Sensitization development in austenitic stainless steel piping

Pacific Northwest Laboratory and the Division of Engineering Technology of the US Nuclear Regulatory Commission are conducting a program to determine a method for evaluating welded and repair-welded stainless steel piping for light-water reactor service. Validated models, based on experimental data, are being developed to predict the degree of sensitization (DOS) and the intergranular stress corrosion cracking (IGSCC) susceptibility in the heat-affected zone (HAZ) of the SS weldments. The cumulative effects of material composition, past fabrication procedures, past service exposure, weldment thermomechanical (TM) history, and projected post-repair component life are being considered.     

Ferrite formation in neutron-irradiated austenitic stainless steel

Magnetic measurements were carried out on type 316, 321 and three modified heats of 316 austenitic stainless steels that had been irradiated to high fluences (1 ? 8 × 10²²n/cm², E > 0.1 MeV) in EBR-II at temperatures ranging from 450–700°C. Most of the specimens showed increases of magnetization after exposure to the reactor environment that can be attributed to formation of numerous small ferrite particles. The amount of ferrite formed during irradiation is a function of alloy composition as well as irradiation temperature and fluence. Specimens with low molybdenum concentrations had a greater ferrite content than specimens with the normal molybdenum content of type 316 stainless steel. A modified heat of type 316 with 0.23 wt% Ti had lower levels of ferrite under given irradiation conditions than the other heats. Some particles with diffraction patterns corresponding to the ferrite phase were found in an irradiated type 321 stainless specimen, but none were observed in the type 316 stainless specimens.     

强流脉冲电子束辐照诱发的AISI 304奥氏体不锈钢中的空位簇缺陷

利用强流脉冲电子束技术对AISI 304奥氏体不锈钢进行辐照处理,采用透射电子显微镜详细地分析了辐照诱发的空位簇缺陷结构.1次辐照未产生空位簇缺陷结构;5、10次辐照后诱发大量的空位簇缺陷结构,缺陷尺寸通常小于10 nm,随辐照次数增加,缺陷簇数量明显增加,但尺寸增加不明显;小缺陷簇主要由空位型位错圈和少量的堆垛层错四面体(SFT,stacking fault tetrahedral)组成,SFT占整个缺陷簇的比例低于1%;少量的孔洞缺陷在孪晶片附近出现.10次辐照后,孪晶片前沿或附近形成大量的大尺寸SFT,其最大尺寸可达250 nm,这些大尺寸SFT的形成机制与小尺寸SFT的有所不同,通过1/3<111>位错攀移在孪晶片前沿和孪晶界上的台阶处形成的压杆位错核心吸收周围丰富的空位而长大可能是大尺寸SFT形成的原因.     

Radiation-induced segregation in desensitized type 304 austenitic stainless steel

Radiation-induced segregation (RIS) in desensitized type 304 stainless steel (SS) was investigated using a combination of electrochemical potentiokinetic reactivation (EPR) test and atomic force microscopy (AFM). Desensitized type 304 SS was irradiated to 0.43 dpa (displacement per atom) using 4.8 MeV protons at 300 °C. The maximum attack in the EPR test for the irradiated desensitized SS was measured at a depth of 70 μm from the surface. Grain boundaries and twin boundaries got attacked and pit-like features within the grains were observed after the EPR test at the depth of 70 μm. The depth of attack, as measured by AFM, was higher at grain boundaries and pit-like features as compared to twin boundaries. It has been shown that the chromium depletion due to RIS takes place at the carbide-matrix as well as at the carbide-carbide interfaces at grain boundaries. The width of attack at grain boundaries after the EPR test of the irradiated desensitized specimen appeared larger due to the dislodgement of carbides at grain boundaries.     

钼离子注入奥氏体不锈钢引发亚结构的变化

用加速电压为５０ｋＶ，剂量为２×１０＾１７ｉｏｎｓ／ｃｍ＾２的钼离子注入经１１５０℃固溶处理的奥氏体不锈钢中，利用透射电子显微镜弱束成像技术研究了被注入材料的横截面样品，发现离子辐照引发被注入亚结构变化的深度远大于离子本身的注入深度，其亚结构为多种位错组成的复杂位错组态，这种亚结构会使被注入材料具发于加工硬化的效果。探讨了离子注入材料中由辐射引发的加工硬化现象，即所谓的“长程效应”的原因。     

