No description available.

The micromechanisms by which ductile fracture extended from a pre-existing crack was experimentally observed for two classes of forged SA 508 pressure vessel steel. The micromechanisms were related to the measured values of fracture toughness characterised by the resistance to crack initiation and growth. This comparison was only possible with the aid of an accurate fracture resistance test technique which could determine the crack growth toughness from a single specimen. An unloading compliance test system was developed and was used for the construction of crack growth resistance curves. Microstructural parameters determined from a specimen were related to the toughness measured on that specimen and this proved invaluable in isolating the controlling parameters. The effect of orientation and location on the toughness of the materials was assessed. The crack growth resistance was sensitive to the orientation of the crack with respect to the maximum hot working direction and the bands of segregation associated with elongated manganese sulphide inclusions. The toughness was high when the crack plane was perpendicular to the segregation bands and low when the crack plane was parallel with the bands. The location of the crack-tip through the thickness of the forging had a minor effect on the crack growth resistance. A limited study of test temperature, strength level and isothermal ageing was undertaken. Testing within the dynamic strain ageing regime of temperature had a marked effect and reduced the crack growth resistance to below the value at room temperature. Increasing the strength level of one steel by re-heat treating had no effect on the crack growth resistance. Subsequent isothermal ageing treatments also had no effect on the resistance curves. The magnitude and extent of void formation around growing cracks was studied and related to the applied loading. The size, shape and distribution of inclusions was characterised for the materials and orientations used in the fracture tests. Correlations between inclusion parameters and toughness revealed the important microstructural parameters controlling initiation and crack growth. Simple models for initiation and crack growth resistance were developed which take the controlling parameters into account. These models are shown to agree reasonably well with some experimental data.

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.

No description available.

No description available.

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.

Ziel des BMWi-Fördervorhabens 1501393 ist es, durch den Einsatz von Untersuchungsmethoden auf der nm-Skala einen Beitrag zur Aufklärung von Flusseffekten und von Late-Blooming-Effekten in bestrahlten RDB-Stählen zu leisten. Zur Untersuchung dieser Effekte wurde auf RDB-Stähle deutscher Reaktoren aus zwei bei der AREVA GmbH abgeschlossenen Vorhaben zurückgegriffen. Die Auswahl der Grundwerkstoffe und Schweißgüter erfolgte so, dass sich optimale Voraussetzungen für das Erreichen des Gesamtziels des Vorhabens ergeben. Die ausgewählten Untersuchungsmethoden umfassen mit der Neutronenkleinwinkelstreuung, der Atomsondentomographie und der Positronen-annihilationsspektroskopie solche Techniken, die die nm-skaligen bestrahlungsinduzierten Defekt-Fremdatom-Cluster bestmöglich und in komplementärer Weise zu detektieren und zu charakterisieren gestatten. Es wurde ein Flusseffekt auf die Größe der bestrahlungsinduzierten Fremdatomcluster, jedoch nicht auf den Volumenanteil und die mechanischen Eigenschaften gefunden. In einem Cu-armen RDB-Schweißgut wurde ein Late-Blooming-Effekt nachgewiesen, der sich in einem steilen Anstieg des Clustervolumenanteils und der Übergangstemperaturverschiebung nach einer Phase schwacher oder fehlender Zunahme niederschlägt. The BMWi project 1501393 aimed at contributing to the clarification of flux effects and late blooming effects in irradiated RPV steels by means of experimental techniques of sensitivity at the nm scale. The investigation of these effects was focussed on RPV steels, both base metal and weld of German reactors selected according to the objectives of the present project from two previous projects performed at AREVA GmbH. The complementary techniques of small-angle neutron scattering, atom probe tomography and positron annihilation spectroscopy were applied to detect and characterize the irradiation-induced nm-scale defect-solute clusters. A flux effect on the size of the irradiation-induced clusters but no flux effect on both cluster volume fraction and mechanical properties was found. For a low-Cu RPV weld, a late blooming effect was observed, which results in a steep slope of both cluster volume fraction and transition temperature shift after an initial stage of small or no change.

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.

Ziel des BMWi-Fördervorhabens 1501393 ist es, durch den Einsatz von Untersuchungsmethoden auf der nm-Skala einen Beitrag zur Aufklärung von Flusseffekten und von Late-Blooming-Effekten in bestrahlten RDB-Stählen zu leisten. Zur Untersuchung dieser Effekte wurde auf RDB-Stähle deutscher Reaktoren aus zwei bei der AREVA GmbH abgeschlossenen Vorhaben zurückgegriffen. Die Auswahl der Grundwerkstoffe und Schweißgüter erfolgte so, dass sich optimale Voraussetzungen für das Erreichen des Gesamtziels des Vorhabens ergeben. Die ausgewählten Untersuchungsmethoden umfassen mit der Neutronenkleinwinkelstreuung, der Atomsondentomographie und der Positronen-annihilationsspektroskopie solche Techniken, die die nm-skaligen bestrahlungsinduzierten Defekt-Fremdatom-Cluster bestmöglich und in komplementärer Weise zu detektieren und zu charakterisieren gestatten. Es wurde ein Flusseffekt auf die Größe der bestrahlungsinduzierten Fremdatomcluster, jedoch nicht auf den Volumenanteil und die mechanischen Eigenschaften gefunden. In einem Cu-armen RDB-Schweißgut wurde ein Late-Blooming-Effekt nachgewiesen, der sich in einem steilen Anstieg des Clustervolumenanteils und der Übergangstemperaturverschiebung nach einer Phase schwacher oder fehlender Zunahme niederschlägt. The BMWi project 1501393 aimed at contributing to the clarification of flux effects and late blooming effects in irradiated RPV steels by means of experimental techniques of sensitivity at the nm scale. The investigation of these effects was focussed on RPV steels, both base metal and weld of German reactors selected according to the objectives of the present project from two previous projects performed at AREVA GmbH. The complementary techniques of small-angle neutron scattering, atom probe tomography and positron annihilation spectroscopy were applied to detect and characterize the irradiation-induced nm-scale defect-solute clusters. A flux effect on the size of the irradiation-induced clusters but no flux effect on both cluster volume fraction and mechanical properties was found. For a low-Cu RPV weld, a late blooming effect was observed, which results in a steep slope of both cluster volume fraction and transition temperature shift after an initial stage of small or no change.