Use of Mini-CT Specimens for Fracture Toughness Characterization of Low Upper-Shelf Linde 80 Weld

2017 ◽  
Author(s):  
Mikhail A. Sokolov
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessels ◽  
Charpy Specimen ◽  
Good Correspondence ◽  
Asme Code ◽  
Surveillance Programs ◽  
Impact Data ◽  
Reactor Pressure ◽  
Fracture Toughness Specimen

Any fracture toughness specimen that can be made out of the broken halves of standard Charpy specimens may have exceptional utility for evaluation of reactor pressure vessels since it would allow one to determine and monitor directly actual fracture toughness instead of requiring indirect predictions using correlations established with impact data. The Charpy V-notch specimen is the most commonly used specimen geometry in surveillance programs and most likely to be used in advanced reactors as per ASME code. The advantage of the Mini-CT specimen technique is that multiple specimens can be machined from one half of a broken Charpy specimen, used in a standard surveillance capsule of a reactor pressure vessel. Up to now, most of the work on validation of this type of the specimens has been performed on base metal. In this study, Mini-CT specimens were used to perform fracture toughness characterization of low upper-shelf Linde 80 weld, designated WF-70. This weld was utilized in the Midland beltline weld and has been previously well characterized at ORNL with various types and sizes of fracture toughness specimens. The Mini-CT specimens were machined from broken previously tested Charpy V-notch specimens. Despite very small size and relatively small number of Mini-CT specimen tested, the transition fracture toughness temperature, To, derived from these Mini-CT specimens is in very good correspondence with To reported from analysis of a large number of larger fracture toughness specimens.

Use of Small Specimens for Evaluation of Irradiated Materials

2013 ◽  
Author(s):  
Mikhail A. Sokolov ◽  
Randy K. Nanstad
Keyword(s):  
Fracture Toughness ◽  
Small Volume ◽  
Radioactive Material ◽  
Surveillance Programs ◽  
Series Of Experiments ◽  
Point Bend ◽  
Impact Data ◽  
Surveillance Specimens ◽  
Reactor Pressure ◽  
Fracture Toughness Specimen

Small specimens are playing the key role in evaluating properties of irradiated materials. The use of small specimens provides several advantages. Typically, only small volume of material can be irradiated in a reactor at desirable conditions in terms of temperature, neutron flux, and neutron dose. Small volume of irradiated material may also allow for easier handling of specimens. Smaller specimens reduce the amount of radioactive material, minimizing personnel exposures and waste disposal. However, use of small specimens imposes variety of challenges as well. These challenges are associated with proper accounting for size effects and transferability of small specimen data to the real structures of interest. The PCVN specimen as well as any fracture toughness specimen that can be made out of the broken halves of standard Charpy specimens may have exceptional utility for evaluation of RPVs. The Charpy V-notch specimen is the most commonly used specimen geometry in surveillance programs. Precracking and testing of Charpy surveillance specimens would allow one to determine and monitor directly actual fracture toughness instead of requiring indirect predictions using correlations established with impact data. However, there is a growing number of indications that there might be a bias in the reference fracture toughness transition temperature, To values derived from PCVN and compact specimens. The present paper summarizes data from the series of experiments that use subsize specimens for evaluation of the transition fracture toughness of reactor pressure vessel (RPV) steels. Two types of compact specimens and three types of three-point bend specimens from five RPV materials were used in these subsize experiments. The current results showed that To determined from PCVN specimens with width (W) to thickness (B) ratio W/B = 1, on average, are lower than To determined from compact specimens with W/B = 2. At the same time, three-point bend specimens with W/B = 2 exhibited To values that were very similar to To values derived from compact specimens. Constraint corrections developed by Dodds et al. are applied to assess the bias.