The role of niobium carbide in radiation induced segregation behaviour of type 347 austenitic stainless steel

The effect of niobium carbide precipitates on radiation induced segregation (RIS) behaviour in type 347 stainless steel was investigated. The material in the as-received condition was irradiated using double-loop 4.8 MeV protons at 300 °C for 0.43 dpa (displacement per atom). The RIS in the proton irradiated specimen was characterized using double-loop electrochemical potentiokinetic reactivation (DL-EPR) test followed by atomic force microscopic examination. The nature of variation of DL-EPR values with the depth matched with the variation of the calculated irradiation damage (dpa) with the depth. The attack on grain boundaries during EPR tests was negligible indicating absence of chromium depletion zones. The interface between niobium carbide and the matrix acts as a sink for point defects generated during irradiation and this had reduced point defect flux toward grain boundaries. The attack was noticed at a few large cluster of niobium carbide after the DL-EPR test at the depth of maximum attack for the irradiated specimen. Pit-like features were not observed within the matrix indicating the absence of chromium depletion regions within the matrix.     

Serrated flow behavior in AL6XN austenitic stainless steel

Serrated flow behavior of the AL6XN austenitic stainless steel has been investigated at different temperatures and strain rates. The results show the serrated flow, peak/plateau in flow stress and negative strain rate sensitivity appearing in tensile deformation of the AL6XN steel at 773-973 K and 3.3 × 10⁻⁵-3.3 × 10⁻³ s⁻¹ (excluding 873 K, 3.3 × 10⁻⁵ s⁻¹), suggesting the occurrence of dynamic strain aging (DSA). The activation energy for type-A and -(A + B) serrations was calculated to be 304 kJ/mol and diffusion of substitutional solutes, such as chromium and molybdenum is considered as the mechanism of serrated flow. TEM observations further revealed a typical planar slip mode in the regime of DSA of the deformed AL6XN steel.     

Mechanisms of stress relief cracking in titanium stabilised austenitic stainless steel

The heat affected zone (HAZ) of AISI 321 welds may exhibit a serious form of cracking during service at high temperature. This form of damage, called ‘stress relief cracking’, is known to be due to work hardening but also to aging due to Ti(C,N) precipitation on dislocations which modifies the mechanical behaviour of the HAZ. The present study aims to analyse the latter embrittlement mechanism in one specific heat of 321 stainless steel. To this end, different HAZs are simulated using an annealing heat-treatment, followed by various cold rolling and aging conditions. Then, we study the effects of work hardening and aging on Ti(C,N) precipitation, on the mechanical (hardness, tensile and creep) behaviour of the simulated HAZs and on their sensitivity to intergranular crack propagation through stress relaxation tests performed on pre-cracked CT type specimens tested at 600 °C. It is shown that work hardening is the main parameter of the involved mechanism but that aging does not promote crack initiation although it leads to titanium carbide precipitation. Therefore, the role of Ti(C,N) precipitation on stress relief cracking mechanisms is discussed. An attempt is made to show that solute drag effects are mainly responsible for this form of intergranular damage, rather than Ti(C,N) precipitation.     

Study of low-cycle fatigue of AL6XN austenitic stainless steel

The low-cycle fatigue (LCF) behaviors of the AL6XN austenitic stainless steel were investigated at strain rates of 3.3 × 10⁻⁵ to 3.3 × 10⁻³ s⁻¹ and at different temperatures, respectively. The weakened cyclic softening, negative strain-rate-stress response and anomalous temperature-dependence of cycle stress reflect the dynamic strain aging (DSA) hardening during LCF at elevated temperatures. Electron microscopy observations revealed that the dislocation structure changes from the cellular structure at room temperature to the planar slip band, serving as crack initiation sites in the regime of DSA. The DSA hardening results in the reduction of fatigue resistance at elevated temperature via reducing the crack initiation and propagation life.     

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.     

