Microstructural development and radiation hardening of neutron irradiated Mo-Re alloys

Stress-relieved specimens and recrystallized specimens of pure Mo and Mo-Re alloys with Re contents of 2, 4, 5, 10, 13 and 41 wt% were neutron irradiated up to 20 dpa at temperatures from 681 to 1072 K. On microstructural observation, sigma phase and chi phase precipitates were found in all irradiated Mo-Re alloys. Voids were observed in all irradiated specimens, and dislocation loops and dislocations were observed in the specimens that were irradiated at lower temperatures. On Vickers hardness testing, all of the irradiated specimens showed hardening. Especially Mo-41Re were drastically embrittled after irradiation at 874 K or below. From these results, the authors discuss about the relation between microstructure development and radiation hardening and embrittlement, and propose the optimum Re content and thermal treatment for Mo-Re alloys to be used under irradiation conditions.     

Welding-induced microstructure in austenitic stainless steels before and after neutron irradiation

The effects of neutron irradiation on the microstructure of welded joints made of austenitic stainless steels have been investigated. The materials were welded AISI 304 and AISI 347, so-called test weld materials, and irradiated with neutrons at 300 °C to 0.3 and 1.0 dpa. In addition, an AISI 304 type from a decommissioned pressurised water reactor, so-called in-service material, which had accumulated a maximum dose of 0.35 dpa at about 300 °C, was investigated. The microstructure of heat-affected zones and base materials was analysed before and after irradiation, using transmission electron microscopy. Neutron diffraction was performed for internal stress measurements. It was found that the heat-affected zone contains, relative to the base material, a higher dislocation density, which relates well to a higher residual stress level and, after irradiation, a higher irradiation-induced defect density. In both materials, the irradiation-induced defects are of the same type, consisting in black dots and Frank dislocation loops. Careful analysis of the irradiation-induced defect contrast was performed and it is explained why no stacking fault tetrahedra could be identified.     

Analysis of WWER-440 and PWR RPV welds surveillance data to compare irradiation damage evolution

It is known that for Russian-type and Western water reactor pressure vessel steels there is a similar degradation in mechanical properties during equivalent neutron irradiation. Available surveillance results from WWER and PWR vessels are used in this article to compare irradiation damage evolution for the different reactor pressure vessel welds. The analysis is done through the semi-mechanistic model for radiation embrittlement developed by JRC-IE. Consistency analysis with BWR vessel materials and model alloys has also been performed within this study. Globally the two families of studied materials follow similar trends regarding the evolution of irradiation damage. Moreover in the high fluence range typical of operation of WWER the radiation stability of these vessels is greater than the foreseen one for PWR.     

Correlation of yield stress and microhardness in 08Cr16Ni11Mo3 stainless steel irradiated to high dose in the BN-350 fast reactor

The relationship between the microhardness and the engineering yield stress in 08Cr16Ni11Mo3 steel after irradiation in the BN-350 reactor has been experimentally derived and agrees with a previously published correlation developed by Toloczko for unirradiated 316 in a variety of cold-work conditions. Even more importantly, when the correlation is derived in the K_Δ format where the correlation involves changes in the two properties, excellent agreement is found with a universal K_Δ correlation developed by Busby and coworkers. Additionally, this report points out that microhardness measurements must take into account that sodium exposure at high temperature and neutron fluence alters the metal surface to produce ferrite, and therefore the altered layers should be removed prior to testing.     

Welding-induced mechanical properties in austenitic stainless steels before and after neutron irradiation

The effects of neutron irradiation on the mechanical properties of welded joints made of austenitic stainless steels have been investigated. The materials are welded AISI 304 and AISI 347, so-called test weld materials, irradiated with neutrons at 573 K to doses of 0.3 and 1.0 dpa. In addition, an AISI 304 from a decommissioned pressurised water reactor, so-called in-service material, which had accumulated a maximum dose of 0.35 dpa at about 573 K, was investigated. The mechanical properties of heat-affected zones and base materials were analysed before and after irradiation. Tensile parameters were determined at room temperature and at 573 K, for all materials and irradiation conditions. In the test weld materials it is found that radiation hardening is lower and loss of ductility is higher in the heat-affected zone than in the base material. In the in-service material radiation hardening is about the same in heat-affected zone and base material. After irradiation, deformation takes place by stacking faults and twins, at both room temperature and high temperature, contrary to unirradiated materials, where deformation takes place by twinning at room temperature and by dislocation cells at high temperature. No defect free channels are observed.     

