Local mechanical properties of Alloy 82/182 dissimilar weld joint between SA508 Gr.1a and F316 SS at RT and 320 °C

The distributions of mechanical and microstructural properties were investigated for the dissimilar metal weld joints between SA508 Gr.1a ferritic steel and F316 austenitic stainless steel with Alloy 82/182 filler metal using small-size tensile specimens. The material properties varied significantly in different zones while those were relatively uniform within each material. In particular, significant gradient of the mechanical properties were observed near the both heat-affected zones (HAZs) of F316 SS and SA508 Gr.1a. Thus, the yield stress (YS) was under-matched with respect to the both HAZs, although, the YS of the weld metal was over-matched with respect to both base metals. The minimum ductility occurred in the HAZ of SA508 Gr.1a at both test temperatures. The plastic instability stress also varied considerably across the weld joints, with minimum values occurring in the SA508 Gr.1a base metal at RT and in the HAZ of F316 SS at 320 °C. The transmission electron micrographs showed that the strengthening in the HAZ of F316 SS was attributed to the strain hardening, induced by a strain mismatch between the weldment and the base metal, which was evidenced by high dislocation density in the HAZ of F316 SS.     

Temperature effects on the mechanical properties of candidate SNS target container materials after proton and neutron irradiation

This report presents the tensile properties of EC316LN austenitic stainless steel and 9Cr-2WVTa ferritic/martensitic steel after 800 MeV proton and spallation neutron irradiation to doses in the range 0.54-2.53 dpa at 30-100 °C. Tensile testing was performed at room temperature (20 °C) and 164 °C. The EC316LN stainless steel maintained notable strain-hardening capability after irradiation, while the 9Cr-2WVTa ferritic/martensitic steel posted negative hardening in the engineering stress-strain curves. In the EC316LN stainless steel, increasing the test temperature from 20 to 164 °C decreased the strength by 13-18% and the ductility by 8-36%. The effect of test temperature for the 9Cr-2WVTa ferritic/martensitic steel was less significant than for the EC316LN stainless steel. In addition, strain-hardening behaviors were analyzed for EC316LN and 316L stainless steels. The strain-hardening rate of the 316 stainless steels was largely dependent on test temperature. A calculation using reduction of area measurements and stress-strain data predicted positive strain hardening during plastic instability.     

Effects of Dissolved Hydrogen and Strain Rate on IASCC Behavior in Highly Irradiated Stainless Steels

IASCC behavior in cold-worked SUS316 stainless steels irradiated to 35 dpa was examined using slow strain rate tensile testing at a strain rate of 6.7 × 10^?8/s in 320°C simulated PWR primary water while varying the dissolved hydrogen (DH) concentration from 0 to 2.8 ppm. The results were compared with those previously obtained at a higher strain rate using specimens of different sizes and with those of the previous interrupted experiment. The initiation and propagation of IASCC enhanced with increasing DH concentration and lower strain rate. The IASCC initiation stress decreased to almost half of the yield strength at high DH. Accompanying slow tensile tests in an argon gas environment showed that a lower strain rate did not change in the initiation stress that exceeded the yield strength, but enhanced the propagation of intergranular cracking.     

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.     

The effect of dose rate on the response of austenitic stainless steels to neutron radiation

Depending on reactor design and component location, austenitic stainless steels may experience significantly different irradiation dose rates in the same reactor. Understanding the effect of dose rate on radiation performance is important to predicting component lifetime. This study examined the effect of dose rate on swelling, grain boundary segregation, and tensile properties in austenitic stainless steels through the examination of components retrieved from the Experimental Breeder Reactor-II (EBR-II) following its shutdown. Annealed 304 stainless steel, stress-relieved 304 stainless steel, 12% cold-worked 316 stainless steel, and 20% cold-worked 316 stainless steel were irradiated over a dose range of 1–56 dpa at temperatures from 371 to 440 °C and dose rates from 0.5 to 5.8 × 10⁻⁷ dpa/s. Density and tensile properties were measured for 304 and 316 stainless steel. Changes in grain boundary composition were examined for 304 stainless steel. Swelling appears to increase at lower dose rates in both 304 and 316 stainless steel, although the effect was not always statistically significant. Grain boundary segregation also appears to increase at lower dose rate in 304 stainless steel. For the range of dose rates examined, no measurable dose rate effect on tensile properties was noted for any of the steels.     

