The embrittlement of reactor pressure vessel (RPV) steels due to neutron irradiation restricts the operating lifetime of nuclear reactors. The reference temperature 𝑇0, obtained from fracture mechanics testing using the Master Curve concept, is a good indicator of the irradiation resistance of a material. The measurement of the shift in 𝑇0 after neutron irradiation, which accompanies the embrittlement of the material, using the Master Curve concept, enables the assessment of the reactor materials. In the context of worldwide life time extensions of nuclear power plants, the limited availability of neutron irradiated materials (surveillance materials) is a challenge. Testing of miniaturized 0.16T C(T) specimens manufactured from already tested standard Charpy-sized specimens helps to solve the material shortage problem. In this work, four different reactor pressure vessel steels with different compositions were investigated in the unirradiated and in the neutron-irradiated condition. A total number of 189 mini-C(T) samples were fabricated and tested. An important component of this study is the transferability of fracture mechanics data from mini-C(T) to standard Charpy-sized specimen. Our results demonstrate good agreement of the reference temperatures from the mini-C(T) specimens with those from standard Charpy-sized specimens. RPV steels containing higher Cu and P contents exhibit a higher increase in 𝑇0 after irradiation. The fracture surfaces were investigated using SEM in order to record the location of the fracture initiators. The fracture modes were also determined. A large number of test results formed the basis for a censoring probability function, which was used to optimally select the testing temperature in Master Curve testing. The effect of the slow stable crack growth censoring criteria from ASTM E1921 on the determination of 𝑇0 was analysed and found to have a minor effect. Our results demonstrate the validity of mini-C(T) specimen testing and confirm the role of the impurity elements Cu and P in neutron embrittlement. We anticipate further research linking microstructure to the fracture properties of materials before and after neutron irradiation and the optimization of Master Curve testing using the results from our statistical analysis.

Radiation damage occurs in reactor pressure vessel (RPV) steel, causing microstructural changes such as point defect clusters, interstitial loops, vacancy-solute clusters, and precipitates, that cause material embrittlement. Radiation damage is a crucial concern in the nuclear industry since many nuclear plants throughout the US are entering the first period of life extension and older plants are currently undergoing assessment of technical basis to operate beyond 60 years. The result of extended operation is that the RPV and other components will be exposed to higher levels of neutron radiation than they were originally designed to withstand. There is currently no nondestructive evaluation technique that can unambiguously assess the amount of radiation damage in RPV steels. Nonlinear ultrasound (NLU) is a nondestructive evaluation technique that is sensitive to microstructural features such as dislocations, precipitates, and their interactions in metallic materials. The physical effect monitored by NLU is the generation of higher harmonic frequencies in an initially monochromatic ultrasonic wave, arising from the interaction of the ultrasonic wave with microstructural features. This effect is quantified with the measurable acoustic nonlinearity parameter, beta. In this work, nonlinear ultrasound is used to characterize radiation damage in reactor pressure vessel steels over a range of fluence levels, irradiation temperatures, and material composition. Experimental results are presented and interpreted with newly developed analytical models that combine different irradiation-induced microstructural contributions to the acoustic nonlinearity parameter.

The paper presents results of a research project founded by the Swiss Federal Nuclear Inspectorate concerning the application of the Master Curve approach in nuclear reactor pressure vessels integrity assessment. The main focus is put on the applicability of pre-cracked 0.4T-SE(B) specimens with short cracks, the verification of transferability of MC reference temperatures T0 from 0.4T thick specimens to larger specimens, ascertaining the influence of the specimen type and the test temperature on T0, investigation of the applicability of specimens with electroerosive notches for the fracture toughness testing, and the quantification of the loading rate and specimen type on T0. The test material is a forged ring of steel 22 NiMoCr 3 7 of the uncommissioned German pressurized water reactor Biblis C. SE(B) specimens with different overall sizes (specimen thickness B=0.4T, 0.8T, 1.6T, 3T, fatigue pre-cracked to a/W=0.5 and 20% side-grooved) have comparable T0. T0 varies within the 1σ scatter band. The testing of C(T) specimens results in higher T0 compared to SE(B) specimens. It can be stated that except for the lowest test temperature allowed by ASTM E1921-09a, the T0 values evaluated with specimens tested at different test temperatures are consistent. The testing in the temperature range of T0 ± 20 K is recommended because it gave the highest accuracy. Specimens with a/W=0.3 and a/W=0.5 crack length ratios yield comparable T0. The T0 of EDM notched specimens lie 41 K up to 54 K below the T0 of fatigue pre-cracked specimens. A significant influence of the loading rate on the MC T0 was observed. The HSK AN 425 test procedure is a suitable method to evaluate dynamic MC tests. The reference temperature T0 is eligible to define a reference temperature RTTo for the ASME-KIC reference curve as recommended in the ASME Code Case N-629. An additional margin has to be defined for the specific type of transient to be considered in the RPV integrity assessment. This margin also takes into account the level of available information of the RPV to be assessed.

The paper presents results of a research project founded by the Swiss Federal Nuclear Inspectorate concerning the application of the Master Curve approach in nuclear reactor pressure vessels integrity assessment. The main focus is put on the applicability of pre-cracked 0.4T-SE(B) specimens with short cracks, the verification of transferability of MC reference temperatures T0 from 0.4T thick specimens to larger specimens, ascertaining the influence of the specimen type and the test temperature on T0, investigation of the applicability of specimens with electroerosive notches for the fracture toughness testing, and the quantification of the loading rate and specimen type on T0. The test material is a forged ring of steel 22 NiMoCr 3 7 of the uncommissioned German pressurized water reactor Biblis C. SE(B) specimens with different overall sizes (specimen thickness B=0.4T, 0.8T, 1.6T, 3T, fatigue pre-cracked to a/W=0.5 and 20% side-grooved) have comparable T0. T0 varies within the 1σ scatter band. The testing of C(T) specimens results in higher T0 compared to SE(B) specimens. It can be stated that except for the lowest test temperature allowed by ASTM E1921-09a, the T0 values evaluated with specimens tested at different test temperatures are consistent. The testing in the temperature range of T0 ± 20 K is recommended because it gave the highest accuracy. Specimens with a/W=0.3 and a/W=0.5 crack length ratios yield comparable T0. The T0 of EDM notched specimens lie 41 K up to 54 K below the T0 of fatigue pre-cracked specimens. A significant influence of the loading rate on the MC T0 was observed. The HSK AN 425 test procedure is a suitable method to evaluate dynamic MC tests. The reference temperature T0 is eligible to define a reference temperature RTTo for the ASME-KIC reference curve as recommended in the ASME Code Case N-629. An additional margin has to be defined for the specific type of transient to be considered in the RPV integrity assessment. This margin also takes into account the level of available information of the RPV to be assessed.