Bootstrap calculation of ultimate strength temperature maxima for neutron irradiated ferritic/martensitic steels

The dependence of mechanical properties of ferritic/martensitic (F/M) steels on irradiation temperature is of interest because these steels are used as structural materials for fast, fusion reactors and accelerator driven systems. Experimental data demonstrating temperature peaks in physical and mechanical properties of neutron irradiated pure iron, nickel, vanadium, and austenitic stainless steels are available in the literature. A lack of such an information for F/M steels forces one to apply a computational mathematical-statistical modeling methods. The bootstrap procedure is one of such methods that allows us to obtain the necessary statistical characteristics using only a sample of limited size. In the present work this procedure is used for modeling the frequency distribution histograms of ultimate strength temperature peaks in pure iron and Russian F/M steels EP-450 and EP-823. Results of fitting the sums of Lorentz or Gauss functions to the calculated distributions are presented. It is concluded that there are two temperature (at 360 and 390 °C) peaks of the ultimate strength in EP-450 steel and single peak at 390 °C in EP-823.     

Study of radiation-induced degradation of RPV steels and model alloys by positron annihilation and Mössbauer spectroscopy

The influence of different microstructural processes on the degradation due to radiation embrittlement has studied by positron annihilation and Mössbauer spectroscopy. The materials studied consisted of WWER-440 base (15Kh2MFA) and weld (10KhMFT) RPV steels which were neutron-irradiated at fluence levels of 0.78 × 10²⁴ m⁻², 1.47 × 10²⁴ m⁻² and 2.54 × 10²⁴ m⁻²; WWER-1000 base (15Kh2NMFAA) and weld (12Kh2N2MAA) irradiated at a fluence level 1.12 × 10²⁴ m⁻²; three different model alloys implanted with protons at two dose levels (up to 0.026 dpa), finally the base metal of WWER-1000 (15Kh2NMFAA) was thermally treated with the intention to simulate the P-segregation process. It has been shown possible to correlate the values of parameters obtained by such techniques and data of mechanical testing (ductile-to-brittle transition temperature and upper shelf energy).     

Evolution of fine-scale defects in stainless steels neutron-irradiated at 275 °C

Six austenitic stainless steel heats (three heats each of 304SS and 316SS) neutron-irradiated at 275 °C from 0.6 to 13.3 dpa have been carefully characterized by TEM and their hardness measured as a function of dose. The characterization revealed that the microstructure is dominated by a very high density of small Frank loops present in sizes as small as 1 nm and perhaps lower, which could be of both vacancy and interstitial-type. Frank loop density saturated at the lowest doses characterized, whereas the Frank loop size distributions changed with increasing dose from an initially narrow, symmetric shape to a broader, asymmetric shape. Although substantial hardening is caused by the small defects, a simple correlation between hardness changes and density and size of defects does not exist. These results indicate that radiation-induced segregation to the Frank loops could play a role in both defect evolution and hardening response.     

Correlation between irradiation-induced changes of microstructural parameters and mechanical properties of RPV steels

Radiation hardening, displayed by the yield stress increase, and irradiation embrittlement, described by the Charpy transition temperature shift, were experimentally determined for a broad variety of irradiation specimens machined from different reactor pressure vessel base and weld materials and irradiated in several VVER-type reactors. Additionally, the same specimens were investigated by small angle neutron scattering. The analysis of the neutron scattering data suggests the presence of nano-scaled irradiation defects. The volume fraction of these defects depends on the neutron fluence and the material. Both irradiation hardening and irradiation embrittlement correlate linearly with the square root of the defect volume fraction. However, a generally valid proportionality is only a rough approximation. In detail, chemical composition and technological pretreatment clearly affect the correlation.     

Evidence for neutron irradiation-induced metallic precipitates in model alloys and pressure-vessel weld steel

We have used positron Doppler-broadening spectroscopy to examine a series of neutron-irradiated model alloys (1 × 10²³ n/m², E>0.5 MeV) and 73W-weld steel (to 1.8 × 10²³ n/m², E>1 MeV. The copper, nickel and phosphorus content of the model alloys was systematically varied. The samples were examined in the as-irradiated state and after post-irradiation isochronal anneals to temperature up to 600 °C. By following the S and W parameters, and especially by plotting the results in (S,W) space, we can infer that the damage is a combination of irradiation-induced metallic precipitates and vacancy-type defect clusters. Samples with either high Cu or with a combination of high Ni and medium Cu (and the pressure-vessel weld steel) showed evidence for both irradiation-induced metallic precipitation, and vacancy-type clusters, while samples without either high Ni or high Cu showed predominantly evidence of annihilations in vacancy-type clusters. These results are discussed in terms of embrittlement models.     

