Fracture toughness of the hydrogen charged EUROFER 97 RAFM steel at room temperature and 120 °C

The static fracture toughness of EUROFER 97 reduced activation ferritic-martensitic steel was investigated in presence of higher content of hydrogen. The hydrogen effect is shown during fracture toughness testing both of base and weld metals at room temperature and at 120 °C. At the room temperature testing the J_0.2 integral values will decrease depending on hydrogen content in the range of 2-4 wppm. The same hydrogen content of 2 wppm manifests itself by an uneven level of hydrogen embrittlement for base metal and weld metal. This corresponds to a different J_0.2 integral value and a different mechanism of fracture mode. At the hydrogen content of 4 wppm more evident decrease of J_0.2 was observed for both metals. At 120 °C hydrogen decreases J_0.2 integral in base metal at a limited scale only in comparison to weld metal. At room temperature and hydrogen content of about 4 wppm the base metal specimen exhibits inter-granular fracture and trans-granular cleavage on practically the whole crack surface. The weld metal fracture has shown inter-granular and trans-granular mechanism with ductile and dimple rupture.     

Post-irradiation mechanical properties of three EUROFER97 joints

Within the development of reduced activation ferritic martensitic (RAFM) steels as prominent structural materials for future fusion reactors, EUROFER97 has recently emerged in Europe as the reference material for the DEMO design. In order to characterise the in-service performance of EUROFER97 as structural material, it is important to assess the properties of welded joints, particularly under irradiation. In the present paper, three EUROFER97 joints (two diffusion welds and one TIG weld) have been irradiated in the BR2 reactor of SCK-CEN at 300 °C up to 1.8 dpa and subsequently characterised for tensile, impact and fracture toughness properties. Comparisons of the results are provided with base EUROFER97 (both unirradiated and irradiated under similar conditions) and, where available, with properties measured on the joints in the unirradiated condition. The post-irradiation mechanical behaviour of both diffusion joints (“laboratory” and “mock-up”) appears similar to that of the base material; therefore, diffusion joining looks a very promising technique. On the other hand, the properties of the TIG joint are affected by the lack of a post-weld heat treatment, which causes the material from the upper part of the weld to be significantly worse than that of the lower region.     

Fine structure behaviour of VVER-1000 RPV materials under irradiation

Changes in the fine structure and mechanical properties of the base metal (BM) and weld metal (WM) of VVER-1000 pressure vessels during accumulation of neutron dose in the range of fluences ∼(3.2-15) × 10²³ m⁻² (E > 0.5 MeV) at 290 °C are studied using methods of transmission electron microscopy, fractographic analysis, and Auger electron spectroscopy. A correlation was found between the changes of mechanical properties and the micro- and nano-structures of the studied steels. Accumulation of neutron dose considerably raises the strength characteristics and transition temperature of VVER-1000 pressure vessel steels. The rate of changes in the mechanical properties of the weld metal is significantly higher than that of the base metal. The slower growth of strength characteristics and transition temperature shift of the base metal under irradiation as compared with the weld metal is due to the slower growth of the density of radiation defects and radiation-induced precipitates. The level of intergranular embrittlement under irradiation in the weld metal is not higher then in the base metal in spite of the higher content of nickel.     

Evolution of the nanostructure of VVER-1000 RPV materials under neutron irradiation and post irradiation annealing

A high nickel VVER-1000 (15Kh2NMFAA) base metal (1.34 wt% Ni, 0.47% Mn, 0.29% Si and 0.05% Cu), and a high nickel (12Kh2N2MAA) weld metal (1.77 wt% Ni, 0.74% Mn, 0.26% Si and 0.07% Cu) have been characterized by atom probe tomography to determine the changes in the microstructure during neutron irradiation to high fluences. The base metal was studied in the unirradiated condition and after neutron irradiation to fluences between 2.4 and 14.9 × 10²³ m⁻² (E > 0.5 MeV), and the weld metal was studied in the unirradiated condition and after neutron irradiation to fluences between 2.4 and 11.5 × 10²³ m⁻² (E > 0.5 MeV). High number densities of ∼2-nm-diameter Ni-, Si- and Mn-enriched nanoclusters were found in the neutron irradiated base and weld metals. No significant copper enrichment was associated with these nanoclusters and no copper-enriched precipitates were observed. The number densities of these nanoclusters correlate with the shifts in the ΔT_{41 J} ductile-to-brittle transition temperature. These nanoclusters were present after a post irradiation anneal of 2 h at 450 °C, but had dissolved into the matrix after 24 h at 450 °C. Phosphorus, nickel, silicon and to a lesser extent manganese were found to be segregated to the dislocations.     

