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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 focus of this paper is on the irradiation response and the effect of thermal annealing in weld materials extracted from decommissioned WWER 440 reactor pressure vessels of the nuclear power plant Greifswald. The characterisation is based on the measurement of the hardness, the yield stress, the Master Curve reference temperature, T0, and the Charpy-V transition temperature through the thickness of multi-layer beltline welding seams in the irradiated and the thermally annealed condition. Additionally, the weld bead structure was characterised by light microscopic studies. We observed a large variation in the through thickness T0 values in the irradiated as well as in thermally annealed condition. The T0 values measured with the T–S-oriented Charpy size SE(B) specimens cut from different thickness locations of the multilayer welding seams strongly depend on the intrinsic weld bead structure along the crack tip. The Master Curve, T0, and Charpy-V, TT47J, based ductile-to-brittle transition temperature progressions through the thickness of the multi-layer welding seam do not correspond to the forecast according to the Russian code. In general, the fracture toughness values at cleavage failure, KJc, measured on SE(B) specimens from the irradiated and large-scale thermally annealed beltline welding seams follow the Master Curve description, but more than the expected number lie outside the curves for 2% and 98% fracture probability. In this case the test standard ASTM E1921 indicates the investigated multi-layer weld metal as not uniform. The multi modal Master Curve based approach describes the temperature dependence of the specimen size adjusted KJc-1T values well. Thermal annealing at 475 °C for 152 h results in the expected decrease of the hardness and tensile strength and the shift of Master Curve and Charpy-V based ductile-to-brittle transition temperatures to lower values.

ODS steels have been known to exhibit anisotropic fracture behaviour and form secondary cracks. In this work, the factors responsible for the anisotropic fracture behaviour have been investigated using scanning electron microscopy and electron backscatter microscopy. Fracture toughness of hot rolled 13Cr ODS steel was determined using unloading compliance method for L-T and T-L orientations at various temperatures. L-T orientation had higher fracture toughness than T-L orientation and also contained more pronounced secondary cracking. Secondary cracks appeared at lower loads than primary cracks in both orientations. Primary crack propagation was found to be preferentially through fine grains in a bimodal microstructure. Grains were aligned and elongated the most towards rolling direction followed by T and S directions resulting in fracture anisotropy. Crystallographic texture and preferential alignment of Ti enriched particles parallel to rolling direction also contributed towards fracture anisotropy.

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.

Reactor pressure vessel (RPV) multilayer welding seams show an inhomogeneous structure. It raises concerns that the evaluation of non-uniform material might not be amenable to the statistical analysis methods on which the Master Curve approach is based. In particular with regard to weld metals, it can be expected that the cleavage fracture toughness is strongly influenced by the orientation of the Charpy size SE(B) specimen. The T-L oriented SE(B) specimen (axis axial and crack propagation in circumferential direction) comprises of various welding beads along the crack front whereas in a L-S specimen (axis axial and crack propagation through the thickness) the crack tip is located in one welding bead with an approximately uniform structure. The paper summarises fracture toughness results measured on welding seams of decommissioned and non-commissioned RPVs of WWER type nuclear reactors and the non-commissioned Biblis-C RPV. Specimens of T-L and T-S orientation were tested. The results show, that in general the cleavage fracture toughness values, KJc-1T, follow the Master Curve description. However, the number of KJc-1T data outside the 2% and 98% tolerance bounds is larger than predicted by the underlying model, which indicates non-uniform material. There is a large variation in the evaluated through thickness T0 values of the investigated multilayer beltline welding seams. Within the sampling range of the surveillance specimens, T0 values vary with a span of 30 to about 60 K depending on the applied welding technology. The fracture toughness strongly depends on the intrinsic weld bead structure. Hence, the position of the fatigue crack tip of the pre-cracked SE(B) specimen at the multilayer welding seam is crucial and defines the cleavage fracture toughness. Modified Master Curve based evaluation procedures like the MC based approach of the SINTAP procedure were applied to get fracture toughness values which are representative for the most brittle fraction the test series. Despite of the pronounced non-homogeneity of the micro-structure along the crack front of T-L specimens, crack initiation sites are randomly distributed along the crack front. This means that one of the basic assumptions in ASTM E1921, i.e. the uniform distribution of initiation sites, is fulfilled also for the T-L specimens from the multilayer weld metal.