Fracture Toughness Characterization of Low Upper-Shelf Linde 80 Weld Using Mini-C(T) Specimens

2018 ◽  
Author(s):  
Michael R. Ickes ◽  
J. Brian Hall ◽  
Robert G. Carter
Keyword(s):  
Fracture Toughness ◽  
Direct Measurement ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Weld Metals ◽  
Surveillance Programs ◽  
Test Reactor ◽  
Reactor Pressure

The Charpy V-notch specimen is the most commonly used specimen geometry in reactor pressure vessel irradiation surveillance programs and there is an extensive stored inventory of irradiated broken Charpy specimens. The advantage of the mini-C(T) (4mm thick C(T)) specimen technique is that multiple specimens can be machined from each half of broken irradiated Charpy specimens. Fracture toughness specimens that can be machined from broken halves of standard Charpy specimens enable the direct measurement of fracture toughness which can be used for engineering evaluation of reactor pressure vessels. Work to validate the mini-C(T) specimens has been performed mostly on unirradiated reactor pressure vessel base and weld metals . In this study, mini-C(T) specimens were tested providing fracture toughness characterization of an irradiated low upper-shelf Linde 80 weld (WF-70). This weld was utilized in the Midland beltline and has been previously well characterized at ORNL with various types and sizes of fracture toughness specimens. The mini-C(T) specimens were machined from broken previously tested Charpy V-notch size specimens which were irradiated in a material test reactor. The effect of different methods of measuring the displacement on the results is assessed. The ASTM E1921 results are compared to previous test data produced from larger fracture toughness specimens. In addition, the sensitivity of T0 to the ASTM E1921 censoring value is discussed.

On Bias in To Values Derived With Compact and PCVN Specimens

2004 ◽  
Author(s):  
Mikhail A. Sokolov ◽  
Randy K. Nanstad
Keyword(s):  
Fracture Toughness ◽  
National Laboratory ◽  
Oak Ridge ◽  
Surveillance Programs ◽  
Series Of Experiments ◽  
Point Bend ◽  
Impact Data ◽  
Surveillance Specimens ◽  
Reactor Pressure ◽  
Fracture Toughness Specimen

The Heavy-Section Steel Irradiation (HSSI) Program at Oak Ridge National Laboratory (ORNL) includes a task to investigate the bias in the reference fracture toughness transition temperature values, To, derived with the pre-cracked Charpy (PCVN) and compact specimens. The PCVN specimen, as well as any other fracture toughness specimen that can be made out of the broken Charpy specimens, may have exceptional utility for the evaluation of RPV steels. The Charpy V-notch specimen is the most commonly used specimen geometry in surveillance programs. Precracking and testing of Charpy surveillance specimens would allow one to determine and monitor directly actual fracture toughness instead of requiring indirect predictions using correlations established with impact data. However, there are a growing number of indications that there might be a bias in To values derived from PCVN and compact specimens. The present paper summarizes data from the series of experiments that use subsize specimens for evaluation of the transition fracture toughness of reactor pressure vessel (RPV) steels conducted within the HSSI Program. Two types of compact specimens and three types of three-point bend specimens from five RPV materials were used in these subsize experiments. The current results showed that To determined from PCVN specimens with width (W) to thickness (B) ratio W/B=1, on average, are lower than To determined from compact specimens with W/B=2. At the same time, three-point bend specimens with W/B=2 exhibited To values that were very similar to To values derived from compact specimens.

Surveillance Specimen Programs for WWER Type Reactors

2005 ◽  
Author(s):  
Milan Brumovsky
Keyword(s):  
Mechanical Properties ◽  
Pressure Vessel ◽  
Pressure Vessels ◽  
Surveillance Program ◽  
Reactor Safety ◽  
Reactor Radiation ◽  
Critical Comparison ◽  
Asme Code ◽  
Surveillance Programs ◽  
Reactor Pressure

Reactor pressure vessels (RPV) are components with the highest importance for the reactor safety and operation as they contain practically whole inventory of fission material but they are damaged/aged during their operation by an intensive reactor radiation. Surveillance specimen programs are the best method for monitoring changes in mechanical properties of reactor pressure vessel materials if they are designed and operated in such a way that they are located in conditions close to those of the vessels. Reactor Codes and standards usually included requirements and conditions for such programs to assure proper vessel monitoring. WWER (Water-Water-Energetic Reactors) reactor pressure vessels are designed according to former Russian Codes and rules with somewhat different requirements using different materials comparing e.g. with ASME Code. Two principal types of WWER reactors were designed, manufactured and are operated in several European countries (and also in China, Iran): WWER-440 and WWER-1000. Their surveillance programs were designed in quite different way, with some modifications due to the time, country of manufacturing and experience gained from their operation. The paper gives a critical comparison of these programs in both types of reactors with requirements of both Russian and ASME/ASTM Codes and Standards. Finally, information about creation of the Integral Surveillance Program for WWER-1000 type reactor pressure vessels covering vessels from several countries is described.