Investigation of the ductile fracture properties of type 304 stainless steel plate, welds, and 4-inch pipe

J-integral fracture toughness tests were performed on welded 304 stainless steel 2-inch plate and 4-inch diameter pipe. The 2-inch plate was welded using a hot-wire automatic gas tungsten arc process. This weldment was machined into 1T and 2T compact specimens for single specimen unloading compliance J-integral tests. The specimens were cut to measure the fracure toughness of the base metal, weld metal and the heat affected zone (HAZ). The tests were performed at 550°F, 300°F and room temperature. The results of the J-integral tests indicate that the J_Ic of the base plate ranged from 4400 to 6100 in lbs/in² at 550°F. The J_Ic values for the tests performed at 300°F and room temperature were beyond the measurement capacity of the specimens and appear to indicate that J_Ic was greater than 8000 in lb/in². The J-integral tests performed on the weld metal specimens indicate that the J_Ic values ranged from 930 to 2150 in lbs/in² at 550°F. The J_Ic values of the weld metal specimens tested at 300°F and room temperature were 2300 and 3000 in lbs/in² respectively. One HAZ specimen was tested at 550°F and found to have a J_Ic value of 2980 in lbs/in² which indicates that the HAZ is an average of the base metal and weld metal thoughness. These test results indicate that there is a significant reduction in the initiation fracture toughness as a result of welding.The second phase of this task dealt with the fracture toughness testing of 4-inch diameter 304 stainless steel pipes containing a gas tungsten arc weld. The pipes were tested at 550°F in four point bending. Three tests were performed, two with a through wall flaw growing circumferentially and the third pipe had a part through radial flaw in combination with the circumferential flaw. These tests were performed using unloading compliance and d.c. potential drop crack length estimate methods. The results of these test indicate that the presence of a complex crack (radial and circumferential) reduces in the initiation toughness and the tearing modulus of the pipe material compared to a pipe with only a circumferentially growing crack.     

Creep rupture strength of activated-TIG welded 316L(N) stainless steel

316L(N) stainless steel plates were joined using activated-tungsten inert gas (A-TIG) welding and conventional TIG welding process. Creep rupture behavior of 316L(N) base metal, and weld joints made by A-TIG and conventional TIG welding process were investigated at 923 K over a stress range of 160-280 MPa. Creep test results showed that the enhancement in creep rupture strength of weld joint fabricated by A-TIG welding process over conventional TIG welding process. Both the weld joints fractured in the weld metal. Microstructural observation showed lower δ-ferrite content, alignment of columnar grain with δ-ferrite along applied stress direction and less strength disparity between columnar and equiaxed grains of weld metal in A-TIG joint than in MP-TIG joint. These had been attributed to initiate less creep cavitation in weld metal of A-TIG joint leading to improvement in creep rupture strength.     

Structural design of high temperature metallic components

In the high temperature engineering test reactor (HTTR), even at normal operation the service temperatures of class 1 metallic components reach temperatures above 900 °C when exposed to primary helium coolant of 950 °C. For these components, Hastelloy XR, which is the improved version of Hastelloy X, was developed and used for high temperature application.Some of the high temperature materials and their service temperatures, including Hastelloy XR, used for the class 1 and reactor internal metallic components of the HTTR are very well beyond the well-established Japanese elevated temperature structural design guideline. Moreover, at very high temperatures, where creep deformation is significant, the component design based on elastic analysis is impossible. Therefore, many research works on structural mechanics behavior were carried out to establish a high temperature structural design guideline and creep analysis methods. This paper reviews structural design of the high temperature components for the HTTR made of Hastelloy XR, 2 1/4Cr–1Mo steel, austenitic stainless steels SUS321TB and SUS316, and 1Cr–0.5Mo–V steel.     

