Use of Master Curve Technology for Assessing Shallow Flaws in a Reactor Pressure Vessel Material

2008 ◽  
Vol 130 (3) ◽  
Author(s):  
N. Taylor ◽  
P. Minnebo ◽  
B. R. Bass ◽  
D. Siegele ◽  
K. Wallin ◽  
...  
Keyword(s):  
Pressure Vessel ◽  
Master Curve ◽  
Biaxial Loading ◽  
Production Quality ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Test Pieces ◽  
Fracture Tests ◽  
Vessel Material ◽  
Reactor Pressure

In the NESC-IV project, an experimental/analytical program was performed to develop validated analysis methods for transferring fracture toughness data to shallow flaws in reactor pressure vessels subject to biaxial loading in the lower-transition temperature region. Within this scope, an extensive range of fracture tests was performed on material removed from a production-quality reactor pressure vessel. The master curve analysis of these data is reported and its application to the assessment of the project feature tests on large beam test pieces is discussed.

Use of Master Curve Technology for Assessing Shallow Flaws in a Reactor Pressure Vessel Material

2006 ◽  
Author(s):  
N. Taylor ◽  
P. Minnebo ◽  
R. B. Bass ◽  
D. Siegele ◽  
K. Wallin ◽  
...  
Keyword(s):  
Pressure Vessel ◽  
Master Curve ◽  
Biaxial Loading ◽  
Production Quality ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Test Pieces ◽  
Fracture Tests ◽  
Vessel Material ◽  
Reactor Pressure

In the NESC-IV project an experimental/analytical program was performed to develop validated analysis methods for transferring fracture toughness data to shallow flaws in reactor pressure vessels subject to biaxial loading in the lower-transition temperature region. Within this scope an extensive range of fracture tests was performed on material removed from a production-quality reactor pressure vessel. The Master Curve analysis of this data is reported and its application to the assessment of the project feature tests on large beam test pieces.

Accounting for the Effects of Inservice Inspection Flaw Depth Sizing Errors in Developing Flaw Size Distribution Tables

2009 ◽  
Author(s):  
S. R. Gosselin ◽  
F. A. Simonen
Keyword(s):  
Pressure Vessel ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Expert Elicitation ◽  
Flaw Size ◽  
Physical Models ◽  
Asme Code ◽  
Valid Comparison ◽  
Vessel Material ◽  
Reactor Pressure

Probabilistic fracture mechanics studies have addressed reactor pressure vessels that have high levels of material embrittlement. These calculations have used flaw size and density distributions determined from precise and optimized laboratory measurements made and validated with destructive methods as well as from physical models and expert elicitation. The experimental data were obtained from reactor vessel material samples removed from cancelled plants (Shoreham and the Pressure Vessel Research Users Facility (PVRUF)). Consequently, utilities may need to compare the numbers and sizes of reactor pressure vessel flaws identified by the plant’s inservice inspection program to the numbers and sizes of flaws assumed in prior failure probability calculations. This paper describes a method to determine whether the flaws in a particular reactor pressure vessel are consistent with the assumptions regarding the number and sizes of flaws used in other analyses. The approach recognizes that ASME Code Section XI examinations suffer from limitations in terms of sizing errors for very small flaws. Direct comparisons of a vessel specific flaw distribution with other documented flaw distributions would lead to pessimistic conclusions. This paper provides a method for a valid comparison that accounts for flaw sizing errors present in ASME Code Section XI examinations.