Applicability of Miniature C(T) Specimen to Evaluation of Fracture Toughness for Reactor Pressure Vessel Steel

2013 ◽  
Author(s):  
Kentaro Yoshimoto ◽  
Takatoshi Hirota ◽  
Hiroyuki Sakamoto ◽  
Takuji Sugihara ◽  
Shohei Sakaguchi ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Evaluation Method ◽  
Pressure Vessels ◽  
Charpy Specimen ◽  
Pressure Vessel Steel ◽  
Fracture Toughness Test ◽  
Irradiation Embrittlement ◽  
Vessel Steel ◽  
Reactor Pressure Vessel Steel ◽  
Reactor Pressure

Irradiation embrittlement of Japanese reactor pressure vessels (RPV) is usually monitored by conducting tests on irradiated RPV material according to surveillance test program. Although fracture toughness specimens are contained in Japanese PWR surveillance capsule, the number of specimens is limited due to capacity of capsule. In order to evaluate lower bound of fracture toughness considering its scatter with higher reliability, it is expected to obtain additional fracture toughness data using remaining broken specimens of irradiated materials. One of solutions to this problem is specimen reconstitution technique. However, it is difficult to make numbers of specimens by reconstitution because of need for specific equipments and time-consuming machining operations. As an alternative method, fracture toughness test using miniature C(T) specimens with dimension of 4×10×10mm, which can be taken from broken halves of Charpy specimen, is proposed and the studies to verify the reliability and robustness of evaluation method have been conducted in the Japanese round robin program since 2010. In this study, fracture toughness tests were performed on Japanese SA 533 Gr.B Cl.1 steel using miniature C(T) specimens and the effect of specimen size on reference temperature T0 was studied by the Master Curve approach. In addition, the issues related to application to irradiated materials were discussed.

Applications of the Master Curve Approach to Irradiated Steels

2004 ◽  
Author(s):  
A. Ballesteros ◽  
J. Bros ◽  
M. Brumovsky
Keyword(s):  
Fracture Toughness ◽  
Master Curve ◽  
Pressure Vessels ◽  
Current Situation ◽  
Ferritic Steels ◽  
Preliminary Results ◽  
Fundamental Understanding ◽  
Asme Code ◽  
Reactor Pressure ◽  
Open Issues

The Master Curve approach is being introduced in the evaluation of the fracture toughness of low-alloy ferritic steels used in reactor pressure vessels. The Master Curve provides significant technical enhancements compared with the methodology currently incorporated in the ASME code. The fundamental understanding of Master Curve has been achieved, but application issues are emerging facing the integrity evaluation of reactor pressure vessels. The current situation of the Master Curve methodology is reviewed in this paper, including the last developments for WWER reactors. The open issues are discussed, and the preliminary results from investigations being carried out in Spain are presented.

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.

scholarly journals Use of Mini-CT Specimens for Fracture Toughness Characterization of Low Upper-Shelf Linde 80 Weld Before and After Irradiation

2018 ◽  
Author(s):  
Mikhail A. Sokolov
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Master Curve ◽  
Collaborative Program ◽  
Before And After ◽  
Minimum Number ◽  
Real Practice ◽  
Reactor Pressure ◽  
Good Agreement

Mini-CT specimens are becoming a highly popular geometry for use in reactor pressure vessel (RPV) community for direct measurement of fracture toughness in the transition region using the Master Curve methodology. In the present study, Mini-CT specimens were machined from previously tested Charpy specimens of the Midland low upper-shelf Linde 80 weld in both, unirradiated and irradiated conditions. The irradiated specimens have been characterized as part of a joint ORNL-EPRI-CRIEPI collaborative program. The Linde 80 weld was selected because it has been extensively characterized in the irradiated condition by conventional specimens, and because of the need to validate application of Mini-CT specimens for low upper-shelf materials — a more likely case for some irradiated materials of older generation RPVs. It is shown that the fracture toughness reference temperatures, To, derived from these Mini-CT specimens are in good agreement with To values previously recorded for this material in the unirradiated and irradiated conditions. However, this study indicates that in real practice it is highly advisable to use a much larger number of specimens than the minimum number prescribed in ASTM E1921.

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