Behavior of thorium-uranium (IV) phosphate-diphosphate sintered samples during leaching tests. Part I - Kinetic study

The dissolution of β-TUPD sintered samples was examined in various conditions of pH, temperature, concentrations of anions in the leachate and leaching flow rates. All the normalized dissolution rates were in the range 10⁻⁷ to 10⁻⁴ g m⁻² day⁻¹ even in very aggressive media, showing the good resistance of these ceramics to aqueous alteration. The first part of this paper describes several parameters exhibiting a significant influence on the normalized dissolution rate of the pellets prepared. Both the partial order relative to the proton concentration (n = 0.39-0.41) and the apparent activation energy (E_app = 49 kJ mol⁻¹) were found in good agreement with the data reported for powdered samples showing that the sintering process does not degrade the chemical durability of the ceramics. Moreover, due to the high thermodynamical constant of complexation of phosphate species for tetravalent uranium and thorium, the influence of other ligands such as nitrate, chloride or sulphate on the normalized dissolution rates was limited. Near the equilibrium, the increasing of the leaching time, the temperature or the leachate acidity led to the thorium precipitation at the surface of the pellets either in static or in dynamic conditions. Consequently, the dissolution became clearly incongruent and controlled by saturation processes which are described in the second part of this paper.     

Irradiation damage in 304 and 316 stainless steels: experimental investigation and modeling. Part II: Irradiation induced hardening

The consequences of irradiation damage in austenitic stainless steels on their mechanical properties, namely the yield stress, are investigated both experimentally and theoretically. The observed hardening is correlated with the quantitative characteristics of irradiation defects population. A simple model allowing for the defaulting of Frank loops under stress predicts the hardening and its saturation at large doses.     

Neural-network analysis of irradiation hardening in low-activation steels

An artificial neural network has been used to model the irradiation hardening of low-activation ferritic/martensitic steels. The data used to create the model span a range of displacement damage of 0-90 dpa, within a temperature range of 273-973 K and contain 1800 points. The trained model has been able to capture the non-linear dependence of yield strength on the chemical composition and irradiation parameters. The ability of the model to generalise on unseen data has been tested and regions within the input domain that are sparsely populated have been identified. These are the regions where future experiments could be focused. It is shown that this method of analysis, because of its ability to capture complex relationships between the many variables, could help in the design of maximally informative experiments on materials in future irradiation test facilities. This will accelerate the acquisition of the key missing knowledge to assist the materials choices in a future fusion power plant.     

Characteristics of microscopic strain localization in irradiated 316 stainless steels and pure vanadium

Characteristics of localized dislocation glide were investigated for 316 and 316LN stainless steels and pure vanadium after ion or neutron irradiation near room temperature and deformation by a uniaxial tensile load or by a multiaxial bending load. In the irradiated 316 stainless steels, both the uniaxial tensile loading and the multiaxial bend loading produced straight localized bands in the form of channels and twins. In vanadium specimens, on the other hand, curved channels were observed after tensile deformation, and these became a common feature after multiaxial bend deformation. No twin was observed in vanadium. A river pattern of channels was observed in the bent samples after irradiation to a high dose of 0.69 dpa. A highly curved channel can be formed by successive cross slip of screw dislocations. Also, the channel width was not constant along the channels; channel widening occurred when weak defect clusters were removed by the gliding screw dislocations changing their paths by cross slip. It is believed that the dissociation of dislocations into partials and high angles between easy glide planes suppresses the formation of curved channels, while a multiaxial stress state, or a higher stress constraint, increases the tendency for channel bending and widening.     

Microstructural characteristics and embrittlement phenomena in neutron irradiated 309L stainless steel RPV clad

The effects of neutron irradiation on the microstructural features and mechanical properties of 309L stainless steel RPV clad were investigated using TEM, SEM, small tensile, microhardness and small punch (SP) tests. The neutron irradiations were performed at 290 °C up to the fluences of 5.1 × 10¹⁸ and 1.02 × 10¹⁹ n/cm² (>1 MeV) in Japan Materials Testing Reactor (JMTR). The microstructure of the clad before and after irradiation was composed of main part of fcc austenite, a few percent of bcc δ-ferrite and small amount of brittle σ phase. After irradiation, not only the yield stress and microhardness, but SP ductile to brittle transition temperature (SP-DBTT) were increased. However, the increase in SP-DBTT is almost saturated, independent of the neutron fluence. Based on the TEM observation, the origin of irradiation hardening was accounted for by the irradiation-produced defect clusters of invisible fine size (<1-2 nm), and the shift of SP-DBTT was primary due to the higher hardening and the preferential failure of δ-ferrite. The embrittlement of the clad was strongly affected by the initial microstructural factors, such as the amount of brittle σ phase, which caused a cracking even in an early stage of deformation.     