Overview of NESC-IV Cruciform Specimen Test Results

2002 ◽  
Author(s):  
N. Taylor ◽  
R. Bass
Keyword(s):  
Pressure Vessel ◽  
Best Practice ◽  
Biaxial Loading ◽  
Production Quality ◽  
Standard Test ◽  
Reactor Pressure Vessel ◽  
Test Analysis ◽  
Post Test ◽  
Specimen Test ◽  
Reactor Pressure

NESC-IV is an experimental/analytical program to develop validated analysis methods for transferring fracture toughness data generated on standard test specimens to shallow flaws in reactor pressure vessel welds subject to biaxial loading in the lower-transition temperature region. It is the fourth major project of the Network for Evaluating Structural Components (NESC). The testing program has exploited material removed from a production-quality reactor pressure vessel. In Part A six clad cruciform specimens containing shallow surface-breaking flaws located in weld material were successfully tested. For Part B a further four beam tests were performed using an innovative test piece design with a simulated embedded flaw. Post-test analysis is now in progress. The implications for current best-practice procedures for evaluation of RPV shallow are also being considered.

Fracture Analysis of Surface-Breaking Shallow Flaw Behaviour in the NESC-IV Bend Beam Tests

2004 ◽  
Author(s):  
N. Taylor ◽  
I. Sattari-Far ◽  
D. Siegele ◽  
I. Varfolomeyev ◽  
L. Stumpfrock
Keyword(s):  
Fracture Mechanics ◽  
Pressure Vessel ◽  
Biaxial Loading ◽  
Production Quality ◽  
Standard Test ◽  
Reactor Pressure Vessel ◽  
Weld Material ◽  
Post Test ◽  
Analytical Program ◽  
Reactor Pressure

NESC-IV is an experimental/analytical program to develop validated analysis methods for transferring fracture toughness data generated on standard test specimens to shallow flaws in reactor pressure vessel welds subject to biaxial loading in the lower-transition temperature region. The testing program exploited material from a production-quality reactor pressure vessel. In Part A six clad cruciform specimens containing shallow surface-breaking flaws located in weld material were successfully tested. Post-test fracture mechanics analyses have been conducted by several organizations to determine fracture mechanics parameters such as K, J, Q and T at the flaws. These have been used to interpret the results with respect to the Master Curve, giving particular attention to constraint aspects.

Failure Probability Assessment for a Boiling Water Reactor Pressure Vessel Under Low Temperature Over-Pressure Event

2012 ◽  
Author(s):  
Hsoung-Wei Chou ◽  
Chin-Cheng Huang ◽  
Bo-Yi Chen ◽  
Hsien-Chou Lin ◽  
Ru-Feng Liu
Keyword(s):  
Low Temperature ◽  
Failure Probability ◽  
Pressure Vessel ◽  
Fracture Probability ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Reactor Pressure

The fracture probability of a boiling water reactor pressure vessel for a domestic nuclear power plant in Taiwan has been numerically analyzed using an advanced version of ORNL’s FAVOR code. First, a model of the vessel beltline region, which includes all shell welds and plates, is built for the FAVOR code based on the plant specific parameters of the reactor pressure vessel. Then, a novel flaw model which describes the flaw types of surface breaking flaws, embedded weld flaws and embedded plate flaws are simulated along both inner and outer vessel walls. When conducting the fracture probability analyses, a transient low temperature over-pressure event, which has previously been shown to be the most severe challenge to the integrity of boiling water reactor pressure vessels, is considered as the loading condition. It is found that the fracture occurs in the fusion-line area of axial welds, but with only an insignificant failure probability. The low through-wall cracking frequency indicates that the analyzed reactor pressure vessel maintains sufficient stability until either the end-of-license or for doubling of the present license of operation.

Measurements of Stress During Thermal Shock in Clad Reactor Pressure Vessel Material Using Time-Resolved In-Situ Synchrotron X-Ray Diffraction

2018 ◽  
Author(s):  
Sam Oliver ◽  
Chris Simpson ◽  
Andrew James ◽  
Christina Reinhard ◽  
David Collins ◽  
...  
Keyword(s):  
Thermal Shock ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Pressure Vessel Steel ◽  
X Ray Diffraction ◽  
Time Resolved ◽  
Vessel Steel ◽  
X Ray ◽  
Reactor Pressure

