Analysis of Crack Front Length Effect Applicable to the Evaluation of Fracture Toughness of Base Metal for PWR Pressure Vessel Steel

2014 ◽  
Author(s):  
Pierre Joly ◽  
Claude Benhamou ◽  
Antoine Andrieu ◽  
Henriette Churier-Bossenec ◽  
Aurore Parrot
Keyword(s):  
Fracture Toughness ◽  
Size Effect ◽  
Base Metal ◽  
Crack Front ◽  
Pressure Vessel ◽  
Pressure Vessel Steel ◽  
Fracture Toughness Test ◽  
Reference Curve ◽  
Length Effect ◽  
Vessel Steel

The objective of the present paper is to review the crack front length effect (also called “specimen size effect”) on Fracture Toughness of PWR Reactor Pressure Vessel Steel base metal. The analysis of the reality and amplitude of this effect is conducted in a first step on a database (the so called “GKSS database”) including fracture toughness test results on a single representative material using specimens of different thicknesses, tested in the same temperature range. A realistic analytical form for describing the size effect observed in this data set is thus derived from statistical analyses and proposed for engineering application. In a second step, this size effect formulation is then applied to a large number of fracture toughness data, obtained in Irradiation Surveillance Programs, and also to the numerous data used for the definition of the ASME (and RCC-M) fracture toughness reference curves. This analysis allows normalizing all the available fracture toughness data with a single specimen width of 100 mm and defining the fracture toughness reference curve as the lower bound of this normalized set of data points. It is thus demonstrated that the fracture toughness reference curve is associated with a reference crack length of 100 mm, and can be used in RPV integrity analyses for other crack front length in association with the crack front length correction formula defined in the first step.

Fracture Toughness Evaluation of Heat-Affected Zone Under Weld Overlay Cladding in Reactor Pressure Vessel Steel

2018 ◽  
Author(s):  
Yoosung Ha ◽  
Tohru Tobita ◽  
Hisashi Takamizawa ◽  
Satoshi Hanawa ◽  
Yutaka Nishiyama
Keyword(s):  
Fracture Toughness ◽  
Base Metal ◽  
Heat Affected Zone ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Pressure Vessel Steel ◽  
Vessel Steel ◽  
Weld Overlay ◽  
Reactor Pressure Vessel Steel ◽  
Reactor Pressure

An evaluation of the fracture toughness of the heat-affected zone (HAZ), which is located under the weld overlay cladding of a reactor pressure vessel (RPV), was performed. Considering inhomogeneous microstructures of the HAZ, 0.4T-C(T) specimens were manufactured from the cladding strips locations, and Mini-C(T) specimens were fabricated from the distanced location as well as under the cladding. The reference temperature (To) of specimens that were aligned with the middle section of a cladding strip (HAZMCS) was ∼12°C higher than that of specimens that were aligned with cladding strips at the overlap (HAZOCS). To values of partial area in the HAZ were obtained using Mini-C(T) specimen. The To values obtained near the side of the cladding were ∼13°C higher than those away from the cladding. To values of HAZ for both 0.4T-C(T) and Mini-C(T) specimens were significantly lower than that of the base metal at a quarter thickness by 40°C–60°C. Compared to the literature data that indicated fracture toughness at the surface without overlay cladding and base metal of a quarter thickness in a pressure vessel plate, this study concluded that the welding thermal history showed no significant effect on the fracture toughness of the inner surface of RPV steel.

Comparison of Predicted Transition Temperature Shifts Between Static Fracture Toughness and Charpy-V Impact Properties Due to Irradiation for an A508 Pressure Vessel Steel

2004 ◽  
Author(s):  
B. Tanguy ◽  
J. Besson ◽  
C. Bouchet ◽  
S. Bugat
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Material Model ◽  
Pressure Vessel Steel ◽  
Fracture Toughness Test ◽  
Irradiation Embrittlement ◽  
Charpy Test ◽  
Vessel Steel ◽  
Rupture Energy ◽  
Reference Index

Nuclear pressure vessel steels are subjected to irradiation embrittlement which is monitored using Charpy tests. Reference index temperatures, such as the temperature for which the mean Charpy rupture energy is equal to 56 J (T56J), are used as embrittlement indicators. The safety integrity evaluation is performed assuming that the shift of RTNDT due to irradiation is equal to the shift of T56J. In this work a material model integrating a description of viscoplasticity, ductile damage and brittle fracture is used to simulate both the Charpy test and the fracture toughness test (CT geometry). The model is adjusted on an unirradiated material. It is then applied to irradiated materials assuming that irradiation affects hardening. It is shown that irradiation probably also affects brittle failure. The shift of RTNDT and the predicted shift of T100MPam are then compared for a given level of irradiation.

