Consideration of Special Effects for the Application of an Optimized Fracture Mechanics Approach for the RPV Assessment: Project CAMERA

2020 ◽  
Author(s):  
Florian Obermeier ◽  
Hieronymus Hein ◽  
Johannes May ◽  
Julia Kobiela ◽  
Marco Kaiser
Keyword(s):  
Fracture Toughness ◽  
Fracture Mechanics ◽  
Transition Region ◽  
Crack Extension ◽  
Master Curve ◽  
Pressure Vessels ◽  
Ductile Material ◽  
Pressurized Water ◽  
Special Effects ◽  
Material Tests

Abstract A large database for fracture toughness data according to the standard ASTM E1921 in the brittle and brittle to ductile transition region was generated within the former research programs CARINA and CARISMA for materials used in western pressurized water reactor (PWR) reactor pressure vessels (RPV). These programs confirmed successfully the application of the Master Curve approach for German RPVs. With respect to the RPV proof of safety during plant operation, in particular for integrity assessments considering the event of a Pressurized Thermal Shock (PTS) further relevant issues appeared in terms of the use of suitable fracture toughness curves for components and of the quantification of safety margins for the irradiated material state: • Validity of fracture toughness curve and Master Curve respectively in the irradiated state at higher test temperatures (ductile material region) • Verification of the WPS effect (warm pre-stress) in the irradiated state for representative loading paths and materials • Impact of material inhomogeneities on the fracture toughness • Transferability of specimen-specific effects on the safety-related RPV integrity assessment To address these open issues a follow-up project was initiated called CAMERA (Consideration of special effects for the application of an optimized fracture mechanics approach for the RPV safety assessment). One main aspect within this program is to assess the possibility of describing the whole fracture toughness curve including the upper shelf based on empirical T0 correlations and the effect of a load case related warm pre-loading. Material tests are carried out and evaluated according to ASTM E1820 to get the crack extension resistance as a function of stable crack extension (J-R curves) in the transition to the ductile region and according to ASTM E1921 (WPS) in the lower shelf region and transition region. The least-square fit curves (J-R) as well as the effect of warm pre-stress are determined for several types of irradiated and unirradiated RPV base and weld metals. The results serve to expand the application window of the Master Curve concept to the ductile region and to complete the fracture toughness curve over the entire temperature range from lower shelf until operating temperature also considering the load scenario based on the effect of warm pre-stress. This paper will summarize and discuss the results of the material tests performed.

Estimation of Through-Wall Cracking Frequency of RPV Under PTS Events Using PFM Analysis Method for Identifying Conservatism Included in Current Japanese Code

2014 ◽  
Author(s):  
Kazuya Osakabe ◽  
Koichi Masaki ◽  
Jinya Katsuyama ◽  
Genshichiro Katsumata ◽  
Kunio Onizawa
Keyword(s):  
Fracture Toughness ◽  
Fracture Mechanics ◽  
Crack Initiation ◽  
Structural Integrity ◽  
Probabilistic Approach ◽  
Pressure Vessels ◽  
Assessment Procedure ◽  
Analysis Method ◽  
Integrity Assessment ◽  
The U.S

To assess the structural integrity of reactor pressure vessels (RPVs) during pressurized thermal shock (PTS) events, the deterministic fracture mechanics approach prescribed in Japanese code JEAC 4206-2007 [1] has been used in Japan. The structural integrity is judged to be maintained if the stress intensity factor (SIF) at the crack tip during PTS events is smaller than fracture toughness KIc. On the other hand, the application of a probabilistic fracture mechanics (PFM) analysis method for the structural reliability assessment of pressure components has become attractive recently because uncertainties related to influence parameters can be incorporated rationally. A probabilistic approach has already been adopted as the regulation on fracture toughness requirements against PTS events in the U.S. According to the PFM analysis method in the U.S., through-wall cracking frequencies (TWCFs) are estimated taking frequencies of event occurrence and crack arrest after crack initiation into consideration. In this study, in order to identify the conservatism in the current RPV integrity assessment procedure in the code, probabilistic analyses on TWCF have been performed for certain model of RPVs. The result shows that the current assumption in JEAC 4206-2007, that a semi-elliptic axial crack is postulated on the inside surface of RPV wall, is conservative as compared with realistic conditions. Effects of variation of PTS transients on crack initiation frequency and TWCF have been also discussed.