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 Master Curve (MC) approach and the associated reference temperature, To, as defined in the test standard ASTM E1921, is rapidly moving from the research laboratory to application in integrity assessment of components and structures. To is the index temperature for the universal MC, which considers the toughness behaviour of a specific material. The Structural Integrity Assessment Procedures for European Industry (SINTAP) contain a MC extension for analysing the fracture behaviour of inhomogeneous ferritic steels. This paper presents the application of the MC approach to the To determination of different types of Russian WWER-type reactor pressure vessel (RPV) steels. In addition the SINTAP-MC approach was applied to determine an alternative reference temperature, T R . The influence of different microstructures and compositions within one type of RPV steel and the effect of irradiation with fast neutrons on To are experimentally evaluated. In general the MC based To is about 72 K below the Charpy V-notch transition temperature related to an impact energy of 48 J. The paper demonstrates the application of MC based To and T R as an alternative reference temperature for neutron embrittled RPV steels used in the RPV integrity assessment.

The Master Curve approach for the fracture toughness evaluation is expected to be a powerful tool to ensure the reliability of long-term used RPV steels. In order to get sufficient number of data for the Master Curve approach coexistent with the present surveillance program for RPVs, the utilization of miniature specimens, which can be taken from broken halves of surveillance Charpy specimens, is important. CRIEPI developed the test technique for the miniature C(T) specimens (Mini-CT), whose dimensions are 4 × 10 × 10 mm, and verified the basic applicability of Master Curve approach by means of Mini-CT for the determination of fracture toughness of typical Japanese RPV steels. A round robin program is organized in order to assure the robustness of the testing procedure to the difference in testing machines or operators. The first and second round robin tests (PVP2012-78661 [1], PVP2013-97936 [2]) suggested that the reference temperature T0 evaluation technique by Mini-CT specimen potentially is fairly robust in regard to difference in testing machines and operators, and gives similar loading rate dependency to the larger C(T) specimens. As the final year of the round robin program, “blind tests” were carried out. Here, detailed material information such as the type of materials, estimated T0, existing fracture toughness data for the material, were not given with the specimen, and 6 organizations independently selected the test temperature based on Charpy full curve of the tested material. The selection of test temperature has the variation of −120 °C to −150 °C among the organizations. 8 to 20 specimens in a set were subjected to the Master Curve evaluation and all the 6 organizations successfully obtained valid T0. The scatter range in T0 was at most 16 °C, which was within the acceptable scatter range specified in ASTM E1921-10e1. The selection of test temperature seems to give limited effect as like as that in larger specimens.Copyright © 2014 by ASME

The reference temperature T0 was measured for both T-S and T-L- specimen orientation in 24 layers across the thickness of the beltline weld of a reactor pressure vessel. It turned out to vary in a bandwidth of more than 40K. Because of a high scatter, no clear pattern of T0 as a function of the thickness position could be recognized. A more detailed analysis revealed that the median of KJc was considerably steeper than predicted by the Master-Curve, which leads to a bias of T0 with respect to the test- temperature relative to T0. By a modified evaluation procedure, the scatter of the reference temperature could be significantly reduced, which enabled the global pattern of T0 to be recognized. By comparing the theoretical lower bound to KJc-data of the used specimens with the individual measured KJc a representative T0 that characterizes the overall toughness behaviour of the weld was determined. It turned out to be about 10 K lower than the maximum local T0.

The investigation of reactor pressure vessel (RPV) material from the decommissioned Greifswald NPP representing the first generation of Russian type WWER-440/V-230 reactors offers the opportunity to evaluate the real toughness response. The Greifswald RPVs represent different material conditions viz. irradiated, irradiated & annealed and irradiated, annealed and re-irradiated. The paper presents test results measured on the trepan taken from the forged base metal ring 0.3.1. located in the reactor core region of the Unit 4 RPV. This unit was shut down after 11 years of operation and represents the irradiated condition. The key part of the testing is focused on the determination of the reference temperature T0 following the ASTM test standard E1921-10. The T0 of 11 thickness locations from the inner to the outer RPV wall varies between 112°C and 130°C with a mean value of 121°C. These are very low values for WWER RPV steel irradiated with fluences between 5.4 to 1.2•1019n/cm2 (E>0.5 MeV) from the inner to the outer RPV wall. The fracture toughness values at cleavage failure, KJc, measured on LS oriented pre-cracked and side-grooved Charpy size SE(B) specimens from defined thickness locations of the forged ring strongly scatter. More than allowed 2% of the specimen size adjusted KJc-1T values lie below the fracture toughness curve for 2% fracture probability. The application of modified MC based evaluation methods indicates the material as non-homogeneous. Because of very low KJc values the SINTAP step 3 evaluation gave a maximum T0SINTAP of -40°C. The multimodal MC evaluation of KJc values from all thickness locations gives a T0MM of -118°C and a standard deviation of 25 K. The course of the fracture toughness curves evaluated by multi modal approach also do not represent the measured KJc values since more than 2% lie below the fracture toughness curve for 2% fracture probability. The reason for the occurrence of very low KJc values is seen in intergranular planes detected on the fractured surfaces of the specimens.

