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.     

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.     

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.     

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.     

Corrosion behaviour of stainless steels in flowing LBE at low and high oxygen concentration

The corrosion behaviours of austenitic steel AISI 316L and martensitic steel T91 were investigated in flowing lead-bismuth eutectic (LBE) at 400 °C. The tests were performed in the LECOR and CHEOPE III loops, which stood for the low oxygen concentration and high oxygen concentration in LBE, respectively. The results obtained shows that steels were affected by dissolution at the condition of low oxygen concentration (C_[O2] = 10⁻⁸-10⁻¹⁰ wt%) and were oxidized at the condition of high oxygen concentration (C_[O2] = 10⁻⁵-10⁻⁶ wt%). The oxide layers detected are able to protect the steels from dissolution in LBE. Under the test condition adopted, the austenitic steel behaved more resistant to corrosion induced by LBE than the martensitic steel.     

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.     

Determination of neutron exposure of AISI 304 stainless steel from a BWR top guide using retrospective dosimetry

Retrospective dosimetry was used to determine the accumulated neutron exposure of AISI 304 stainless steel removed from the top guide of a boiling water reactor located at the Oyster Creek nuclear power station. The material was removed from areas adjacent to cracks that were observed after ∼20 years of operation. Using the plant operational history and a variety of measurements of various radioisotopes or non-radioactive transmutation products produced by irradiation, it was possible to determine the integrated neutron fluence experienced by the cracked region and to specify the accumulated displacement dose. Dose estimates on two separate specimens adjacent to the cracks were found to average 1.5 ± 0.2 dpa, possibly reflecting some uncertainty in measurement but more likely suggesting a small gradient in neutron flux-spectra within the section from which the various analysis specimens were cut. This report demonstrates that it is possible to examine defective components lying outside of the core region and where neutron flux-spectra are not well known, and to use the induced transmutation products to determine the neutron exposure with some confidence by using the examined specimen as its own dosimeter.     

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.     

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.     

Irradiation damage in 304 and 316 stainless steels: experimental investigation and modeling. Part I: Evolution of the microstructure

Irradiation damage in three austenitic stainless steels, SA 304L, CW 316 and CW Ti-modified 316, is investigated both experimentally and theoretically. The density and size of Frank loops after irradiation at 320 and 375 °C in experimental EBR II, BOR-60 and OSIRIS reactors for doses up to 40 dpa are characterized by TEM. The evolution of the initial dislocation network under irradiation is evaluated. A cluster dynamics model is proposed to account quantitatively for the experimental findings.     

Swelling rate versus swelling correlation in 20% cold-worked 316 stainless steels

A quantitative correlation of the swelling rate versus swelling is presented. Swelling data of 20% cold-worked 316 stainless steels were analyzed using the power law swelling equation. The prolonged transient region with keeping the suppressed swelling rate was clearly demonstrated for the improved 316 stainless steels like PNC316 and 316Ti. Rate theory analyses lead to the role of precipitates as point defect sinks for retardation of the void growth.     

Detection of intergranular cracking susceptibility due to hydrogen in irradiated austenitic stainless steel with a superconducting quantum interference device (SQUID) sensor

To investigate the detection method of intergranular (IG) cracking susceptibility by hydrogen in irradiated austenitic stainless steel (SS), magnetic and mechanical properties were examined after two repeats of hydrogen charging and discharging (hydrogen treatment) in Type 304 SS which had been irradiated during use in different reactor cores. The residual magnetic flux density (Br) was measured with a superconducting quantum interference device sensor and Br increased with increased neutron fluence and repeated hydrogen treatments. Elongation decreased with an increase of Br and IG cracking appeared above Br of 2×10⁻⁵ T for this measuring method after repeated hydrogen treatments. These phenomena would be caused by hydrogen-induced martensite phase being formed on grain boundaries. It was thought the appearance of IG cracking susceptibility due to hydrogen in irradiated SS could be predicted by measuring the Br of the steel.     

Dissimilar welding of nickel-based Alloy 690 to SUS 304L with Ti addition

This study investigates the effects of Ti addition on the weldability, microstructure and mechanical properties of a dissimilar weldment of Alloy 690 and SUS 304L. Shielding metal arc welding (SMAW) is employed to butt-weld two plates with three welding layers, where each layer is deposited in a single pass. To investigate the effects of Ti addition, the flux coatings of the electrodes used in the welding process are modified by varying additions of either a Ti-Fe compound or a Ti powder. The results indicate that the microstructure of the fusion zone (FZ) is primarily dendritic. With increasing Ti content, it is noted that the microstructure changes from a columnar dendritic to an equiaxed dendritic, in which the primary dendrite arm spacing (PDAS) becomes shorter. Furthermore, it is observed that the amount of Al-Ti oxide phase increases in the inter-dendritic region, while the amount of Nb-rich phase decreases. Moreover, the average hardness of the FZ increases slightly. The results indicate that Ti addition prompts a significant increase in the elongation of the weldment (i.e. 36.5%, Ti: 0.41 wt%), although the tensile strength remains relatively unchanged. However, at an increased Ti content of 0.91 wt%, an obvious reduction in the tensile strength is noted, which can be attributed to a general reduction in the weldability of the joint.     

