The properties of WWER-440 type reactor pressure vessel steels cut out from operated units

This paper presents the results of study on radiation degradation occurring in WWER-440 reactor pressure vessel (RPV) steel, using subsize impact specimens (5×5×27.5 mm³). The results of testing trepans and templates cut out from WWER-440 reactor pressure vessels are considered. Ductile-to-brittle transition temperatures (DBTT) obtained using standard Charpy and subsize impact specimens are compared. The relation between these two values is established.     

Thermal annealing as a method to predict results of high temperature irradiation embrittlement

In order to assess the validity of post-irradiation annealing as a method to predict results of high temperature irradiation a new analysis of experimental data has been performed revealing the combined influence of annealing temperature and impurities content on residual embrittlement after annealing. For 2CrMoV (WWER-440 reactor pressure vessel) steel with low contents of copper and phosphorus, the comparison of two embrittlement dependencies has been done: on irradiation temperature and post-irradiation annealing temperature. It is demonstrated that data for both the transition temperature shift after irradiation, ΔT_k, and the residual transition temperature shift after post-irradiation annealing, ΔT_res, fall within the same scatter band. A similarly close correlation is observed by comparison of yield strength increases after irradiation and after post-irradiation annealing.     

Defect formation and annealing kinetics in ion irradiated carbon nanotube buckypapers

We studied the frequency shifts in G, D and D^* Raman modes in freestanding multiwall carbon nanotube buckypapers. Upon ion irradiation by 140 keV He⁺ or 3 MeV H⁺ ions, the intensity ratio of D–G modes linearly increases with fluence before amorphization. The ratio is used to quantitatively measure the level of disorder in the buckypaper. The study shows that, upon post-irradiation annealing, defect removal requires little energy addition in lightly damaged buckypaper, which is evidenced by an activation energy of 0.36 eV. Once amorphized, defect removal becomes very difficult. The D–G intensity ratio has no reduction in heavily damage sample after annealing up to 850 °C.     

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 molybdenum neutron-irradiated at a high temperature

Molybdenum specimens prepared by two processes, powder-metallurgy (PM) and electron-beam melting (EB), were irradiated to a fast neutron fluence of 2.74 × 10²⁴n/m² (E_n? 1 MeV) at about 600°C (873 K), and their mechanical properties were studied in detail. It was shown that the degree of irradiation embrittlement in EB-Mo was smaller than that in PM-Mo, which might be caused by stronger grain-boundaries and probably smaller irradiation-hardening in the former. From the relation between the recovery of ductility and microstructural changes in post-irradiation annealed PM-Mo at 800 (1073 K), 1000 (1273 K) and 1200°C (1473 K), it was concluded that the recovery resulted from a decrease of irradiation hardening due to a rearrangement and a disappearance of depleted-zones, dislocation-loops and voids in order with increasing annealing temperature. An anomalous mode of fracture was observed in as-irradiated specimens, which consisted of inhomogeneous deformation, then brittle fracture not at the center but at the root of the deformation neck. This mode was observed in a narrow temperature range near the DBTT. A possible mechanism is discussed.     

Deformation and fracture properties in neutron irradiated pure Mo and Mo alloys

The effect of neutron irradiation on the mechanical properties of select molybdenum materials, unalloyed low carbon arc-cast (LCAC) Mo, Mo-0.5% Ti-0.1% Zr (TZM) alloy, and oxide dispersion-strengthened (ODS) Mo alloy, was characterized by analyzing the temperature dependence of mechanical properties. This study assembles the tensile test data obtained through multiple irradiation and post-irradiation experiments, in which tensile specimens were irradiated up to 13.1 dpa at 80-1000 °C and tested at −194 to 1000 °C. Irradiation at 80-609 °C increased yield stress significantly, up to 170%, while the increase of yield stress after irradiation at 784-936 °C was not significant. The plastic instability stress was strongly dependent on test temperature but was nearly independent of irradiation dose and temperature. The true fracture stress showed weak dependences on test temperature, irradiation dose and temperature when ductile failure occurred. Among the test materials the stress-relieved ODS material in the longitudinal direction (ODS-LSR) displayed the highest resistance to irradiation embrittlement due to its relatively high fracture stress. The critical temperature for shear failure (CTSF) was defined and evaluated for the test materials and the CTSF values were compared with the ductile-to-brittle transition temperatures (DBTT) based on ductility data.     