Nuclear reactor pressure vessels must be able to withstand thermal shock due to emergency cooling during a loss of coolant accident. Demonstrating structural integrity during thermal shock is difficult due to the complex interaction between thermal stress, residual stress, and stress caused by internal pressure. Finite element and analytic approaches exist to calculate the combined stress, but validation is limited. This study describes an experiment which aims to measure stress in a slice of clad reactor pressure vessel during thermal shock using time-resolved synchrotron X-ray diffraction. A test rig was designed to subject specimens to thermal shock, whilst simultaneously enabling synchrotron X-ray diffraction measurements of strain. The specimens were extracted from a block of SA508 Grade 4N reactor pressure vessel steel clad with Alloy 82 nickel-base alloy. Surface cracks were machined in the cladding. Electric heaters heat the specimens to 350°C and then the surface of the cladding is quenched in a bath of cold water, representing thermal shock. Six specimens were subjected to thermal shock on beamline I12 at Diamond Light Source, the UK’s national synchrotron X-ray facility. Time-resolved strain was measured during thermal shock at a single point close to the crack tip at a sample rate of 30 Hz. Hence, stress intensity factor vs time was calculated assuming K-controlled near-tip stress fields. This work describes the experimental method and presents some key results from a preliminary analysis of the data.

Investigation of the Beltline Welding Seam and Base Metal of the Greifswald WWER-440 Unit 1 Reactor Pressure Vessel

2009 ◽  
Author(s):  
Jan Schuhknecht ◽  
Hans-Werner Viehrig ◽  
Udo Rindelhardt
Keyword(s):  
Weld Metal ◽  
Base Metal ◽  
Pressure Vessel ◽  
Master Curve ◽  
Reactor Pressure Vessel ◽  
Reference Temperature ◽  
Metal Ring ◽  
Welding Seam ◽  
Root Region ◽  
Reactor Pressure

The investigation of reactor pressure vessel (RPV) materials from decommissioned NPPs offers the unique opportunity to scrutinize the irradiation behaviour under real conditions. Material samples taken from the RPV wall enable a comprehensive material characterisation. The paper describes the investigation of trepans taken from the decommissioned WWER-440 first generation RPVs of the Greifswald NPP. Those RPVs represent different material conditions such as irradiated (I), irradiated and recovery annealed (IA) and irradiated, recovery annealed and re-irradiated (IAI). The working program is focussed on the characterisation of the RPV steels (base and weld metal) through the RPV wall. The key part of the testing is aimed at the determination of the reference temperature T0 following the ASTM Test Standard E1921-05 to determine the fracture toughness of the RPV steel in different thickness locations. In a first step the trepans taken from the RPV Greifswald Unit 1 containing the X-butt multilayer submerged welding seam and from base metal ring 0.3.1 both located in the beltline region were investigated. Unit 1 represents the IAI condition. It is shown that the Master Curve approach as adopted in ASTM E1921 is applicable to the investigated original WWER-440 weld metal. The evaluated T0 varies through the thickness of the welding seam. The lowest T0 value was measured in the root region of the welding seam representing a uniform fine grain ferritic structure. Beyond the welding root T0 shows a wavelike behaviour. The highest T0 of the weld seam was not measured at the inner wall surface. This is important for the assessment of ductile-to-brittle temperatures measured on sub size Charpy specimens made of weld metal compact samples removed from the inner RPV wall. Our findings imply that these samples do not represent the most conservative condition. Nevertheless, the Charpy transition temperature TT41J estimated with results of sub size specimens after the recovery annealing was confirmed by the testing of standard Charpy V-notch specimens. The evaluated Charpy-V TT41J shows a better accordance with the irradiation fluence along the wall thickness than the Master Curve reference temperature T0. The evaluated T0 from the trepan of base metal ring 0.3.1 varies through the thickness of the RPV wall. T0 increases from −120°C at the inner surface to −104°C at a distance of 33 mm from it and again to −115°C at the outer RPV wall. The KJc values generally follow the course of the MC, although the scatter is large. The re-embrittlement during 2 campaigns operation can be assumed to be low for the weld and base metal.

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