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.     

Development of Manufacturing Process of PNC-FMS Wrapper Tube with SUS316 Short Joint

When a ferritic-martensitic stainless steel (PNC-FMS) wrapper tube having far greater swelling resistance against neutron irradiation is applied in the JOYO or MONJU reactor, it becomes necessary to weld it with SUS316 austenitic stainless steel (entrance nozzle and handling head). Such welding between PNC-FMS and SUS316 causes the delta (δ) ferrite formation at heat-affected zone, which leads to significant toughness degradation. In addition, bending of wrapper tube caused by their differential thermal expansion should be straightened. For preventing those problems, manufacturing process of the complex wrapper tube was developed. This process involves TIG-welding with SUS316 short pipe joints in 50mm length to both ends of a PNC-FMS round tube, and then performing the drawing and normalizing and tempering. Normalizing induces complete disappearance of the δ ferrite in the course of wrapper tube manufacturing. The mechanical properties of PNC-FMS/SUS316 welded zone were confirmed to be equivalent to those of the base metal even after thermal aging.     

Effect of thermal aging on the room temperature tensile properties of AISI type 316LN stainless steel

Round tensile specimens of AISI type 316LN stainless steel, thermally aged at 1123 K for 0, 2, 10, 25, 100, 500 and 1000 h, were tested for tensile properties at room temperature at a strain rate of 7.7 × 10⁻³ s⁻¹. The changes in tensile properties were correlated to the transmission electron microscopic studies. The various stages of nitrogen repartitioning including Cr–N cluster formation, intragranular and subsequent cellular precipitation of Cr₂N were found to have a strong influence on the yield strength (YS) and ductility of the material. However, the changes in ultimate tensile strength (UTS) with aging were negligible. The results of electrochemical extraction of secondary phases clearly indicated a two-slope behavior. X-ray diffraction analysis of electrochemically extracted residue suggested that the initial smaller sloped line corresponded to the precipitation of the Cr₂N phase while the line with larger slope at longer aging time corresponded to the domination of chi phase precipitation.     

Creep-fatigue interaction: data base analysis and applications

Creep-fatigue tests were conducted with a Type 304 stainless steel at 650°C using a wide variety of strain wave forms. Wave shape and hold-time effects were of special interest.Two distinct approaches were developed for analyzing the above test results in 10⁻⁹ mbar vacuum. Both approaches are based on a concept of “pure” creep-fatigue interaction.Data obtained in various environments such as air, 10⁻⁶ mbar vacuum and sodium are also listed. Creep-fatigue behavior in different environments are compared and the role of environmental effect is qualitatively discussed.     

Low temperature fatigue crack propagation in neutron-irradiated type 316 steel and weld metal

The fast cycling fatigue crack propagation characteristics of type 316 steel and weld metal have been investigated at 380°C after irradiation to 1.72?1.92 × 10²⁰ n/cm²($\text{E > 1}$ MeV) and 2.03 × 10²¹ n/cm² ($\text{E > 1}$ MeV)at the same temperature. With mill-annealed type 316 steel, modest decreases in the rates of crack propagation were observed for both dose levels considered, whereas for cold-worked type 316 steel irradiation to 2.03 × 10²¹ n/cm² ($\text{E > 1}$ MeV) caused increases in the rate of crack propagation. For type 316 weld metal, increases in the rate of crack propagation were observed for both dose levels considered.The diverse influences of irradiation upon fatigue crack propagation in these materials are explained by considering a simple continuum mechanics model of crack propagation, together with the results of control tensile experiments made on similarly irradiated materials.     