Results on research of templates from Kozloduy-1 reactor pressure vessel

The studies on the specimens manufactured from the templates cut out from the weld 4 of Kozloduy NPP Unit 1 reactor vessel have been conducted. The data on chemical composition of the weld metal have been obtained. Neutron fluence, mechanical properties, ductile to brittle transition temperature (DBTT) using mini Charpy samples have been determined. The phosphorus and copper content averaged over all templates is 0.046 and 0.1 wt.%, respectively. The fluence amounted up to 5×10¹⁸ n cm⁻² within 15–18 fuel cycles, and about 5×10¹⁹ n cm⁻² for the whole period of operation. These values agree well with calculated data. DBTT was determined after irradiation (T_k) to evaluate the vessel metal state at the present moment, then after heat treatment at the temperature of 475°C to simulate the vessel metal state after thermal annealing (T_an), and after heat treatment at 560°C to simulate the metal state in the initial state (T_k0). As a result of the tests the following values were obtained: T_k, +91.5°C; T_an, +63°C; and T_k0, 54°C. The values of T_k and T_an obtained by measurements were found to be considerably lower than those predicted in accordance with the conservative method accepted in Russia (177°C for T_k and 100°C for T_an). Thus, the obtained results allowed to make a conclusion that it is not necessary to anneal Kozloduy NPP Unit 1 reactor vessel for the second time. The fractographic and electron-microscopic research allowed to draw some conclusions on the embrittlement mechanism.     

Assessment of irradiation response of WWER-440 welds using samples taken from Novovoronezh unit 3 and 4 reactor pressure vessels

The results of the study on Novovoronezh unit 3 and 4 (NV NPP-3 and 4) reactor pressure vessel (RPV) radiation embrittlement measured using subsize impact specimens (5×5×27.5 mm³) fabricated from samples taken from the corresponding RPV walls are presented. The post-irradiation annealing effectiveness and the embrittlement kinetics of Novovoronezh unit 3 and 4 RPV welds under re-irradiation are discussed. Ductile-to-brittle transition temperatures (DBTT) obtained using standard Charpy (TT^10×10) and subsize impact (TT^5×5) specimens of trepans cut out from Novovoronezh unit 2 RPV are compared. A new relation between TT^10×10 and TT^5×5 has been developed.     

Fracture behavior of a neutron-irradiated stainless steel submerged arc weld cladding overlay

Stainless steel weld overlay cladding was irradiated at temperatures and fluences relevant to power reactor operation. The cladding was applied to a pressure vessel steel plate by the submerged arc, single-wire, oscillating-electrode method. Three layers of cladding were applied. The first layer was type 309, and the upper two layers were type 308 stainless steel. The type 309 was diluted considerably by excessive melting of the base plate. Charpy V-notch and tensile specimens were irradiated at 288°C to a fluence of 2 × 10²³ neutrons/m² (> 1 MeV).When irradiated, both types 308 and 309 cladding increased 5 to 40% in yield strength and slightly increased in ductility in the temperature range from 25 to 288°C. All cladding exhibited ductile-to-brittle transition behavior during impact testing caused by temperature dependent failure of the δ-ferrite phase. The type 308 cladding, microstructurally typical of that in reactor pressure vessels, showed very little degradation in either upper-shelf energy or transition temperature due to irradiation. Conversely, the impact properties of the specimens containing the highly diluted type 309 cladding, microstructurally similar to that produced during some off-normal welding conditions in existing reactors, experienced significant increases in transition temperature and drops of up to 50% in upper-shelf energy.     

Formation of Gas Pores in Nickel Alloys and Structural Steel under Irradiation by Helium Ions

Transmission electron microscopy is used to study the development of helium porosity in binary alloys of nickel with elements possessing a different dimensional atomic mismatch with nickel – from negative (beryllium and silicon) to positive (molybdenum, tungsten, aluminum, titanium, tantalum, tin, and zirconium), in structural steels ChS-68, ÉP-150, and the nickel alloy KhNM. The gas pores were produced by irradiation with 40 keV He⁺ up to fluence 5·10²⁰ m^–2 at 650 and 20°C followed by annealing at 650°C for 1 h. It is shown that under high-temperature annealing beryllium and silicon, relative to nickel, give rise to the formation of larger bubbles, while elements with a larger positive size mismatch with nickel atoms substantially decrease the size and increase the density of the bubbles. On the whole, as atomic radius and the concentration of the alloying element increases in alloys, the gas swelling of the irradiated layer decreases. Under post-irradiation annealing, bubbles with the largest diameter and the lowest density develop in nickel. Any alloying used decreases the size and increases the density of bubbles. The data obtained are discussed from the standpoint of the formation of various vacancy complexes of helium and their thermal stability.     

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.     