The precracked Charpy single-edge notched bend, SE(B), specimen (PCC) is the most likely specimen type to be used for determination of the reference temperature, T0, with reactor pressure vessel (RPV) surveillance specimens. Unfortunately, for many RPV steels, significant differences have been observed between the T0 temperature for the PCC specimen and that obtained from the 25-mm thick compact specimen [1TC(T)], generally considered the standard reference specimen for T0. This difference in T0 has often been designated a specimen bias effect, and the primary focus for explaining this effect is loss of constraint in the PCC specimen. The International Atomic Energy Agency (IAEA) has developed a coordinated research project (CRP) to evaluate various issues associated with the fracture toughness Master Curve for application to light-water RPVs. Topic Area 1 of the CRP is focused on the issue of test specimen geometry effects, with emphasis on determination of T0 with the PCC specimen and the bias effect. Topic Area 1 has an experimental part and an analytical part. Participating organizations for the experimental part of the CRP performed fracture toughness testing of various steels, including the reference steel JRQ (A533-B-1) often used for IAEA studies, with various types of specimens under various conditions. Additionally, many of the participants took part in a round robin exercise on finite element modeling of the PCVN specimen, discussed in a separate paper. Results from fracture toughness tests are compared with regard to effects of specimen size and type on the reference temperature T0. It is apparent from the results presented that the bias observed between the PCC specimen and larger specimens for Plate JRQ is not nearly as large as that obtained for Plate 13B (−11°C vs −37°C) and for some of the results in the literature (bias values as much as −45°C). This observation is consistent with observations in the literature that show significant variations in the bias that are dependent on the specific materials being tested. There are various methods for constraint adjustments and two methods were used that reduced the bias for Plate 13B from −37°C to −13°C in one case and to − 11°C in the second case. Unfortunately, there is not a consensus methodology available that accounts for the differences observed with different materials. Increasing the Mlim value in the ASTM E-1921 to ensure no loss of constraint for the PCC specimen is not a practicable solution because the PCC specimen is derived from CVN specimens in RPV surveillance capsules and larger specimens are normally not available. Resolution of these differences are needed for application of the master curve procedure to operating RPVs, but the research needed for such resolution is beyond the scope of this CRP.

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

In the final evaluation for the application of the Master Curve in the IAEA Coordinated Research Project Phase 5 (CRP-5), one of the areas which was identified as needing further work concerned the effects of loading rate on the reference temperature To up to impact loading conditions. This subject represents one of the three topic areas within the current CRP-8. The effect of loading rate can be broken down into two distinct aspects: 1) the effect of loading rate on the Master Curve To values for loading rates within the specified in ASTM E1921-05 for quasi-static loading (0.1–2 MPa√m/s); 2) the effect of loading rate on To values for higher loading rates, including impact conditions using instrumented precracked Charpy (PCC) specimens. The new CRP includes both aspects, but primarily focuses on the second element of loading rate effects, i.e. loading rates above 2 MPa√m/s. These issues are investigated within the topic area #2 of CRP-8 (Loading Rate Effect). The mandatory portion of this topic area required participation in a round-robin exercise (RRE) to validate the application of the Master Curve approach to PCC specimens tested in the ductile-to-brittle transition region using an instrumented pendulum (10 tests per participant on the JRQ material). The current status of the RRE is presented in [1]. The non-mandatory portion of this topic area consists in providing Master Curve data obtained at different loading rates on various RPV steels, in order to assess the loading rate dependence of To and compare it with an empirical model proposed by Wallin. Moreover, additional topics will be addressed, such as: • comparison of results from unloading compliance and monotonic loading in the quasi-static range; • estimation of fracture toughness from Charpy V-notch data; • assessment of crack arrest properties from instrumented Charpy results; • effect of irradiation on the relationship between static and dynamic fracture toughness.Copyright © 2007 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

