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The multi-layer beltline welding seam of the Biblis C reactor pressure vessel was characterized by hardness, tensile, ISO-V impact and fracture toughness testing. The reference temperature, T0, was determined according to the test standard ASTM E1921 at different thickness positions of the multi-layer welding seam. Additionally, the influence of the specimen orientation on the ISO-V ductile-to-brittle transition temperature and T0 was investigated. In contrast to the T-S orientation (crack extension through the thickness) the crack front of the T-L oriented specimens (crack extension in welding direction) penetrates several welding beads. By means of fractographic and metallographic analyses of the fractured surface of fracture mechanics SE(B) specimens was shown that the distribution of the crack initiation sites is not necessarily correlated to the structure of the different welding beads along the crack front. Furthermore, it was found that the scatter of the fracture toughness values at cleavage failure, KJc, determined with T-S specimens is significantly higher than in case of the T-L specimens. T0 values measured at different thickness locations of the multi-layer welding seam vary in a range of about 40 K. The evaluated T0 values are used to determine the reference temperature RTTo for indexing the lower bound curve KIc(T) according to the Regulatory Guide ENSI-B01 for the ageing surveillance of nuclear power plants in Switzerland. It could be shown that the KIc values converted from the KJc values are enveloped by the lower bound curves.

The fracture behaviour of a component or specimen that contains a sharp notch is governed essentially by the same theoretical relations known from cracks. The blunt notch root only causes an increase of the resistance against crack initiation, which depends on the fracture mechanism. In the present paper, the relation between fracture toughness and notch toughness is investigated by simple analytical models. The derived formulas were compared with experimental results obtained from fracture toughness tests on RPV-steel 24 NiCrMo 3-7 at various temperatures. 1T-CT- and 0.4T-SEB-specimens that contained a sharp notch with a root radius of 0.06 mm introduced by spark erosion (EDM) instead of the standard fatigue crack were used. The predictions were found to agree well with the experimental data. The effect of the notch radius on fracture toughness is most pronounced in the brittle to ductile transition regime, where fracture toughness can be characterized by the master curve and the corresponding reference temperature T0 according to ASTM E1921. Accordingly, the effect of the notch radius can be quantified by a shift of T0. Since the shape of the transition curve depends on the notch radius, the procedure of ASTM E1921 to determine T0 is not applicable. An alternative is suggested. As limiting cases, ductile tearing and brittle fracture are also considered.

LONGLIFE (“Treatment of long term irradiation embrittlement effects in RPV safety assessment”) was a collaborative project of the 7th Framework Programme of EURATOM under the umbrella of NULIFE/NUGENIA, aiming at an improved understanding of irradiation effects in reactor pressure vessel steels under conditions representative of long term operation. The LONGLIFE project was completed by the end of January 2014. The paper gives an overview of the main project results and their implications for future research, as discussed at the final project workshop. The microstructural database for neutron-irradiated RPV steels was extended considerably and existing gaps on mechanical property data were closed. Indications of late blooming effects (LBE) were found in some cases, but clear criteria for the occurrence/exclusion in terms of irradiation conditions and chemical composition have still to be developed. The commonly accepted trend, that low flux and low irradiation temperature promotes LBE, is supported. A significant flux effect on the size of defect clusters was observed in two high Cu weld materials, while the changes of mechanical properties are not affected by the neutron flux. The database requires completion in particular for low-Cu RPV steels. The shift of reference temperature T 0 over the thickness location of a VVER-440 welding seam does not follow the prediction Russian code, because of the strong variation of the intrinsic weld bead structure. Therefore, the effect of the initial microstructure and of the heterogeneity on the radiation behaviour has to be addressed in future works. Existing embrittlement trend curves models were applied to the LONGLIFE data base. None of the trend curves could predict the behaviour of the entirety of the LONGLIFE materials sufficiently. A guideline for monitoring radiation embrittlement during life extension periods was developed.

The paper presents data measured for trepans sampled from the decommissioned WWER-440 reactor pressure vessel of the NPP Greifswald Unit 4. The main focus of this work is on fracture toughness characterisation according to test standard ASTM E1921. Large variations in the evaluated reference temperature values, T0, across the wall of the multilayer beltline welding seam were observed. Generally, the ductile-to-brittle transition temperature shift predicted by the Russian code for the present content of deleterious elements P and Cu and the accumulated neutron fluences lies within the amount of the scatter of the measured T0 values. Metallographic investigations show that the T0 values measured with T–S oriented Charpy size SE(B) specimens from different thickness locations of the multilayer welding seams strongly depend on the microstructure at the specimen crack tip, and, consequently, on the initial structure of the multilayer welding seam. The RPV integrity is discussed, taking into account a pressurised thermal shock scenario.