Uncertainties in Pressurized Thermal Shock Analyses

2019 ◽  
Author(s):  
M. Niffenegger ◽  
O. Costa Garrido ◽  
D. F. Mora ◽  
G. Qian ◽  
R. Mukin ◽  
...  
Keyword(s):  
Fracture Mechanics ◽  
Thermal Shock ◽  
Probabilistic Approach ◽  
Pressure Vessels ◽  
Pressurized Water Reactor ◽  
Ansys Fluent ◽  
Pressurized Water ◽  
Integrity Assessment ◽  
Safety Margins ◽  
Pressurized Thermal Shock

Abstract Integrity assessment of reactor pressure vessels (RPVs) can be performed either by deterministic fracture mechanics (DFM) or/and by probabilistic fracture mechanics (PFM) analyses. In European countries and Switzerland, only DFM analyses are required. However, in order to establish the probabilistic approach in Switzerland, the advantages and shortcomings of the PFM are investigated in the frame of a national research project. Both, the results from DFM and PFM depend strongly on the previous calculated thermal-hydraulic boundary conditions. Therefore, complete integrity analyses involving several integrated numerical codes and methods were performed for a reference pressurized water reactor (PWR) RPV subjected to pressurized thermal shock (PTS) loads. System analyses were performed with the numerical codes RELAP5 and TRACE, whereas for structural and fracture mechanics calculations, the FAVOR and ABAQUS codes were applied. Additional computational fluid dynamics analyses were carried out with ANSYS/FLUENT, and the plume cooling effect was alternatively considered with GRS-MIX. The results from the different analyses tools are compared, to judge the expected overall uncertainty and reliability of PTS safety assessments. It is shown that the scatter band of the stress intensities for a fixed crack configuration is rather significant, meaning that corresponding safety margins should be foreseen. The conditional probabilities of crack initiation and RPV failure might also differ, depending on the considered random parameters and applied rules.

Fracture mechanics in design and service: ‘living with defects’ - Fracture mechanics as an aid to design and operation of nuclear plant

1981 ◽  
Vol 299 (1446) ◽  
pp. 227-236 ◽  
Keyword(s):  
Fracture Mechanics ◽  
Material Properties ◽  
Pressure Vessels ◽  
Nuclear Plant ◽  
Plant Design ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Plant Safety ◽  
Usual Criterion ◽  
Design Protection

Fracture mechanics analyses are an important part of nuclear plant design, supplementing the conventional design protection against failure to cover the possibility of the presence of crack-like defects. The degree of detail and accuracy required for a particular application depends on the possible consequences of a failure and whether the assessment is concerned with plant safety or with aspects of reliability. In the former case, a conservative approach is necessary and the prevention of initiation is the usual criterion. This approach is typified by the safety assessment applied to pressurized water reactor pressure vessels, which is outlined and discussed in relation to elastic plastic approaches and the importance of plant transient conditions, material properties (especially in weldments) and possible defect distributions. Fracture mechanics can help in defining quality control and quality assurance procedures, including both requirements for mechanical property appraisal and nondestructive testing. The latter aspects extend into operation, in respect of monitoring of plant conditions, surveillance of changes in material properties and the use of periodic inspection and plant condition monitoring techniques. A number of examples are quoted and recommendations made to permit improved fracture mechanics assessments.

Prediction of Fracture Toughness KIC Transition Curves of Pressure Vessel Steels From Charpy V-Notch Impact Test Results

1994 ◽  
Vol 116 (4) ◽  
pp. 353-358 ◽  
Author(s):  
T. Iwadate ◽  
Y. Tanaka ◽  
H. Takemata
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Master Curve ◽  
Prediction Method ◽  
Pressure Vessels ◽  
Life Extension ◽  
Low Alloy Steels ◽  
Excess Temperature ◽  
Pressure Vessel Steels ◽  
Transition Curves