Prediction of irradiation embrittlement of vanadium alloyed low nickel steel for future reactors

A comparison between pearlitic 2CrMoV steel (WWER-440) and 9% Cr based ferritic-martensitic steels (EUROFER 97 and LA12TaLC) is presented as regards irradiation induced ductile-brittle transition temperature shifts. For neutron doses of 1.5-2 dpa and irradiation temperatures around 300 °C the transition temperature shifts for WWER-440 steel and EUROFER 97 welds are comparable. In the temperature range 350-500 °C the radiation embrittlement levels of both steels are low. Moreover, post-irradiation annealing is proposed as a promising method to predict results of high temperature irradiation embrittlement from existing lower temperature irradiation embrittlement data.     

Mechanical properties of irradiated ODS-EUROFER and nanocluster strengthened 14YWT

Irradiations to 1.5 dpa at 300-750 °C were conducted to investigate the changes in mechanical properties of an advanced nanocluster strengthened ferritic alloy, designated 14YWT, and an oxide dispersion strengthened ferritic alloy ODS-EUROFER. Two non-dispersion strengthened variants, 14WT and EUROFER 97, were also irradiated and tested. Tensile results show 14YWT has very high tensile strengths and experienced some radiation-induced hardening, with an increase in room temperature yield strength of 125 MPa after irradiation, while results for ODS-EUROFER show a 275 MPa increase following irradiation. Master curve fracture toughness analysis show 14YWT has a cryogenic T_o reference temperatures before and after irradiation of about −188 and −176 °C, respectively, and upper-shelf K_JIc values between 175 and 225 MPa√m. The favorable fracture toughness properties and resistance to radiation-induced changes in mechanical properties observed for 14YWT are attributed to a fine grain structure and high number density of Y-Ti-O nanoclusters.     

Liquid metal embrittlement susceptibility of ferritic-martensitic steel in liquid lead alloys

The susceptibility of the ferritic-martensitic steels T91 and EUROFER97 to liquid metal embrittlement (LME) in lead alloys has been examined under various conditions. T91, which is currently the most promising candidate material for the high temperature components of the future accelerator driven system (ADS) was tested in liquid lead bismuth eutectic (LBE), whereas the reduced activation steel, EUROFER97 which is under consideration to be the structural steel for fusion reactors was tested in liquid lead lithium eutectic. These steels, similar in microstructure and mechanical properties in the unirradiated condition were tested for their susceptibility to LME as function of temperature (150-450 °C) and strain rate (1 × 10⁻³-1 × 10⁻⁶ s⁻¹). Also, the influence of pre-exposure and surface stress concentrators was evaluated for both steels in, respectively, liquid PbBi and PbLi environment. To assess the LME effect, results of the tests in liquid metal environment are compared with tests in air or inert gas environment. Although both unirradiated and irradiated smooth ferritic-martensitic steels do not show any or little deterioration of mechanical properties in liquid lead alloy environment compared to their mechanical properties in gas as function of temperature and strain rate, pre-exposure or the presence of surface stress concentrators does lead to a significant decrease in total elongation for certain test conditions depending on the type of liquid metal environment. The results are discussed in terms of wetting enhanced by liquid metal corrosion or crack initiation processes.     

Effect of thermal aging on fracture toughness of RPV steel

The effect of thermal aging on mechanical properties and fracture toughness was investigated on pressure vessel steel of light water reactors. Submerged are welded plates of ASME SA508 C1.3 steel were isothermally aged at 350°C, 400°C and 450°C for up to 10,000 hrs. Tensile, Charpy impact and fracture toughness testings were conducted on the base metal and the weld heat affected zone (HAZ) material to evaluate whether thermal aging induced by the plant operation is critical for the integrity of the pressure vessel or not. Tensile properties of the base metal was not changed by thermal aging as far as the thermal aging conditions were concerned. Relatively distinct degradation was observed in fracture toughness J_IC and J-resistance properties of both the base metal and the weld HAZ material, while only slight changes were observed in Charpy impact properties for both of them. However, it was concluded that the effect of thermal aging estimated by 40–80 years of plant operation on fracture toughness of both materials is small.     