The real toughness response of RPV material can only be determined after the final shut down of the NPP. Such a chance is given now by investigating material from the former Greifswald NPP (VVER-440/230). In the first part the paper deals with fast neutron fluence calculations and retrospective dosimetry based on Niobium. Unfortunately, a second neutron reaction besides 93 Nb( n , n ’) leading to 93m Nb-activity is the reaction 92 Mo( n , γ ) 93 Mo. Based on the found Nb and Mo contents in the RPV material, it turned out that the 93m Nb generation on the Mo path mostly dominates over the fast neutron induced generation from Nb. The comparison between the calculated and the measured 93m Nb activities typically resulted in deviations of 50%. Possible reasons for the observed differences are discussed. In the second part first results of fracture mechanic investigations are reported. SE(B) specimens from three thickness positions were tested and evaluated according to the test standard ASTM E1921-05. Cleavage fracture toughness values, K Jc , were determined and Master Curve based reference temperatures ( T 0 ) were evaluated. The T 0 measured at the inner surface of the RPV did not represent the conservative condition. The T 0 of disc 1-1.3 located between the surface and 1/4 thickness is about 40K higher compared with those of the surface. The measured K Jc values are not enveloped by the 5% fractile indexed with T 0 according to the Master Curve concept. However, the 5% fractile indexed with the VERLIFE reference temperature RTT o that includes an additional margin envelops the measured K Jc values. Therefore the VERLIFE lower bound curve conservatively describes the fracture toughness of the investigated weld metal.

While the Master Curve (MC) method is gradually entering brittle fracture safety assessment procedures world-wide, knowledge is still lacking about its limits of applicability to highly neutron irradiated material. In this paper two reactor pressure vessel (RPV) steels A533B Cl.1 (IAEA reference material code JRQ) and A508 Cl.3 (code JFL) were scrutinized for possible deviations of the postulated invariant MC shape and the MC validity for macroscopically inhomogeneous microstructure. Besides tensile and Charpy-V tests, MC tests were performed on Charpy-size three-point bend specimens in the unirradiated, neutron irradiated with fluences up to nearly 10 20 n/cm 2 ( E > 1 MeV) and recovery heat treated condition. Evaluation procedures include Master Curve reference temperature T 0 determination according to ASTM E1921-05 as well as additional analysis methods such as SINTAP, multi-modal MC method (MML) and the Unified Curve (UC). Integrity assessment according to ASME Code Cases N-629 and N-631 has been applied. It is shown that the standard MC concept provides a precise description of the fracture toughness for all conditions, even exceptionally well for the highly irradiated state. No MC shape change could be observed, whereas the UC concept indicates a significant influence of irradiation on the fracture toughness curves for the highly irradiated JRQ.

The master curve (MC) approach as standardised in the ASTM Standard Test Method E1921 was applied to weld metal of the reactor pressure vessel (RPV) beltline welding seam of Greifswald unit 8 RPV. Charpy size SE(B) specimens from 13 locations equally spaced over the thickness of the welding seam were tested. The orientation of the specimens within the welding seam is TL and TS according to ASTM E399. The fracture toughness values measured on the SE(B) specimens with both orientations follow the course of the master curve. Nearly all values lie within the fracture toughness curves for 2% and 98% fracture probability. There is a strong variation of the reference temperature T 0 through the thickness of the welding seam, which can be explained by microstructural differences. The scatter is more pronounced for the TS SE(B) specimens. It can be shown that specimens with TL and TS orientation in the welding seam have a differentiating and integrating behaviour, respectively.