Behavior of steels in flowing liquid PbBi eutectic alloy at 420-600 °C after 4000-7200 h

This paper presents the results of steel exposure up to 7200 h in flowing LBE at elevated temperatures and is a follow-up paper of that with results of an exposure of up to 2000 h. The examined AISI 316 L, 1.4970 austenitic and MANET 10Cr martensitic steels are suitable as a structural material in LBE (liquid eutectic Pb₄₅Bi₅₅) up to 550 °C, if 10⁻⁶ wt% of oxygen is dissolved in the LBE. The martensitic steel develops a thick magnetite and spinel layer while the austenites have thin spinel surface layers at 420 °C and thick oxide scales like the martensitic steel at 550 °C. The oxide scales protect the steels from dissolution attack by LBE during the whole test period of 7200 h. Oxide scales that spall off are replaced by new protective ones. At 600 °C severe attack occurs already after 2000 and 4000 h of exposure. Steels with 8-15 wt% Al alloyed into the surface suffer no corrosion attack at all experimental temperatures and exposure times.     

Impact properties of 304L stainless steel GTAW joints evaluated by high strain rate of compression tests

This paper presents an investigation into the high velocity impact of 304L stainless steel gas tungsten arc welded (GTAW) joints at strain rates between 10⁻³ and 7.5 × 10³ s⁻¹ using a compressive split-Hopkinson bar. The results show that the impact properties and fracture characteristics of the tested weldments depend strongly on applied strain rate. This rate-dependent behavior is in good agreement with model predictions using the hybrid Zerilli-Armstrong constitutive law. It is determined that the tested weldments fail as a result of adiabatic shearing. The fracture surfaces of the fusion zone and base metal regions are characterized by the presence of elongated dimples. The variation in the observed dimple features with strain rate is consistent with the results of the impact stress-strain curves.     

TEM and PAS study of neutron irradiated VVER-type RPV steels

Conventional transmission electron microscopy and positron lifetime and Doppler broadening positron annihilation spectroscopy techniques have been used to investigate the radiation-induced microstructural changes in surveillance specimens of VVER-type reactor pressure vessel (RPV) steels, and RPV steels irradiated in the research reactor. Defects visible in transmission electron microscopy consist of black dots, dislocation loops and precipitates concentrated along the dislocation substructure. Their size and density depend on the neutron flux and fluence. The parallel set of thermally aged specimens, specimens recovery annealed after irradiation and specimens irradiated in a lower neutron flux was investigated too. No defects discernible in transmission electron microscopy were found after accelerated irradiation in the research reactor. In addition to visible defects, the small-volume vacancy clusters were identified by positron annihilation spectroscopy.     

The neutron irradiation effect on mechanical properties of HIP joint material

Dispersion strengthened copper (DSCu) and stainless steel are the candidate material for the heat sink and the structural material of the ITER shielding blanket and these materials are joined by hot isostatic pressing (HIP). In this study, the neutron irradiation effect on mechanical properties of HIP joint material was examined by tensile and impact tests using specimens with irradiation damage of about 1.5 dpa. The results of tensile tests show that tensile strength of HIP joint material was about the same as that of DSCu base material, and this trend did not change after neutron irradiation. On the other hand, the impact value of HIP joint material was smaller than that of DSCu base material because of the diffusion of main elements at joint boundary. It was shown that embrittlement by the neutron irradiation effect is smaller than that of the effect by HIP joint.     

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.     

Development of a fundamental crack tip strain rate equation and its application to quantitative prediction of stress corrosion cracking of stainless steels in high temperature oxygenated water

A formulation for the quantitative calculation of the stress corrosion cracking (SCC) growth rate was proposed based on a fundamental-based crack tip strain rate (CTSR) equation that was derived from the time-based mathematical derivation of a continuum mechanics equation. The CTSR equation includes an uncertain parameter r₀, the characteristic distance away from a growing crack tip, at which a representative strain rate should be defined. In this research, slow strain rate tensile tests on sensitized 304L stainless steel in oxygenated high temperature water were performed. By curve fitting the experimental results to the numerically calculated crack growth rate, the parameter r₀ was determined. Then, the theoretical formulation was used to predict the SCC growth rates. The results indicate that r₀ is on the order of several micrometers, and that the application of the theoretical equation in predicting the crack growth rate provides satisfactory agreement with the available data.     

Strain rate dependence of impact properties of sintered 316L stainless steel

This paper uses a material testing system (MTS) and a compressive split-Hopkinson bar to investigate the impact behaviour of sintered 316L stainless steel at strain rates ranging from 10⁻³ s⁻¹ to 7.5 × 10³ s⁻¹. It is found that the true stress, the rate of work hardening and the strain rate sensitivity vary significantly as the strain rate increases. The flow behaviour of the sintered 316L stainless steel can be accurately predicted using a constitutive law based on Gurson’s yield criterion and the flow rule proposed by Khan, Huang and Liang (KHL). Microstructural observations reveal that the degree of localized grain deformation increases, but the pore density and the grain size decrease, with increasing strain rate. Adiabatic shear bands associated with cracking are developed at strain rates higher than 5.6 × 10³ s⁻¹. The fracture surfaces exhibit ductile dimples. The depth and density of these dimples decrease with increasing strain rate.