Behavior of ferritic/martensitic steels after n-irradiation at 200 and 300 °C

High chromium ferritic/martensitic (F/M) steels are considered as the most promising structural materials for accelerator driven systems (ADS). One drawback that needs to be quantified is the significant hardening and embrittlement caused by neutron irradiation at low temperatures with production of spallation elements. In this paper irradiation effects on the mechanical properties of F/M steels have been studied and comparisons are provided between two ferritic/martensitic steels, namely T91 and EUROFER97. Both materials have been irradiated in the BR2 reactor of SCK-CEN/Mol at 300 °C up to doses ranging from 0.06 to 1.5 dpa. Tensile tests results obtained between −160 °C and 300 °C clearly show irradiation hardening (increase of yield and ultimate tensile strengths), as well as reduction of uniform and total elongation. Irradiation effects for EUROFER97 starting from 0.6 dpa are more pronounced compared to T91, showing a significant decrease in work hardening. The results are compared to our latest data that were obtained within a previous program (SPIRE), where T91 had also been irradiated in BR2 at 200 °C (up to 2.6 dpa), and tested between −170 °C and 300 °C. Irradiation effects at lower irradiation temperatures are more significant.     

Investigation of samples taken from Kozloduy unit 2 reactor pressure vessel

Within the framework of the 6 month WANO program, small samples were cut from the inside surface of the Kozloduy NPP unit 2 reactor pressure vessel to assess the actual condition of the pressure vessel material before and after annealing. The actual values of the weld metal characteristics required for estimating radiation-limited lifetime—the ductile-to-brittle transition temperature (DBTT) in the initial state (T_ko) and the phosphorus and copper contents which affect the radiation stability of steel—were not determined during manufacturing. The Kozloduy unit 2 pressure vessel had no surveillance program. Radiation stability was evaluated using dependencies based on analysis results for surveillance samples taken from other VVER-440 reactors. For this reason, the actual pressure vessel characteristics and their changes in the course of reactor operation, as well as comparison of experimental with calculated data were the principle objectives of the study.Instrumented impact tests were carried out on sub-size specimens of base and weld metal. Correlation dependencies were used with standard tests to determine DBTTs for the base and weld metal (in accordance with Russian standards): base metal before annealing 40 °C, after annealing 16 °C; weld metal before annealing 212 °C, after annealing 70 °C.The estimated value of T_ko, for the initial, unirradiated weld metal, was 50 °C. The experimental results were compared with a prediction of the extent of radiation-induced embrittlement of Kozloduy unit 2 pressure vessel materials. It was confirmed that radiation-induced embrittlement of the base metal does not impose any limits on the radiation-limited lifetime of the pressure vessel.The predicted increase in the DBTT of the weld metal as a result of irradiation (about 165 °C) is practically equal to the experimental result (162 °C). However, the value of T_f obtained from tests before annealing (212 °C) is about 40 °C higher that the estimated value, i.e. the calculation does not produce a conservative estimate. This was explained by a low estimate of T_ko (10 °C), which had been calculated using data from chemical analysis of the weld metal, performed by the manufacturer. The investigations on the samples, however, yielded an estimated value of T_ko = 50 °C.The effectiveness of annealing in restoring the mechanical properties of irradiated VVER-440 reactor pressure vessels was confirmed. Recovery annealing lowered the DBTT of the weld metal by 85% or more of its radiation-induced shift.     

Tensile and fracture toughness properties of neutron-irradiated CuCrZr

Tensile and fracture toughness properties of a precipitation-hardened CuCrZr alloy were investigated in two heat treatment conditions: solutionized, water quenched and aged (CuCrZr SAA), and hot isostatic pressed, solutionized, slow-cooled and aged (CuCrZr SCA). The second heat treatment simulated the manufacturing cycle for large components, and is directly relevant for the ITER divertor components. Specimens were neutron irradiated at ∼80 °C to two fluences, 2 × 10²⁴ and 2 × 10²⁵ n/m² (E > 0.1 MeV), corresponding to displacement doses of 0.15 and 1.5 displacements per atom (dpa). Tensile and fracture toughness tests were carried out at room temperature. Significant irradiation hardening and plastic instability at yield occurred in both heat treatment conditions with a saturation dose of ∼0.1 dpa. Neutron irradiation slightly reduced fracture toughness in CuCrZr SAA and CuCrZr SCA. The fracture toughness of CuCrZr remained high up to 1.5 dpa (J_Q > 200 kJ/m²) for both heat treatment conditions.     