High temperature design curves for high nitrogen grades of 316LN stainless steel

Nitrogen alloyed low carbon grade 316L(N) stainless steel (SS) is a major structural material for high temperature structural components of sodium cooled fast reactors. With a view to significantly enhance the high temperature mechanical properties of 316L(N) SS and thereby increase the design life of structural components from 40 years to 60 years, the influence of nitrogen content on the tensile and creep properties of this steel has been investigated. Four heats of 316LN SS with 0.07, 0.11, 0.14, and 0.22 wt.% nitrogen were used in this investigation. Tensile tests were carried out at various temperatures between room temperature and 850 °C. Creep tests were carried out at 650 °C at various stress levels in the range of 140-225 MPa. The maximum rupture life in these tests was 16,000 h. The tensile and creep data were analysed according to RCC-MR nuclear code procedures and the design curves have been generated. The tensile and creep strength of 316L(N) SS have been found to improve significantly by increasing the nitrogen content.     

Low temperature fatigue crack propagation in neutron-irradiated type 316 steel and weld metal

The fast cycling fatigue crack propagation characteristics of type 316 steel and weld metal have been investigated at 380°C after irradiation to 1.72–1.92 × 10²⁰n/cm² (E>1 MeV) and 2.03×10²¹n/cm² (E>1 MeV) at the same temperature. With mill-annealed type 316 steel, modest decreases in the rates of crack propagation were observed for both dose levels considered, whereas for cold-worked type 316 steel irradiation to 2.03 ×10²¹ n/cm² (E>1 MeV) caused increases in the rate of crack propagation. For type 316 weld metal, increases in the rate of crack propagation were observed for both dose levels considered.The diverse influences of irradiation upon fatigue crack propagation in these materials are explained by considering a simple continuum mechanics model of crack propagation, together with the results of control tensile experiments made on similarly irradiated materials.     

Analysis of tensile deformation and failure in austenitic stainless steels: Part I - Temperature dependence

This paper describes the temperature dependence of deformation and failure behaviors in the austenitic stainless steels (annealed 304, 316, 316LN, and 20% cold-worked 316LN) in terms of equivalent true stress-true strain curves. The true stress-true strain curves up to the final fracture were calculated from tensile test data obtained at −150 to 450 °C using an iterative finite element method. Analysis was largely focused on the necking and fracture: key parameters such as the strain hardening rate, equivalent fracture stress, fracture strain, and tensile fracture energy were evaluated, and their temperature dependencies were investigated. It was shown that a significantly high strain hardening rate was retained during unstable deformation although overall strain hardening rate beyond the onset of necking was lower than that of the uniform deformation. The fracture stress and energy decreased with temperature up to 200 °C and were nearly saturated as the temperature came close to the maximum test temperature 450 °C. The fracture strain had a maximum at −50 to 20 °C before decreasing with temperature. It was explained that these temperature dependencies of fracture properties were associated with a change in the dominant strain hardening mechanism with test temperature. Also, it was seen that the pre-straining of material has little effect on the strain hardening rate during necking deformation and on fracture properties.     

Effects of hold time and neutron irradiation on the low-cycle fatigue behaviour of type 316-CL and their consideration in a damage model

The creep fatigue behaviour of AISI type 316 L(N) plate material has been investigated in the temperature range of 450–750 °C by performing axial strain controlled tests with GRIM specimens. The creep and creep fatigue behaviour of austenitic stainless steel material is known to be prone to neutron irradiation-induced embrittlement. Therefore, the irradiation behaviour was studied by performing irradiation experiments in the High Flux Reactor (HFR) of Petten at 550 °C. A newly developed damage model for time-dependent damage was applied to describe the failure behaviour of AISI 316 L(N) in the cyclic tests performed.     