Behavior of thorium-uranium (IV) phosphate-diphosphate sintered samples during leaching tests. Part II. Saturation processes

Sintered pellets of thorium-uranium (IV) phosphate-diphosphate solid solutions (β-Th_4−xU_x(PO₄)₄P₂O₇, β-TUPD) were altered in several acidic media. All the results reported in the first part of this paper confirmed the good chemical durability of the samples. The evolution of the normalized weight loss showed that, in several media, thorium quickly precipitates in a neoformed phosphate-based phase while uranium (IV) is released in the leachate due to its oxidation into the uranyl form. The characterization of neoformed phases was carried out through several techniques involving grazing XRD, infrared and μ-Raman spectroscopies, EPMA, SEM and TEM. SEM micrographies showed that the dissolution mainly occurs at the grain boundaries, leading to the break away of the grains: only the first 15 μm are altered for 2 months in 10⁻¹ M HNO₃. From EPMA and BET measurements, neither the chemical composition nor the specific surface area are significantly modified. Near equilibrium, two neoformed phases were observed and identified by grazing XRD and/or μ-Raman spectroscopy at the surface of the leached pellets: one is found to be amorphous and progressively turns into the crystallized thorium phosphate-hydrogenphosphate hydrate (TPHPH). From the results obtained, a chemical scheme of the dissolution of β-TUPD sintered samples is proposed. The behavior of the actinides in the gelatinous phase appears mainly driven by their oxidation state: thorium remains in the tetrapositive state and is quickly and quantitatively precipitated while uranium (IV) is oxidized into uranyl then released in the leachate. The Th-precipitation as TPHPH first appears scattered then covers the entire surface of the pellet, inducing a delay of the actinides release in the leachate. Both phases act as protective layers and should induce the significant delay of the release of actinides (Th, U) to the biosphere.     

Corrosion behaviour of aluminized martensitic and austenitic steels in liquid Pb-Bi

The Pb-Bi liquid alloy is under consideration as a spallation target material in the hybrid systems due to its suitable nuclear and physical properties. In order to limit the risks of corrosion of the structural elements in contact with the liquid Pb-Bi, protection by means of aluminized coatings was investigated for 316L austenitic steel and T91 martensitic steel. For both steels, no damages were observed after immersions in static Pb-Bi up to 500 °C for low oxygen concentrations and long durations. However, at 600 °C in the same conditions, a non-uniform degradation of the coatings was observed. Only coated 316L was tested in dynamic conditions. The results were generally satisfying for temperatures from 350 to 600 °C and for fluid velocities up to 2.3 m s⁻¹. However, in both the IPPE loops and the CICLAD device, some localized damage of the coatings, attributed to erosion, was observed.     

Change in flow stress and ductility of δ-phase Pu-Ga alloys due to self-irradiation damage

An internal state variable model for the mechanical behavior of aged Pu-Ga alloys is developed and used to predict the change of the material with accumulated self-irradiation damage or age. The material model incorporates microstructural data such as the primary irradiation-induced defect density from cascades, the density and average size of helium bubbles, the initial dislocation density, and the initial average segment length of the dislocation density as input parameters, and then evaluates the stress-strain response of a representative volume element of the material. Given this response at a single material point, the deformation behavior of tensile specimens is predicted, and it forecasts increased strength, decreased strain hardening, and more strain localization with aging. Although the material point behavior showed some slight strain softening, this strain softening is expected to be masked by statistical variations of different volume elements and by the strain rate sensitivity of the material. Hence, it is not expected to appear in the stress-strain response of macroscopic tensile specimens, and only the increase in flow stress will be measured.     

Influence of irradiation temperature and dose gradients on the microstructural evolution in neutron-irradiated 316SS

A cold worked 316SS baffle bolt was extracted from the Tihange pressurized water reactor and sectioned at three different positions. The temperature and dose at the 1-mm bolt head position were 593 K and 19.5 dpa respectively, whereas at two shank positions the temperature and dose was 616 K and 12.2 dpa at the 25-mm position and 606 K and 7.5 dpa at the 55-mm position. Microstructural characterization revealed that small faulted dislocation loops and cavities were visible at each position, but the cavities were most prominent at the two shank positions. Measurable swelling exists in the shank portions of this particular bolt, and accompanying this swelling is the retention of very high levels of hydrogen absorbed from the environment. The observation of cavities in the CW 316SS at temperatures and doses relevant to LWR conditions has important implications for pressurized water reactors since SA 304SS plates surround the bolts, a steel that usually swells earlier due to its lower incubation period for swelling.     

Change of fuel-to-cladding gap width with the burn-up in FBR MOX fuel irradiated to high burn-up

In order to study the dependence of the gap width change on the burn-up, the fuel-to-cladding gap widths were investigated by ceramography in a large number of FBR MOX fuel pins irradiated to high burn-up. The dependence of gap widths on the burn-up was closely connected with the formations of JOG (joint oxyde-gaine) and rim structure. The gap widths decreased gradually due to the fuel swelling until ∼30 GWd/t, but beyond this burn-up the dependence showed two different tendencies. With the increase of burn-up, the gap widths decreased due to the increase of fuel swelling in the low fuel temperature region where the rim structure was observed, but they increased in the high fuel temperature region where the JOG rich in Cs and Mo formed in the gap.