In the final evaluation for the application of the Master Curve in the IAEA Coordinated Research Project Phase 5 (CRP-5), one of the areas which was identified as needing further work concerned the effects of loading rate on the reference temperature To up to impact loading conditions. This subject represents one of the three topic areas within the current CRP-8. The effect of loading rate can be broken down into two distinct aspects: 1) the effect of loading rate on the Master Curve To values for loading rates within the specified in ASTM E1921-05 for quasi-static loading (0.1–2 MPa√m/s); 2) the effect of loading rate on To values for higher loading rates, including impact conditions using instrumented precracked Charpy (PCC) specimens. The new CRP includes both aspects, but primarily focuses on the second element of loading rate effects, i.e. loading rates above 2 MPa√m/s. These issues are investigated within the topic area #2 of CRP-8 (Loading Rate Effect). The mandatory portion of this topic area required participation in a round-robin exercise (RRE) to validate the application of the Master Curve approach to PCC specimens tested in the ductile-to-brittle transition region using an instrumented pendulum (10 tests per participant on the JRQ material). The current status of the RRE is presented in [1]. The non-mandatory portion of this topic area consists in providing Master Curve data obtained at different loading rates on various RPV steels, in order to assess the loading rate dependence of To and compare it with an empirical model proposed by Wallin. Moreover, additional topics will be addressed, such as: • comparison of results from unloading compliance and monotonic loading in the quasi-static range; • estimation of fracture toughness from Charpy V-notch data; • assessment of crack arrest properties from instrumented Charpy results; • effect of irradiation on the relationship between static and dynamic fracture toughness.

The EURATOM 6th Framework Integrated Project PERFECT (Prediction of Irradiation Damage Effects in Reactor Components) addresses irradiation damage in RPV materials and components by multi-scale modeling. This approach offers many potential advantages over the conventional empirical methods used in current practice of nuclear plant lifetime management. Launched in January 2004, PERFECT is a 48-month project focusing on two main components of nuclear power plants which are subject to irradiation damage: the ferritic steel reactor pressure vessel (RPV), and the austenitic steel internals. It is the purpose of the present paper to provide an update of progress of work being carried out in the Mechanics Sub-project of PERFECT to predict the fracture behavior of RPVs in PWR and WWER systems.Copyright © 2007 by ASME

Within topic area 2 (Loading Rate Effects) of the IAEA Coordinated Research Project Phase 8 (CRP-8), a round-robin exercise (RRE) was performed. This was aimed at measuring impact fracture toughness in the ductile-to-brittle transition region using precracked Charpy specimens of the JRQ steel tested on instrumented pendulum machines. The objective of the round-robin was twofold: 1) assessing the applicability of the Master Curve approach (ASTM E1921-05) to data obtained at impact loading rates of the order of 105 MPa√m/s; 2) validating a draft test procedure developed by sub-committee TC5 of the European Structural Integrity Society (ESIS), in the absence of an official testing standard (ISO or ASTM). A comprehensive analysis of the RRE results received to date will be presented in this paper, including an assessment of the overall data scatter and the lab-to-lab variability. In this context, an outlier data set has been identified and investigated in great detail. Based on the results presented, it is noted that the test procedure appears reliable and easy to follow for the various participants.Copyright © 2007 by ASME