The Master Curve (MC) method is a promising technique for evaluating the fracture toughness of ferritic steels. It enables the determination of the reference temperature, To, of a probabilistic fracture toughness curve using small specimens. The Central Research Institute of Electric Power Industry (CRIEPI) investigated this capability of the MC method using different size of C(T) specimens, and it was found that 0.16T-C(T) specimen with the dimensions of 4 by 10 by 9.6 mm, hereafter called “Mini-CT specimen,” can be used to obtain valid To values. The advantage of this Mini-CT specimen technique is that multiple specimens can be machined from one of the broken halves of Charpy size specimens, which are used in a standard surveillance capsule of a reactor pressure vessel (RPV). In order to ensure the robustness of this technique, a round-robin test was planned. The idea is to perform MC tests using Mini-CT specimens by different investigators to see if consistent To values can be obtained. All the specimens used were machined and pre-cracked by one fabricator from unique RPV material to avoid any possible effect of specimen preparation on To values. Seven institutes participated in this exercise, and obtained valid To values. No specific difficulty was found in the MC tests performed in accordance with the ASTM E1921-10el protocol. The scatter of the obtained To values was well within the uncertainty range defined in Appendix X4.2 of ASTM E1921, indicating the robustness of the Mini-CT specimen technique in terms of the testing procedure. Throughout this activity, we could obtain 182 KJc(1Teq) for a single material. We investigated the statistics of this large database, and found that there is no remarkable difference not only in the To values, but also in the fracture toughness distribution between the Mini-CT specimen and the standard 1T-C(T) specimen results.

The master curve (MC) approach used to measure the transition temperature, T 0 , was standarised in the ASTM Standard Test Method E 1921 in 1997. The basic MC approach for analysis of fracture test results is intended for macroscopically homogeneous steels with a body centred cubic (ferritic) structure only. In reality, due to the manufacturing process, the steels in question are seldom fully macroscopically homogeneous. The fracture toughness values measured on Charpy size SE(B) specimens of base metal from the Greifswald Unit 8 rector pressure vessel (RPV) show large scatter. The basic MC evaluation following ASTM E1921 supplies a MC with many fracture toughness values which lie below the 5% fracture probability line. It is therefore suspected that this material is macroscopically inhomogeneous. In this paper, two recent extensions of the MC for inhomogeneous materials are applied to these fracture toughness data.

The investigation of reactor pressure vessel (RPV) materials from decommissioned NPPs offers the unique opportunity to scrutinize the irradiation behaviour under real conditions. Material samples taken from the RPV wall enable a comprehensive material characterisation. The paper describes the investigation of trepans taken from the decommissioned WWER-440 first generation RPVs of the Greifswald NPP. Those RPVs represent different material conditions such as irradiated (I), irradiated and recovery annealed (IA) and irradiated, recovery annealed and re-irradiated (IAI). The working program is focussed on the characterisation of the RPV steels (base and weld metal) through the RPV wall. The key part of the testing is aimed at the determination of the reference temperature T0 following the ASTM Test Standard E1921-05 to determine the fracture toughness of the RPV steel in different thickness locations. In a first step the trepans taken from the RPV Greifswald Unit 1 containing the X-butt multilayer submerged welding seam and from base metal ring 0.3.1 both located in the beltline region were investigated. Unit 1 represents the IAI condition. It is shown that the Master Curve approach as adopted in ASTM E1921 is applicable to the investigated original WWER-440 weld metal. The evaluated T0 varies through the thickness of the welding seam. The lowest T0 value was measured in the root region of the welding seam representing a uniform fine grain ferritic structure. Beyond the welding root T0 shows a wavelike behaviour. The highest T0 of the weld seam was not measured at the inner wall surface. This is important for the assessment of ductile-to-brittle temperatures measured on sub size Charpy specimens made of weld metal compact samples removed from the inner RPV wall. Our findings imply that these samples do not represent the most conservative condition. Nevertheless, the Charpy transition temperature TT41J estimated with results of sub size specimens after the recovery annealing was confirmed by the testing of standard Charpy V-notch specimens. The evaluated Charpy-V TT41J shows a better accordance with the irradiation fluence along the wall thickness than the Master Curve reference temperature T0. The evaluated T0 from the trepan of base metal ring 0.3.1 varies through the thickness of the RPV wall. T0 increases from −120°C at the inner surface to −104°C at a distance of 33 mm from it and again to −115°C at the outer RPV wall. The KJc values generally follow the course of the MC, although the scatter is large. The re-embrittlement during 2 campaigns operation can be assumed to be low for the weld and base metal.Copyright © 2009 by ASME