Nondestructive Magnetic Adaptive Testing of nuclear reactor pressure vessel steel degradation

Inspection of neutron-irradiation-generated degradation of nuclear reactor pressure vessel steel (RPVS) is a very important task. In ferromagnetic materials, such as RPVS, the structural degradation is connected with a change of their magnetic properties. In this work, applicability of a novel magnetic nondestructive method (Magnetic Adaptive Testing, MAT), based on systematic measurement and evaluation of minor magnetic hysteresis loops, is shown for inspection of neutron irradiation embrittlement in RPVS. Three series of samples, made of JRQ, 15CH2MFA and 10ChMFT type steels were measured by MAT. The samples were irradiated by E > 1 MeV energy neutrons with total neutron fluence of 1.58 × 10¹⁹–11.9 × 10¹⁹ n/cm². Regular correlation was found between the optimally chosen MAT degradation functions and the neutron fluence in all three types of the materials. Shift of the ductile–brittle transition temperature, ΔDBTT, independently determined as a function of the neutron fluence for the 15CH2MFA material, was also evaluated. A sensitive, linear correlation was found between the ΔDBTT and values of the relevant MAT degradation function. Based on these results, MAT is shown to be a promising (at least) complimentary tool of the destructive tests within the surveillance programs, which are presently used for inspection of neutron-irradiation-generated embrittlement of RPVS.     

Fluence rate effect semi-mechanistic modelling on WWER-type RPV welds

Effort at JRC-IE is ongoing in order to develop a semi-mechanistic model to forecast radiation embrittlement. The understanding and the quantification of the influence of the fluence rate is of particular importance for the correct interpretation of data obtained in material testing reactors or in surveillance capsules, which are accelerated with respect to embrittlement of the reactor pressure vessel wall itself. To verify the applicability of the fluence rate as included in the semi-mechanistic model and tuning the model parameters various WWER-type vessel weld material have been studied. For the selected welds, copper ranges from 0.08 to 0.18 mass%, while phosphorus variation is from 0.013 to 0.036 mass%. The fluence range is up to 2 × 10²⁰ n cm⁻² obtained at two fluence rates of 4 × 10¹¹ and 3.5 × 10¹² n cm⁻² s⁻¹, typical for WWER-440 surveillance positions. Significant fluence rate effect has been observed for the welds containing low copper and moderate phosphorus, and adaptation of the semi-mechanistic model’s parameters for the high flux data is required. To verify the consistency and the limits of the findings other similar data coming from RPV surveillance programmes are also included in this analysis.     

Investigation of the Effect of Radiation and High-Temperature Treatment on the x-Ray Luminescence of Quartz Glasses

The x-ray luminescence of KI, KV, and KU-1 quartz glasses, irradiated with and n– radiation in the dose range 10²–10⁷ Gy and neutron fluence range 10¹⁵–10¹⁷ cm^–2 and subjected to high-temperature annealing in air at 450 and 900°C is investigated. It is shown that the spectra of the nonirradiated and the and n– irradiated glasses of the first two types are a superposition of bands with _max = 410 and 460 nm, which are due to an impurity center initially present in the glasses (_max = 410 nm) and the initial and radiation-generated with dose 10⁶ Gy and fluence 10¹⁶ cm^–2 E' centers (_max = 460 nm). X-Ray luminescence is not observed in nonirradiated KU-1 glasses; a band with _max = 460–470 nm, due to radiation-generated E' centers, appears in the spectra of and n– irradiated glasses. As the radiation dose and the neutron fluence increase, the number of impurity centers decreases and the number of E' centers increases. It is established that the 410 nm band is due to the component of the n– radiation. High-temperature annealing in air at 900°C induces in the spectra new bands with _max = 470 and 520–540 nm, which are believed to be due to interstitial defects of the type O^– and O₂ ^–, formed when oxygen from air diffuses into the glass and localizes in interstices. 6 figures, 7 references.     

Strain rate effects in nuclear steels at room and higher temperatures

An investigation of strain rate, temperature and size effects in three nuclear steels has been conducted. The materials are: ferritic steel 20MnMoNi55 (vessel head), austenitic steel X6CrNiNb1810 (upper internal structure), and ferritic steel 26NiCrMo146 (bolting). Smooth cylindrical tensile specimens of three sizes have been tested at strain rates from 0.001 to 300 s⁻¹, at room and elevated temperatures (400–600 °C). Full stress–strain diagrams have been obtained, and additional parameters have been calculated based on them. The results demonstrate a clear influence of temperature, which amounts into reducing substantially mechanical strengths with respect to RT conditions. The effect of strain rate is also shown. It is observed that at RT the strain rate effect causes up shifting of the flow stress curves, whereas at the higher temperatures a mild downshifting of the flow curves is manifested. Size effect tendencies have also been observed. Some implications when assessing the pressure vessel structural integrity under severe accident conditions are considered.     

High velocity tensile test of austenitic stainless steel at elevated temperatures

This paper presents experimental results concerning the tensile properties of JIS Type SUS 316 stainless steel. The test was carried out at room temperature, 400°C and 550°C at strain rates of 10⁻³ 1/s and 10² 1/s. Base metal, weld joint and weld metal specimens were chosen for the test. The aim of this test is to clarify the effects of strain rate and test temperature on the mechanical properties such as 0.2% yield strength, ultimate tensile strength and elongation of JIS Type SUS 316 stainless steel.     

