Crack Arrest Test Results of Unirradiated and Irradiated German RPV Steels

2013 ◽  
Author(s):  
Florian Obermeier ◽  
Julia Barthelmes ◽  
Elisabeth Keim ◽  
Hieronymus Hein ◽  
Hilmar Schnabel ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Lower Bound ◽  
Pressure Vessel ◽  
Charpy Impact ◽  
Crack Arrest ◽  
Reactor Pressure Vessel ◽  
Test Method ◽  
Fracture Toughness Test ◽  
Test Results ◽  
Reactor Pressure

In the CARISMA[1] and CARINA[2] projects comprehensive tensile, Charpy-impact and fracture toughness tests were performed for unirradiated and irradiated original reactor pressure vessel (RPV) steel specimens from German pressurized water reactors (PWR) up to neutron fluences in the range of 60 operational years and beyond. In addition, crack arrest fracture toughness tests were performed to demonstrate the crack arrest behavior of the materials. To determine the crack arrest properties of ferritic steels, the designated test method according to ASTM E1221 [3] was used. However, in particular for irradiated reactor pressure vessel materials with higher irradiation embrittlement, the prescribed standard test specimen does not always provide adequate test results. During starter notch preparation annealing effects occurred in the heat affected zone (HAZ) of the brittle weld of the starter notch causing crack arrest in the HAZ after unstable crack initiation. Therefore a modified test method to perform crack arrest tests with so called duplex specimens was investigated. In this paper this modified method and the test results of five base and four weld metals with a fluence up to 4,69E+19 cm−2 (E >1 MeV) are discussed. The available test results show that the duplex specimen is an appropriate alternative to the standard compact crack arrest (CCA) specimen. The measured KIa fracture toughness data are enveloped by the “lower bound” of the ASME KIa-curve indexed with RTNDTj or TKIa but not all data are enveloped by indexing the “lower bound” curve with RTT0 like described in the ASME Code Case N-629 [4]. Furthermore correlations of the crack arrest test results with Charpy-impact and fracture toughness test results will be shown.

Evaluation of Brittle Crack Arrest Toughness for Highly-Irradiated Reactor Pressure Vessel Steels

2021 ◽  
Author(s):  
Keiko Iwata ◽  
Kuniki Hata ◽  
Tohru Tobita ◽  
Takatoshi Hirota ◽  
Hisashi Takamizawa ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Impact Test ◽  
Charpy Impact ◽  
Crack Arrest ◽  
Charpy Impact Test ◽  
Reactor Pressure Vessel ◽  
Reference Temperature ◽  
Arrest Toughness ◽  
Reactor Pressure

Abstract The crack arrest fracture toughness, KIa, values for highly-irradiated reactor pressure vessel (RPV) steels are estimated according to the linear relationship between crack arrest toughness reference temperature, TKIa, and the temperature corresponding to a fixed arrest load, equal to 4 kN, TFa4kN, obtained by instrumented Charpy impact test. The relationship between TKIa derived from the instrumented Charpy impact test and fracture toughness reference temperature, To, was expressed as an equation proposed in a previous study. The coefficients in the equation could be fine-tuned to obtain a better fitting curve using the present experimental data and previous KIa data. The KIa curve for RPV;A533B class 1 steels irradiated up to 1.3 × 1020 n/cm2 (E > 1 MeV) was compared with a KIR curve defined in JEAC4206-2016. The KIR curve was always lower than the 1%ile curve of KIa for these irradiated RPV steels. This result indicates that the conservatism of the method defined in JEAC4206-2016 to evaluate KIa using the KIR curve is confirmed for highly-irradiated RPV steels.

Effect of Plastic Constraint and Cladding on Semi-Elliptical Shaped Crack in Fracture Toughness Evaluation for a Reactor Pressure Vessel Steel

2021 ◽  
Author(s):  
Masaki Shimodaira ◽  
Tohru Tobita ◽  
Yasuto Nagoshi ◽  
Kai Lu ◽  
Jinya Katsuyama
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Crack Tip ◽  
Reactor Pressure Vessel ◽  
Pressure Vessel Steel ◽  
Fracture Toughness Test ◽  
Integrity Assessment ◽  
Plastic Constraint ◽  
Shaped Crack ◽  
Reactor Pressure

Abstract In the structural integrity assessment of a reactor pressure vessel (RPV), the fracture toughness (KJc) should be higher than the stress intensity factor at the crack tip of a semi-elliptical shaped under-clad crack (UCC), which is prescribed in JEAC4206-2016. However, differences in crack depth and existence of cladding between the postulated crack and fracture toughness test specimens would be affected to the plastic constraint state and KJc evaluation. In this study, we performed fracture toughness tests and finite element analyses to investigate the effect of plastic constraint and cladding on the semi-elliptical shaped crack in KJc evaluation. The apparent KJc value evaluated at the deepest point of the crack exceeded 5% fracture probability based on the Master Curve method estimated from C(T) specimens, and the conservativeness of the current integrity assessment method was confirmed. Few initiation sites were observed along the tip of semi-elliptical shaped crack other than the deepest point. The plastic constraint state was also analyzed along the crack tip, and it was found that the plastic constraint at the crack tip near the surface was lower than that for the deepest point. Moreover, it was quantitatively showed that the UCC decreased the plastic constraint. The local approach suggested higher KJc value for the UCC than that for the surface crack, reflecting the low constraint effect for the UCC.