Between 1973 and 1990 4 units of the Russian NPP type WWER-440/230 were operated in Greifswald (former GDR). The operation was stopped after the German reunification, because the units did not completely follow western nuclear safety standards. Material probes from the pressure vessels were gained in the frame of the ongoing decommissioning procedure. The investigations of this material started with material from the circumferential core weld of unit 1. This weld was annealed after 13 cycles and operated further for 2 cycles. Additionally, starting with cycle 11, dummy assemblies were inserted to reduce the neutron fluence in the RPV wall. Firstly this paper presents results of the RPV fluence calculations depending on different loading schemes and on the axial weld position based on the Monte Carlo code TRAMO. The results show, that the use of the dummy assemblies reduces the flux by a factor of 2 – 5 depending on the azimuthal position. The fluence increase is reduced to 1/6 at the position of the maximum fluence. The neutron fluence at the different circumferential welds is closely related to their distance to the core. The circumferential core weld (SN0.1.4) received a fluence of 2.4·1019 neutrons/cm2 at the inner surface, it decreases to 0.8·1019 neutrons/cm2 at the outer surface. The neutron fluences at the both other welds are 2 resp. 4 orders of magnitude smaller according to their distances to the core. It should be mentioned that in this cases the fluence gradient can be negative through the wall. The material investigations were done using a trepan from the circumferential core weld. Master Curve and Charpy V-notch testing were applied. Specimens from 7 locations through the thickness of the welding seam were tested. The reference temperature T0 was calculated with the measured fracture toughness values, KJc , at brittle failure of the specimen. Generally the KJc values measured on pre-cracked and side-grooved Charpy size SE(B) specimens of the investigated weld metal follows the course of the Master Curve. The KJc values show a remarkable scatter. In addition the MC SINTAP procedure was applied to determine T0 SINTAP of the brittle fraction of the data set. There are remarkable differences between T0 and T0 SINTAP indicating macroscopic inhomogeneous weld metal. The highest T0 was about 50°C at a distance of 22 mm from the inner surface of the weld. It is 40 K higher compared with T0 at the inner surface. This is important for the assessment of ductile-to-brittle temperatures measured with sub size Charpy specimens made of weld metal from the inner RPV wall. This material does not represent the most conservative condition. Nevertheless, the Charpy transition temperature TT41J estimated with results of sub size specimens after the recovery annealing was confirmed by the testing of standard Charpy V-notch specimens. The VERLIFE procedure prepared for the integrity assessment of WWER RPV was applied on the measured results. It enables the determination of a reference temperature, RTT0 to index a lower bound fracture toughness curve. This curve agrees with the MC 5% fractile as specified in ASTM E1921-05. The measured KJc values are not enveloped by this lower bound curve. However, the VERLIFE lower bound curve indexed with the SINTAP reference temperature RTT0 SINTAP envelops the KJc values. Therefore for a conservative integrity assessment the fracture toughness curve indexed with a RT representing the brittle fraction of a dataset of measured KJc values has to be applied.Copyright © 2008 by ASME

Other than the brittle failure, the ductile behavior of the aged nuclear reactor pressure vessel (RPV) steels is also of interest for the integrity assessment and evaluation of the irradiation response. The fracture toughness of high toughness materials like RPV steels can be characterized by a J-R curve. Since the RPV steel material available for testing purposes, like surveillance specimens, is limited, the single specimen method is used for the J-R curve determination. In this study, J-R curves were measured on Charpy size SE(B) and 1T-C(T) specimens of different RPV steels in the unirradiated and irradiated conditions. It was observed that despite the available sophisticated instrumentation and strict implementation of the recommended test procedures, the J-initiation value for all the different material specimens tested could not be ascertained according to the test standards ASTM E1820 and ISO 12 135. For Charpy size SE(B) specimens, it was found that though valid JIC/J0.2BL values could be obtained in irradiated conditions, in un-irradiated conditions, especially for high toughness RPV steels, it was not possible. The evaluation showed that the aoq fit of the ASTM standard compensates uncertainties in the initial J-Δa value resulting in reliable and more number of qualified test results. But these uncertainties strongly influence the A parameters of the ISO fit and the J0.2BL(B) value. Additionally, in the ISO evaluation the lower offset of the first exclusion line and a higher slope results in lower J0.2BL values compared to the ASTM analysis. Furthermore, for the two specimen geometries the course of J-R curves up to the JQ value was similar even for high toughness materials, but the lower specimen size was disqualified due to the lower prescribed Jlimit. Similarly, the J-R curves for un-irradiated and irradiated condition had a similar course up to the J0.2BL value, even for extremely high irradiation induced embrittlement.