Specimen Size Effect on Fracture Toughness of Reactor Pressure Vessel Steel Following Warm Pre-Stressing

2016 ◽  
Author(s):  
Keiko Iwata ◽  
Tohru Tobita ◽  
Hisashi Takamizawa ◽  
Yasuhiro Chimi ◽  
Kentaro Yoshimoto ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Size Effect ◽  
Pressure Vessel ◽  
Specimen Size ◽  
Reactor Pressure Vessel ◽  
Transient Condition ◽  
Pressure Vessel Steel ◽  
Vessel Steel ◽  
Reactor Pressure Vessel Steel ◽  
Reactor Pressure

The effect of warm pre-stressing (WPS) on fracture toughness was evaluated for a reactor pressure vessel steel. Various types of thermomechanical loadings were applied to 1T-CT specimens. The results were compared with predictions from several analytical WPS engineering models. The specimen size effect was subsequently investigated under the load-unload-cool-fracture transient condition using 1T-, 0.4T-, and 0.16T-CT specimens. Analyses of the plastic zone distribution and residual stress were conducted to identify the difference in the WPS effect among the specimens.

Fracture Toughness Characterization of Midland Beltline Weld After High Dose of Irradiation

2006 ◽  
Author(s):  
Mikhail A. Sokolov ◽  
Randy K. Nanstad
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Neutron Fluence ◽  
National Laboratory ◽  
Reactor Pressure Vessel ◽  
High Dose ◽  
Pressure Vessel Steel ◽  
Vessel Steel ◽  
Oak Ridge ◽  
Reactor Pressure

The Heavy-Section Steel Irradiation (HSSI) Program at Oak Ridge National Laboratory includes a task to investigate the shape of the fracture toughness master curve for reactor pressure vessel steel highly embrittled as a consequence of irradiation exposure, and to examine the ability of the Charpy 41-J shift to predict the fracture toughness shift. As part of this task, a low upper-shelf WF-70 weld obtained from the beltline region of the Midland Unit 1 reactor pressure vessel was characterized in terms of static initiation and Charpy impact toughness in the unirradiated and irradiated conditions. Irradiation of this weld was performed at the University of Michigan Ford Reactor at 288°C to neutron fluence of 3.4×1019 neutron/cm2 in the HSSI irradiation-anneal-reirradiation facility. This reusable facility allowed the irradiation of either virgin or previously irradiated material in a well-controlled temperature regime, including the ability to perform in-situ annealing. This was the last capsule irradiated in this facility before reactor shut down. Thus, the Midland beltline weld was irradiated within the HSSI Program to three fluences — 0.5×1019; 1.0×1019; and 3.4×1019 neutron/cm2. It was anticipated that it would provide an opportunity to address fracture toughness curve shape and Charpy 41-J shift compatibility issues at different levels of embrittlement, including the highest dose considered to be in the range of the current end of life fluence. It was found that the Charpy 41-J shift practically saturated after neutron fluence of 1.0×1019 neutron/cm2. The transition fracture toughness shift after 3.4×1019 neutron/cm2 was only slightly higher than that after 1.0×1019 neutron/cm2. In all cases, transition fracture toughness shifts were lower than predicted by the Regulatory Guide 1.99, Rev. 2 equation.

Assessment of Pressure Vessel Steel Irradiation Embrittlement Up to 40 Years Using Local Approach to Fracture Modelling: Application to the French Surveillance Program

2011 ◽  
Author(s):  
B. Tanguy ◽  
A. Parrot ◽  
F. Cle´mendot ◽  
G. Chas
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Chemical Elements ◽  
Experimental Results ◽  
Physical Models ◽  
Surveillance Program ◽  
Pressure Vessel Steel ◽  
Irradiation Embrittlement ◽  
Vessel Steel ◽  
Different Levels

For western pressure vessel reactors, assessment of pressure vessel steels irradiation embrittlement due to neutron irradiation is based on a semi-empirical formulae which predicts the shift of a reference lower bound fracture toughness curve as a function of fluence and embrittlement-involved chemical elements. Periodically, in order to monitor the embrittlement of each RPV, the predictions of the formulae is confronted to experimental results obtained from Charpy specimens located in surveillance capsules irradiated with a higher fluence level than the pressure vessel itself. Historically only the shift of the temperature index defined for a given level of energy, e.g. 56J in the French surveillance program, is used. In support to the French surveillance program methodology, for some of the French RPVs, physical models of fracture (for both cleavage and ductile fracture) are used to analyse in details the whole experimental basis available at different levels of fluence. This study presents the methodology developed in order to analyse the experimental results of a RPV steel from the french surveillance program, including Charpy and fracture toughness tests at different levels of fluence i.e. of embrittlement. The methodology applied aims to use the numerous Charpy tests results available in order to assess, at the same fluence levels, the fracture toughness embrittlement. The results are then compared to available fracture toughness results for a given level of embrittlement.

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