Alternative Reference Temperature Based on Master Curve Approach in Japanese Reactor Pressure Vessel Steel

2013 ◽  
Author(s):  
Takatoshi Hirota ◽  
Takashi Hirano ◽  
Kunio Onizawa
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Master Curve ◽  
Reactor Pressure Vessel ◽  
Test Method ◽  
Reference Temperature ◽  
Ferritic Steels ◽  
Pressure Vessel Steel ◽  
Reactor Pressure

Master Curve approach is the effective method to evaluate the fracture toughness of the ferritic steels accurately and statistically. The Japan Electric Association Code JEAC 4216-2011, “Test Method for Determination of Reference Temperature, To, of Ferritic Steels” was published based on the related standard ASTM E 1921-08 and the results of the investigation of the applicability of the Master Curve approach to Japanese reactor pressure vessel (RPV) steels. The reference temperature, To can be determined in accordance with this code in Japan. In this study, using the existing fracture toughness data of Japanese RPV steels including base metals and weld metals, the method for determination of the alternative reference temperature RTTo based on Master Curve reference temperature To was statistically examined, so that RTTo has an equivalent safety margin to the conventional RTNDT. Through the statistical treatment, the alternative reference temperature RTTo was proposed as the following equation; RTTo = To + CMC + 2σTo. This method is applicable to the Japan Electric Association Code JEAC 4206, “Method of Verification Tests of the Fracture Toughness for Nuclear Power Plant Components” as an option item.

Reconstitution of V-Notched Charpy Impact Specimen Made of Short Insert

2006 ◽  
Author(s):  
Toru Osaki ◽  
Hiroshi Matsuzawa
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Charpy Impact ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Absorption Energy ◽  
Impact Specimen ◽  
Insert Length ◽  
Reactor Pressure

Reconstitution in this paper means to constitute the original size V-notched Charpy impact specimen, which is made of the irradiated insert cut out from broken piece and un-irradiated tabs welded to the insert. It is a promising technique to secure an adequate number of surveillance specimens for long-term operation of nuclear power plants. Every Japanese nuclear power plant has its own surveillance test program, and is operated considering its unique surveillance test results along with the general reduction tendency of fracture toughness. This practice should be continued and enhanced if possible, after the full use of originally installed specimens, because its fracture toughness is lower than before. Reconstitution of V-notched Charpy impact specimens to the original shape by using a short insert was studied. Charpy absorption energy is generally shifted by reconstitution, if the insert length is as short as 10 mm. Reconstitution with a short insert is necessary when the transverse property of the original specimen is required although only the longitudinal surveillance specimen is installed as in some early constructed reactor pressure vessels in Japan. This case is important when the reactor pressure vessel is suspected to be a so-called low upper shelf toughness reactor pressure vessel. The minimum required insert length to avoid affect on the specimen properties depends on the Charpy absorption energy of the insert and reconstitution weld condition. Correlation between Charpy absorption energy and plastic deformation size, and short time annealing properties of irradiated pressure vessel steels were investigated. A method to evaluate the minimum required insert length was proposed, which depends on the expected Charpy absorption energy and thermal transient during reconstitution. It was demonstrated that the reconstituted specimens of 10 mm-long irradiated inserts, whose upper shelf absorption energy was 69J and required insert length was 9.5mm, showed little shift of upper shelf absorption energy.

An Investigation of Cladding Effects on Shallow-Flaw Fracture Toughness of Reactor Pressure Vessel Steel Under Prototypic Biaxial Loading

1999 ◽  
Vol 121 (3) ◽  
pp. 257-268 ◽  
Author(s):  
B. R. Bass ◽  
W. J. McAfee ◽  
J. W. Bryson ◽  
W. E. Pennell
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Biaxial Loading ◽  
Reactor Pressure Vessel ◽  
Biaxial Stress ◽  
Pressure Vessel Steel ◽  
Test Results ◽  
Vessel Steel ◽  
Nonlinear Stress ◽  
Reactor Pressure

Potential structural-integrity benefits or liabilities of the stainless steel cladding on the inner surface of a reactor pressure vessel (RPV) are important considerations in the effort to refine or improve safety assessment procedures applied to RPVs. Clad-beam tests were carried out to investigate and quantify effects of the clad structure on fracture initiation toughness of through-clad shallow surface flaws in RPV material. A cruciform beam specimen was developed at ORNL to introduce a prototypic, far-field, out-of-plane biaxial stress component that provides a linear approximation of the nonlinear stress distribution generated by thermo-mechanical loading transients in an RPV. The cruciform specimens (102-mm-thick test section) were fabricated from RPV shell segments available from a canceled pressurized-water reactor plant. The specimens were tested under biaxial load ratios ranging from 0.0 (uniaxial) to 1.0 (full biaxial), the ratio being defined as the total load applied to the transverse beam arms divided by that applied to the longitudinal arms. The test results imply that biaxial loading is effective in reducing the shallow-flaw fracture toughness of the clad/heat-affected zone/structural-weld region of the RPV shell below that determined from uniaxial loading conditions. The lowest toughness value from the clad cruciform specimens tested under biaxial loading is only slightly above the ASME Section XI KIc curve. For all biaxiality ratios, the test results imply that shallow-flaw fracture toughness data from the RPV structural weld material are significantly lower than that obtained from a high-yield strength plate material.

Sign in / Sign up

Export Citation Format

Share Document