Pressure vessel integrity assessment after long-term service irradiation is commonly based on surveillance program results. Nevertheless, only the investigation of RPV material from decommissioned NPPs enables the evaluation of the real toughness response. Such a chance is given now through the investigation of material from the former Greifswald NPP (VVER-440/230) to evaluate the material state of a standard RPV design and to assess the quality of prediction rules and assessment tools. The operation of the four Greifswald units was finished in 1991 after 12--15 years of operation. In autumn 2005 the first trepans (diameter 120 mm) were gained from the unit 1 of this NPP. Some details of the trepanning procedure will be given. The paper mainly deals with the retrospective dosimetry based on Niobium, which is a trace element of the RPV material. The reaction {sup 93}Nb(n,n'){sup 93m}Nb with an energy dependence highly correlated to radiation damage and a half life of the reaction product of 16.13 years is well suited for retrospective fast neutron dosimetry. Fluence calculations using the code TRAMO were based on pin-wise time dependent neutron sources and an updated nuclear data base (ENDF/B-VI release 8). The neutron spectra were determined at the trepan positions. Themore » different loading schemes of unit 1 (standard and with 4 or 6 dummy assemblies) were taken into account. The calculated specific {sup 93m}Nb activities for February, 2006 at the sample positions were determined to 16.3 Bq/{mu}g Nb for sample 1, (0.1 cm distance from inner wall), and 4.0 Bq/{mu}g Nb for sample 2 (11.5 cm distance from inner wall). Unfortunately, a second neutron reaction besides {sup 93}Nb(n,n') leading to {sup 93m}Nb-activity is the reaction {sup 92}Mo(n,{gamma}){sup 93}Mo. {sup 93}Mo decays by electron capture to {sup 93m}Nb with a half life of 4000 years and a branching ratio b{sub r} = 0.88. As (n,{gamma})-reactions are produced mainly by low energy neutrons, being less important for material damage, the {sup 93m}Nb-activity generated through the Mo-path should be determined separately and subtracted from the measured activity. For the sample 1 in the maximum flux azimuthal position of weld SN4 with a Nb-content of 8 ppm and an Mo-content of 4000 ppm for February 3, 2006 was obtained a Mo-induced {sup 93m}Nb-activity of 80 Bq/g steel, amounting to 37.7% of the total {sup 93m}Nb-activity. It turns out that the {sup 93m}Nb generation on the second path is nearly of the same order as the fast neutron induced generation from Niobium. For the experimental determination of the {sup 93m}Nb-activity the Nb-content was determined by ICP-MS (inductive coupled plasma mass spectrometry) after dissolution of the material sample. The radiochemical isolation of Nb was done by anion exchange separation. The radiochemical separation was accompanied by determination of the chemical yield of Nb using again the ICP-MS method. The measurement of the {sup 93m}Nb activity was realized by Liquid Scintillation Spectrometry (LSC). The first comparison between the calculated and the measured {sup 93m}Nb activities resulted in deviations between 15 and 50%. Possible reasons for the observed differences are discussed. (authors)« less

The EURATOM 6th Framework Integrated Project PERFECT (Prediction of Irradiation Damage Effects in Reactor Components) addresses irradiation damage in RPV materials and components by multi-scale modeling. This state-of-the-art approach offers many potential advantages over the conventional empirical methods used in current practice of nuclear plant lifetime management. Launched in January 2004, this 48-month project is focusing on two main components of nuclear power plants which are subject to irradiation damage: the ferritic steel reactor pressure vessel, and the austenitic steel internals. It is the purpose of the present paper to provide an overview of work being carried out in the RPV Mechanics Sub-project of PERFECT to predict the fracture behavior of PWR, BWR and WWER systems.Copyright © 2005 by ASME

The activities of nuclides produced via the neutron irradiation of reactor pressure vessel (RPV) steel are used to validate respective fluence calculations. Niobium, nickel, and technetium isotopes from RPV trepans of the decommissioned NPP Greifswald (VVER-440) have been analyzed. The activities were determined by TRAMO (Monte-Carlo) fluence calculations, newly applying 640 neutron-energy groups and ENDF/B7 data. Relative to earlier results, fluences up to 20% higher have been computed, leading to somewhat better agreement between measurement and calculation, particularly in the case of Tc-99. (authors)

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.

The investigation of reactor pressure vessel (RPV) materials from decommissioned nuclear power plants (NPP) offers the unique opportunity to scrutinize the irradiation behavior under real conditions. The paper describes the investigation of trepans taken from the decommissioned WWER-440 RPVs of the Greifswald NPP. The key part of the testing is aimed at the determination of the reference temperature T0 following the ASTM Test Standard E1921 to determine the fracture toughness of the RPV steel in different thickness locations. In a first step, the trepan taken from the RPV Greifswald Unit 1 containing the multilayer welding seam located in the beltline region was investigated. This welding seam represents the irradiated, recovery annealed, and reirradiated 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. After an initial increase of T0 from 10°C at the inner surface to 49°C at 22 mm distance from it, T0 decreases to −32°C at a distance of 70 mm, finally increasing again to 61°C near the outer RPV wall. The lowest T0 value was measured in the root region of the welding seam representing a uniform fine grain ferritic structure. 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 subsize 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 subsize specimens after the recovery annealing, was confirmed by the testing of standard Charpy V-notch specimens.