A single and generalized prediction method of fracture toughness KIC transition curves of pressure vessel steels has been greatly desired by engineers in the petro-chemical and nuclear power industries, especially from the viewpoint of life extension of reactor pressure vessels. In this paper, the toughness degradation of Cr-Mo steels during long-term service was examined and the two prediction methods of fracture toughness KIC transition curves were studied using the data of 54 heats. 1) The toughness degradation of 2 1/4Cr-1Mo steels levels off within around 50,000 h service. 2) The FATT versus J-factor (=(Si+Mn)(P+Sn)×104) and/or X (=(10P+5Sb+4Sn+As)x10−2) relationships to estimate the maximum embrittlement of Cr-Mo steels were obtained. 3) A master curve method developed by authors et al.; that is, the method using a KIC/KIC−US versus excess temperature master curve of each material was presented for 2 1/4Cr-1Mo, 1 1/4Cr-1/2Mo, 1Cr and 1/2Mo chemical pressure vessel steels and ASTM A508 C1.1, A508 C1.2, A508 C1.3 and A533 Gr.B C1.1 nuclear pressure vessel steels, where KIC−US is the upper-shelf fracture toughness and excess temperature is test temperature minus FATT. 4) A generalized prediction method to predict the KIC transition curves of any low-alloy steels was developed. This method consists of KIC/KIC−US versus T–T0 master curve and temperature shift ΔT between fracture toughness and CVN impact transition curves versus yield strength relationship, where To is the temperature showing 50 percent KIC−US of the material. 5) The KIC transition curves predicted using both methods showed a good agreement with the lower bound of measured KJC values obtained from JC tests.

Adjustments to Master Curve Methodology and Development of Fracture Toughness Estimation

2011 ◽  
Author(s):  
R. S. Kulka
Keyword(s):  
Fracture Toughness ◽  
Fracture Mechanics ◽  
Driving Force ◽  
Master Curve ◽  
Structural Components ◽  
Fracture Test ◽  
Element Analysis ◽  
Master Curve Method ◽  
Curve Method ◽  
Constraint Effects

In conventional fracture mechanics assessments, there is often an inadequate treatment of in-plane constraint effects on the apparent toughness of structural components, leading to significant conservatism. Modifications to the Master Curve method, to account for these effects, have previously been suggested. A study of these proposed modifications has identified that less conservative toughness estimates could be made from the analysis of fracture mechanics test specimens. An approach has been developed for allowing a comparison of a variation of fracture toughness values throughout a component, to a variation of the localised effective driving force. Cracked-body finite element analysis has been used to assess fracture test specimens with varying levels of in-plane constraint, to provide fracture mechanics data for use with the approach that has been developed.

Correlations Between Charpy V-Notch Impact Energy and Fracture Toughness of Nuclear Reactor Pressure Vessel (RPV) Steels

2015 ◽  
Author(s):  
Meifang Yu ◽  
Zhen Luo ◽  
Y. J. Chao
Keyword(s):  
Fracture Toughness ◽  
Transition Region ◽  
Test Data ◽  
Impact Energy ◽  
Master Curve ◽  
Reference Temperature ◽  
Master Curve Method ◽  
Curve Method ◽  
Reactor Pressure ◽  
Actual Test

Both Charpy V-notch (CVN) impact energy and fracture toughness are parameters reflecting toughness of the material. Charpy tests are however easy to perform compared to standard fracture toughness tests, especially when the material is irradiated and quantity is limited. Correlations between the two parameters are therefore of great significance, especially for reactor pressure vessel (RPV) structural integrity assessment. In this paper, correlations between CVN impact energy and fracture toughness of three commonly used RPV steels, namely Chinese A508-3 steel, USA A533B steel, Euro 20MnMoNi55 steel, are investigated with two methods. One method applies a direct conversion using empirical formulas and the other adopts the Master Curve method. It is found that when the empirical formula is used, the difference between the predicted fracture toughness (from the CVN impact energy) and actual test data is relatively small in upper shelf, lower shelf and the bottom of transition region, while relatively large in other parts of the transition region. When the Master Curve method is adopted, whether the reference temperature T0 is estimated through temperature at 28J or 41J CVN impact energy, the predicted fracture toughness values of the three steels are consistent with actual test data. The reference temperature T0 is also estimated through the IGC-parameter correlation and through a combination of empirical formula and multi-temperature method. Both procedures show excellent agreement with test results. The mean value of T0 estimated from T28J, T41J, IGC-parameters and the combination method is denoted by TQ-ave and is then used as the final reference temperature T0 for the Master Curve determination. Accuracy of TQ-ave (and therefore the Master Curve method) is demonstrated by comparison with actual test data of the three RPV steels. It is concluded that Master Curve method provides a reliable procedure for predicting fracture toughness in the transition region utilizing limited CVN impact energy data from open literature.