Monitoring the embrittlement of reactor pressure vessel steels by using the Seebeck coefficient

The degree of embrittlement of the reactor pressure vessel (RPV) limits the lifetime of nuclear power plants. Therefore, neutron irradiation-induced embrittlement of RPV steels demands accurate monitoring. Current federal legislation requires a surveillance program in which specimens are placed inside the RPV for several years before their fracture toughness is determined by destructive Charpy impact testing. Measuring the changes in the thermoelectric properties of the material due to irradiation, is an alternative and non-destructive method for the diagnostics of material embrittlement. In this paper, the measurement of the Seebeck coefficient () of several Charpy specimens, made from two different grades of 22 NiMoCr 37 low-alloy steels, irradiated by neutrons with energies greater than 1 MeV, and fluencies ranging from 0 up to 4.5 × 10¹⁹ neutrons per cm², are presented. Within this range, it was observed that increased by ≈500 nV/°C and a linear dependency was noted between and the temperature shift ΔT_41 J of the Charpy energy vs. temperature curve, which is a measure for the embrittlement. We conclude that the change of the Seebeck coefficient has the potential for non-destructive monitoring of the neutron embrittlement of RPV steels if very precise measurements of the Seebeck coefficient are possible.     

Assessment of neutron irradiation effects on RAFM steels

The objective of the current work is assessment of the effects of neutron irradiation on the mechanical properties of blanket and divertor materials of a future fusion energy generation Demonstration Power Plant (DEMO). The emphasis is put on the review of the tensile, Charpy impact and fracture toughness properties of EUROFER97 and EUROFER ODS (9%Cr) steels irradiated up to a displacement damage dose of 80 dpa in order to address (i) irradiation dose dependence of mechanical properties, (ii) irradiation temperature dependence of the mechanical properties, (iii) helium effects on the mechanical properties. The assessment will be used to give recommendations on the operating temperature range for the First Wall and helium cooled Breeding Blanket materials and to identify needs for structural materials R&D.     

Atom probe tomography characterization of radiation-sensitive KS-01 weld

The microstructure of a radiation-sensitive KS-01 test weld has been characterized by atom probe tomography. The levels of copper, manganese, nickel and chromium in this weld were amongst the highest of all the steels used in Western reactor pressure vessels. After neutron irradiation to a fluence of 0.8 × 10²³ n m⁻² (E>1 MeV) at a temperature of 288 °C, this weld exhibited a large Charpy T_41J shift of 169 K, a large shift of the fracture toughness transition temperature of 160 K, a decrease in upper shelf energy from 118 to ∼78 J, and an increase in the yield strength from 600 to 826 MPa. However, the mechanical properties data conformed to the master curve. Atom probe tomography revealed a high number density (∼3 × 10²⁴ m⁻³) of Cu-, Mn-, Ni-, Si- and P-enriched precipitates and a lower number density (∼1  × 10²³ m⁻³) of P clusters.     

Radiation-damage in molybdenum-rhenium alloys for space reactor applications

Various Mo-Re alloys are attractive candidates for use as fuel cladding and core structural materials in spacecraft reactor applications. Molybdenum alloys with rhenium contents of 41-47.5% (wt%), in particular, have good creep resistance and ductility in both base metal and weldments. However, irradiation-induced changes such as transmutation and radiation-induced segregation could lead to precipitation and, ultimately, radiation-induced embrittlement. The objective of this work is to evaluate the performance of Mo-41Re and Mo-47.5Re after irradiation at space reactor relevant temperatures. Tensile specimens of Mo-41Re and Mo-47.5Re alloys were irradiated to ∼0.7 displacements per atom (dpa) at 1073, 1223, and 1373 K and ∼1.4 dpa at 1073 K in the High Flux Isotope Reactor at Oak Ridge National Laboratory. Following irradiation, the specimens were strained to failure at a rate of 1 × 10⁻³ s⁻¹ in vacuum at the irradiation temperature. In addition, unirradiated specimens and specimens aged for 1100 h at each irradiation temperature were also tested. Fracture mode of the tensile specimens was determined. The tensile tests and fractography showed severe embrittlement and IG failure with increasing temperatures above 1100 K, even at the lowest fluence. This high temperature embrittlement is likely the result of irradiation-induced changes such as transmutation and radiation-induced segregation. These factors could lead to precipitation and, ultimately, radiation-induced embrittlement. The objective of this work is to examine the irradiation-induced degradation for these Mo-Re alloys under neutron irradiation.     

