotch corner are presented to determine the defect tolerance criteria. This includes calculating the limiting defect size at the start of plant life when considering the end of life critical defect size and through life Fatigue Crack Growth (FCG).

scholarly journals Quick Evaluation Method for Defect Exceeding the Allowable Flaw Size in Pressure Vessel of Nuclear Reactor for Power Plant and Space Heating

2021 ◽  
Vol 9 ◽  
Author(s):  
Pan Liu ◽  
Yuebing Li ◽  
Ting Jin ◽  
Dasheng Wang
Keyword(s):  
Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Structural Integrity ◽  
Pressure Vessels ◽  
Atmospheric Pollutants ◽  
Space Heating ◽  
Integrity Assessment ◽  
Evaluation Approach

Nuclear power can be used for power generation, space heating, and other fields, producing a limited level of greenhouse gases and no atmospheric pollutants. However, the safety of nuclear reactors is always a public concern. The reactor pressure vessels (RPVs) play an important role in the safe operation of a nuclear power plant. When a defect is inspected in the RPV, complex analytical evaluation procedures, including fatigue analysis and fracture assessment, are necessary to ensure the structural integrity of the defective component. Based on the RSE-M, a quick evaluation approach for RPVs with defects exceeding acceptance standards is proposed in this work to reduce the computational complexity and analysis time. The flaw evaluation is simplified by adjusting the inspection period based on the analysis of fatigue crack growth. The new method was applied to the RPVs with embedded defects and underclad semi-elliptical defects, respectively. The proposed evaluation approach was verified by the case of a typical RPV cylinder containing an embedded crack, where all possible transients during the operation of nuclear power plants are considered. During the allowable residual life obtained of 5-years, failure would not occur in the defective component via the conventional method, which gives confidence to the availability of the new approach. Consequently, the proposed method can be a valid reference for the structural integrity assessment of nuclear reactor components with defects exceeding acceptance standards.

Becoming the Nuclear-Front-End

2017 ◽  
Vol 47 (189) ◽  
Author(s):  
Patrick Schukalla
Keyword(s):  
Nuclear Reactor ◽  
Chemical Elements ◽  
High Technology ◽  
Power Production ◽  
Uranium Mining ◽  
Nuclear Industry ◽  
Current State ◽  
Front End ◽  
Production And Consumption ◽  
Uranium Exploration

Uranium mining often escapes the attention of debates around the nuclear industries. The chemical elements’ representations are focused on the nuclear reactor. The article explores what I refer to as becoming the nuclear front – the uranium mining frontier’s expansion to Tanzania, its historical entanglements and current state. The geographies of the nuclear industries parallel dominant patterns and the unevenness of the global divisions of labour, resource production and consumption. Clearly related to the developments and expectations in the field of atomic power production, uranium exploration and the gathering of geological knowledge on resource potentiality remains a peripheral realm of the technopolitical perceptions of the nuclear fuel chain. Seen as less spectacular and less associated with high-technology than the better-known elements of the nuclear industry the article thus aims to shine light on the processes that pre-figure uranium mining by looking at the example of Tanzania.

Comparative analysis and verification of engineering methods of shallow cracks effect for fracture toughness prediction for reactor pressure vessels

2020 ◽  
pp. 140-165
Author(s):  
V. I. Kostylev ◽  
B. Z. Margolin
Keyword(s):  
Experimental Data ◽  
Fracture Toughness ◽  
Comparative Analysis ◽  
Nuclear Reactor ◽  
Probabilistic Approach ◽  
Pressure Vessels ◽  
Local Methods ◽  
European Method ◽  
Shallow Crack ◽  
Reactor Pressure

The main features of shallow cracks fracture are considered, and a brief analysis of methods allowing to predict the temperature dependence of the fracture toughness KJC (T) for specimens with shallow cracks is given. These methods include DA-method, (JQ)-method, (J-T)-method, “local methods” with its multiparameter probabilistic approach, GP method uses power approach, and also two engineering methods – RMSC (Russian Method for Shallow Crack) and EMSC (European Method for Shallow Crack). On the basis of 13 sets of experimental data for national and foreign steels, a detailed verification and comparative analysis of these two engineering methods were carried out on the materials of the VVER and PWR nuclear reactor vessels considering the effect of shallow cracks.

French RPV PTS Assessment: An Overview of EDF Research and Development in Progress on Thermalhydraulic, Materials and Mechanical Aspects

2003 ◽  
Author(s):  
Dominique Moinereau ◽  
Jean-Michel Frund ◽  
Henriette Churier-Bossennec ◽  
Georges Bezdikian ◽  
Alain Martin
Keyword(s):  
Structural Integrity ◽  
Three Dimensional ◽  
Crack Arrest ◽  
Pressure Vessels ◽  
Short Description ◽  
Constraint Effect ◽  
Materials Properties ◽  
Integrity Assessment ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

A significant extensive Research & Development work is conducted by Electricite´ de France (EDF) related to the structural integrity re-assessment of the French 900 and 1300 MWe reactor pressure vessels in order to increase their lifetime. Within the framework of this programme, numerous developments have been implemented or are in progress related to the methodology to assess flaws during a pressurized thermal shock (PTS) event. The paper contains three aspects: a short description of the specific French approach for RPV PTS assessment, a presentation of recent improvements on thermalhydraulic, materials and mechanical aspects, and finally an overview of the present R&D programme on thermalhydraulic, materials and mechanical aspects. Regarding the last aspect on present R&D programme, several projects in progress will be shortly described. This overview includes the redefinition of some significant thermalhydraulic transients based on some new three-dimensional CFD computations (focused at the present time on small break LOCA transient), the assessment of vessel materials properties, and the improvement of the RPV PTS structural integrity assessment including several themes such as warm pre-stress (WPS), crack arrest, constraint effect ....

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