Crack Initiation and Propagation in Static Loaded Fracture Mechanics Tests in Steels Containing Atomic Hydrogen

2020 ◽  
Author(s):  
Philippa L. Moore ◽  
Menno Hoekstra ◽  
Alex Pargeter
Keyword(s):  
Fracture Toughness ◽  
Fracture Mechanics ◽  
Crack Initiation ◽  
Plastic Zone ◽  
Crack Growth ◽  
Crack Extension ◽  
Low Alloy Steels ◽  
Ductile Tearing ◽  
Initiation Fracture Toughness ◽  
Stable Tearing

Abstract Hydrogen is well known to have a detrimental influence on the ductility of low alloy steels, reducing the fracture toughness. Standard test methods to characterize fracture toughness of steels in terms of ductile tearing resistance curves have not been developed to account for any hydrogen-driven contribution to the crack extension, Δa. Simply plotting J or CTOD against Δa is not necessarily appropriate for defining the initiation fracture toughness for tests performed in a hydrogen-charging environment. This paper explores a method to further analyse experimental data collected during fracture toughness tests, which allows the contribution of plasticity (i.e. when blunting precedes ductile tearing) to be considered separately from the initiation of crack extension (which could be by stable tearing and/or by hydrogen-driven crack extension). The principle is based on the assumption that a crack growing by a hydrogen-driven mechanism in a quasi-static fracture mechanics test performed in environment may not be associated with significant ductility in the plastic zone (which would accompany crack growth by stable tearing). The analytical method presented in this paper compares the different points of deviation from linear behavior of the components of J, to isolate the effects of ductility within the plastic zone from pure crack extension. In this way, the point of crack initiation can be defined in order to determine the relevant initiation fracture toughness; whether by blunting and stable tearing, or by hydrogen-driven crack growth. This approach offers a screening method which is illustrated using examples of fracture mechanics specimens tested in environments of varying severity (air, seawater with cathodic protection, and sour service). This method can be used to identify the relevant definition of initiation fracture toughness while allowing for a combination of ductile tearing, hydrogen-driven crack extension, or both, to be present during the test.

Study on Life Extension of Aged RPV Material Based on Probabilistic Fracture Mechanics: Japanese Round Robin

1995 ◽  
Vol 117 (1) ◽  
pp. 7-13 ◽  
Author(s):  
G. Yagawa ◽  
S. Yoshimura ◽  
N. Handa ◽  
T. Uno ◽  
K. Watashi ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Fracture Mechanics ◽  
Surface Crack ◽  
Round Robin ◽  
Life Extension ◽  
Loading Conditions ◽  
Pressurized Water ◽  
Unique Problem ◽  
Probabilistic Fracture Mechanics ◽  
Failure Probabilities

This paper is concerned with round-robin analyses of probabilistic fracture mechanics (PFM) problems of aged RPV material. Analyzed here is a plate with a semi-elliptical surface crack subjected to various cyclic tensile and bending stresses. A depth and an aspect ratio of the surface crack are assumed to be probabilistic variables. Failure probabilities are calculated using the Monte Carlo methods with the importance sampling or the stratified sampling techniques. Material properties are chosen from the Marshall report, the ASME Code Section XI, and the experiments on a Japanese RPV material carried out by the Life Evaluation (LE) subcommittee of the Japan Welding Engineering Society (JWES), while loads are determined referring to design loading conditions of pressurized water reactors (PWR). Seven organizations participate in this study. At first, the procedures for obtaining reliable PFM solutions with low failure probabilities are examined by solving a unique problem with seven computer programs. The seven solutions agree very well with one another, i.e., by a factor of 2 to 5 in failure probabilities. Next, sensitivity analyses are performed by varying fracture toughness values, loading conditions, and pre and in-service inspections. Finally, life extension simulations based on the PFM analyses are performed. It is clearly demonstrated from these analyses that failure probabilities are so sensitive to the change of fracture toughness values that the degree of neutron irradiation significantly influences the judgment of plant life extension.

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