Assessment of toughness in long term service CrMo low alloy steel by fracture toughness and small punch test

Small punch test (SPT) is a miniature sample test technique which can evaluate in-service material properties with an almost non-destructive method. In this paper, the 2.25Cr1Mo steel samples serviced for 10 years in hydrogenation reactor (with temper embrittlement), 1.25Cr0.5Mo supper-pressure vapor pipe serviced for 14 years at 520 °C and several other low alloy steels have been studied by J_IC fracture toughness and SPT. The linear relationship between the small punch (SP) equivalent fracture strain and the fracture toughness of J_IC was created. The correlations applied to the experimental data indicated advantages of using SPT for the determining fracture toughness of in-serviced low alloy steels. Additionally, size affects the fracture pattern. Small punch samples of small size show dimple fractures whereas large fracture toughness samples show quasi-cleavage fractures.     

Review of B & W reactor vessel irradiated materials data

The neutron embrittlement response of the Mn-Mo-Ni/Linde 80 submerged-arc weld metals are evaluated. The original evaluations of the weld metals indicated a high sensitivity to neutron radiation. This paper reviews studies to determine the role of initial properties on the irradiated properties, the relationship of weld wire and weld flux on cross referencing irradiated data, and the role of irradiation temperature. Evaluation of irradiated data as a function of initial toughness value suggest that it may not be necessary to know the unirradiated values to be able to predict the irradiated behavior. This would reduce the uncertainty of the irradiated value. A phenomenon defined as stabilization, was observed wherein further irradiation damage is counteracted by the annihilation, or annealing, of previously formed damage. A correlation is presented to accommodate differences in irradiation temperature for power reactor data. A correlation is presented for predicting the irradiation induced reduction of Charpy upper-shelf energy based on the irradiated value rather than a ratio based on the unirradiated upper-shelf energy. A standard error for the irradiated Charpy upper-shelf data is given.     

Low activation steels welding with PWHT and coating for ITER test blanket modules and DEMO

EUROFER weldability is investigated in support of the European material properties database and TBM manufacturing. Electron Beam, Hybrid, laser and narrow gap TIG processes have been carried out on the EUROFER-97 steel (thickness up to 40 mm), a reduced activation ferritic-martensitic steel developed in Europe. These welding processes produce similar welding results with high joint coefficients and are well adapted for minimizing residual distortions. The fusion zones are typically composed of martensite laths, with small grain sizes. In the heat-affected zones, martensite grains contain carbide precipitates. High hardness values are measured in all these zones that if not tempered would degrade toughness and creep resistance. PWHT developments have driven to a one-step PWHT (750 °C/3 h), successfully applied to joints restoring good material performances. It will produce less distortion levels than a full austenitization PWHT process, not really applicable to a complex welded structure such as the TBM. Different tungsten coatings have been successfully processed on EUROFER material. It has shown no really effect on the EUROFER base material microstructure.     

The effect of neutron irradiation on mechanical properties of YAG laser weldments using previously irradiated material

The objective of this study is to make clear the effect of neutron irradiation on mechanical properties of laser weldments using irradiated material. This estimation is necessary for the application to joining coolant piping of the ITER blanket. Irradiation testing was performed at Japan Material Testing Reactor (JMTR). On the irradiation condition for weldments using irradiated material, fast neutron fluence was 1.4 × 10²⁴ n/m², which corresponds to a displacement damage rate of 0.26 displacement per atom (dpa) and irradiation temperature 200 °C. The results of this study show that tensile properties of all weldments changed into that of base material by the effect of neutron irradiation. The results of hardness tests show that irradiation hardening at an irradiation damage dose of 0.3 dpa is almost same as that at irradiation damage 0.6 dpa. It is concluded that irradiated weldments using irradiated material were moved toward irradiated base material on tensile and hardness properties up to 0.6 dpa. On the other hand, tensile properties of base material were changed by the effect of neutron irradiation up to about 0.3 dpa, and with much less change from 0.3 dpa to 0.6 dpa. It is inferred that the effect of neutron irradiation of SS316LN-IG almost saturated up to 0.3 dpa.