Ion-irradiation-induced athermal annealing of helium bubbles in SiC

We have compared the microstructural evolution of helium bubbles under ion irradiation and high temperature annealing. 4H-SiC was irradiated first by 140 keV He ions to a fluence of 1.0 × 10¹⁷ cm⁻² and then annealed at 1200 K for 30 min. Then, the samples were either irradiated by 2 MeV He ions to a fluence of 3.0 × 10¹⁶ cm⁻² at room temperature or annealed additionally at 1200 K for 30 min. Before and after 2 MeV He ion irradiation, significant microstructural changes were observed, similar to effects of high temperature annealing. Thus, the study provides evidence of ion-irradiation-induced athermal annealing on defect Ostwald ripening process and bubble evolution. Possible mechanisms are discussed.     

Fracture toughness of narrow-gap welded joints in the nuclear pressure vessel steel 22 NiMoCr 37

The mechanical testing of narrow-gap welded joints in 100 and 200 mm thick sections of the steel 22 NiMoCr 37 has revealed that the weld metal, and not the heat affected zone (HAZ) or the weld metal-parent metal boundary. is the critical region. This modified gas-shielded welding process operates with a very low heat input of the order of 6.500 J cm⁻¹ pass⁻¹ and the combination of small diameter welding wires and high welding speeds contributes to the excellent joint properties in the as-welded condition.To investigate the effect of preheating and post-welding heat treatment on the mechanical properties of narrow-gap welds, tensile, notch impact, flat bend and fracture toughness test specimens were extracted from joints welded with the following conditions: (1) no preheating: no post-weld heat treatment; (2) no preheating: soaking at 300°C: (3) no preheating: stress-relief heat treatment at 600°C; (4) preheating 200–250°C; no post-weld heat treatment; (5) preheating 200–250°C; soaking at 300°C; (6) preheating 200–250°C; stress relief heat treatment at 600°C. Tensile testing at room temperature and at 250°C of round specimens oriented across the seam revealed the ultimate fracture to be always located in the base material remote from the welded zone. Although pores or slag inclusions had an influence on bend-test results of specimens in the as-welded condition, the results generally show failure free bends to 180°C with no evidence of cracking in the HAZ or at the fusion boundary.Using sharp-notched impact bend specimens with the notch located in the centre of the seam as well as in and across the HAZ, absorbed energy-test temperature curves have been determined for each welding condition. In comparison with the base material impact toughness, the weld exhibits superior toughness in the temperature range − 60 – 0°C, but yielded lower values at room temperature. After stress relieving at 600°C, the impact toughness of the weld reduced significantly, apparently due to precipitations occurring in the weld-metal microstructure. Test results from welded specimens with the no notch in the HAZ show this region to have superior notch impact toughness to the base material.Crack opening displacement (COD) specimens 45 × 90 × 380 mm with the fatigue crack located in the weld metal and in the HAZ were tested at 0 and 20°C using both the recommendation in BS DD 19: 1972 as well as acoustic emission measurements for the determination of COD values. For this method of fracture toughness testing it has been shown that the occurrence of a critical event must be clearly defined as corresponding to stable crack growth or alternatively to unstable crack propagation.     

Recharging processes, radiation induced strain and changes of OH bands under H ion implantation in Ti doped lithium niobate

A systematic analysis of variations in structural and optical characteristics of Z-cut plates of titanium doped congruent lithium niobate single crystals implanted with 120 keV proton beam at various fluences of 10¹⁵, 10¹⁶ and 10¹⁷ protons/cm² is presented. Through, high resolution X-ray diffraction, atomic force microscopy, Fourier transform infrared and UV-visible-NIR analysis of congruent lithium niobate, the correlation of properties before and after implantation are discussed. HRXRD (0 0 6) reflection by Triple Crystal Mode shows that both tensile and compressive strain peak are produced by the high fluence implantation. A distinct tensile peak was observed from implanted region for a fluence of 10¹⁶ protons/cm². AFM micrographs indicate mountain ridges, bumps and protrusions on target surface on implantation. UV-visible-NIR spectra reveal an increase in charge transfer between Ti³⁺/Ti⁴⁺ and ligand oxygen for implantation with 10¹⁵ protons/cm², while spectra for higher fluence implanted samples show complex absorption band in the region from 380-1100 nm. Variations of OH⁻ stretching vibration mode were observed for cLN Pure, cLNT2% virgin, and implanted samples with FTIR spectra. The concentration of OH⁻ ion before and after implantation was calculated from integral absorption intensity. The effect of 120 keV proton implantation induced structural, surface and optical studies were correlated.