Stress corrosion cracking susceptibility of austenitic stainless steels in supercritical water conditions

The presented paper summarizes the results of general corrosion and stress corrosion cracking (SCC) susceptibility tests in supercritical water (SCW), studied for austenitic stainless steel 316L, with the aim to identify maximum SCW temperature usability and specific failure mechanisms prevailing during slow strain-rate tensile (SSRT) tests in ultra-pure demineralized SCW solution with controlled oxygen content. The general corrosion tests clearly revealed the applicability of austenitic stainless steel in SCW to be limited to 550 °C as maximum temperature as oxidation rates of austenitic stainless steels 316L increase dramatically above 550 °C. The SSRT tests were performed using a step-motor controlled loading device in an autoclave at 550 °C SCW. Besides the strain rate (resp. crosshead speed), the oxygen content was varied in the series of tests. The obtained results showed that even at the lowest strain rate, a serious increase of SCC susceptibility, as typically characterized by IGSCC crack growth, was not observed. The fractography confirmed that failure was due to a combination of transgranular SCC and transgranular ductile fracture. Based on fractographic findings a phenomenological map describing the SCC regime of SSRT test parameters could be proposed for AISI 316L.     

Some advances in fracture studies under the heavy-section steel technology program

Recent results are summarized from HSST studies in three major areas that relate to assessing nuclear reactor pressure vessel integrity under pressurized-thermal-shock (PTS) conditions. These areas are irradiation effects on the fracture properties of stainless steel cladding, crack run-arrest behavior under non-isothermal conditions, and fracture behavior of a thick-wall vessel under combined thermal and pressure loadings.Since a layer of tough stainless steel weld overlay cladding on the interior of a pressure vessel could assist in limiting surface crack extension under PTS conditions, its resistance to radiation embrittlement was examined. A stainless steel overlay cladding, applied by a submerged arc, single-wire, oscillating-electrode method, was irradiated to 2 × 10²³ neutrons/m² (> 1 MeV) at 288°C. Yield strength increases up to 27% and a slight increase in ductility were observed. Charpy V-Notch data showed a ductile-to-brittle transition behavior caused by temperature-dependent failure of the 8-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.Crack-arrest behavior of A533 grade B class 1 steel was examined for temperatures extending above the onset of Charpy upper-shelf. Crack-arrest experiments that use wide-plate specimens have shown crack arrest occurring prior to transition to tearing or tensile instability. High values of crack-arrest toughness have been recorded (static values above 400 MPa that are well above the maximum value that safety assessment criteria assume such materials can exhibit.A validation experiment was performed by exposing an intentionally flawed HSST intermediate test vessel to combined pressure and thermal transients. The experiment addressed warm-prestressing phenomena, crack propagation from brittle to ductile regions, and crack stabilization in ductile regions. Test and analysis results are summarized.     

Mechanical response of LMFBR and P1 cladding tubes under transient heating

Type 316 stainless steel tubing specimens comparable to LMFBR cladding were burst tested with relatively constant internal pressure in the 219–836 psi range and with increasing temperature. Continuous measurements of diameter change, temperature, and pressure were recorded as the samples were heated to temperatures near the melting point at rates from 10–1800°F/sec. The effects of varying initial wall thickness, cold work level, length, and thermal experience were explored. Ductile failures were observed at 10°F/sec, and stable strains at time of failure were greater than those reported by HEDL. At 200°F/sec the initially 20% cold-worked, 15-mil wall tubing produced brittle failures; while initially 40% cold worked, 10-mil wall samples displayed a mixture of ductile and brittle features. At 1000°F/sec the behavior of the latter material was prodominately brittle, although stable strains as large as 6% were observed. Failure temperatures were generally above 2000°F. When substantial ductility was displayed, an exponentially increasing stable strain was recorded as temperature and time progressed: from such curves temperatures corresponding to 1% strain were derived. Factors controlling the mechanical response appear to be separable by analysis based on the recorded data and the variety of materials and conditions of the tests.