An experimental study on thermal-stress ratchetting of austenitic stainless steel by a three bars specimen

A new test specimen for thermal-stress ratchetting is proposed and tested for standard AISI-304 stainless steel. Its purpose is to avoid the drawbacks of a tubular specimen. It is a special plate-shaped specimen with two symmetrical, straight air gaps in the center portion and load chucks at the ends. Since three surfaces out of four on the side bars are enveloped by sheathed trace heaters, the middle bar becomes a cold element and the rest side bars hot elements during thermal loading. Experiments on thermal-stress ratchetting are performed with this specimen. Interesting incremental strain growth behaviour of the specimen subjected to cyclic thermal stresses under a steady primary load is obtained with regard to an actual strain-hardening material. It was confirmed that incremental strain growth decreased with increase in the cumulative strain growth. There is expected to be an asymptotic cumulative strain growth for any test condition. It is also found that the tendency for theoretical strain growth is fairly similar to that found experimentally. However, the experimental data are considerably greater than the theoretical data. This can be explained qualitatively by a decrease in both yield strength and Young's modulus at high temperature for the hot elements.     

Void swelling and precipitation in a titanium-modified austenitic stainless steel under proton irradiation

The correlation between void swelling and precipitation behavior in a 10% cold worked Fe-16.2Ni-14.6Cr-2.37Mo-1.79Mn-0.53Si-0.24Ti-0.06C alloy was examined with 200 keV proton irradiation. Swelling peak temperature after the proton irradiation to 10 dpa was about 823 K, and void swelling decreased steeply with increase in irradiation temperature from 823 to 923 K. Void swelling increased rapidly from 1.9 to 12.1% with increase in irradiation dose from 20 to 45 dpa at 873 K. Fine intragranular TiC precipitates, which were formed during initial stage of irradiation, dissolved gradually with increase in irradiation dose from 10 to 45 dpa at 873 K, while the amount of precipitation of needle-shaped Fe₂P phase containing titanium increased with increasing dose. The reduction of sink strength of the TiC precipitates due to the dissolution during irradiation was thought to cause the increase of swelling rate with increase in irradiation dose from 20 to 45 dpa at 873 K.     

Development of inelastic constitutive model for austenitic stainless steel for design use

To prevent creep-fatigue failure or excessive deformation in high-temperature components of fast reactor plants, accurate estimation of inelastic deformation is essential. In performing inelastic analysis, employment of constitutive models, which can precisely reproduce inelastic deformation of the material is of critical importance. The authors have been engaged in the development of inelastic constitutive model for the use in structural design assessment of liquid metal-cooled fast reactor plants. Various improvements were made on the nonlinear hardening model proposed by Ohno and Wang, placing an emphasis on capability to simulate inelastic deformation behavior of austenitic stainless steels, under regular or irregular cyclic loading possibly with temperature variation and hold time. It was demonstrated that the model can simulate the inelastic deformation behavior under various loading conditions with a sufficient accuracy.     

Modelling the high temperature mechanical behaviour of helium embrittled austenitic stainless steel

Current theories for the growth of grain-boundary defects during plastic deformation have been used to model the ductility of an irradiated austenitic steel as a function of strain rate and thermal-neutron fluence. Failure of irradiated steels is via the growth and linkage of helium bubbles on the grain boundaries produced by an (n, ) reaction between the thermal neutrons and boron present in the steel as a controlled trace element.

Reasonable agreement is obtained between theoretically predicted and experimentally measured ductilities over a range of strain rates between 10⁻¹⁰ and 10⁻²/s. However, in order to obtain such agreement over the range of strain rates examined, variables such as defect size and spacing had to be carefully selected. The values used were internally self-consistent with the amount of helium known to be availablc to nucleate grain-boundary bubbles.

The models of defect growth have also been used to make some predictions on the creep-fatigue performance of such materials.     

A study on the redistribution of ion-implanted nitrogen in Ti-modified austenitic steel

Titanium modified austenitic steel has been subjected to nitrogen implantation to different fluences. There is a considerable deviation of the experimental depth profile of nitrogen from the theoretical profile obtained using Monte-Carlo codes. Available literature ascribes such redistribution to sputtering; by following the implanted nitrogen and the vacancy-defects alongside using resonance nuclear reaction analysis and positron annihilation spectroscopy, clear experimental evidence for the formation of vacancy-nitrogen complexes is established. The possibility of role of complexes in redistributing the implanted nitrogen is explored here for the first time.     

Time-temperature-sensitization diagrams and critical cooling rates of different nitrogen containing austenitic stainless steels

Nitrogen-alloyed 316L stainless steel is being used as structural material for high temperature fast breeder reactor components with a design life of 40 years. With a view to increase the design life to 60 years and beyond, high nitrogen stainless steels are being considered for certain critical components which may be used at high temperatures. Since carbon and nitrogen have major influence on the sensitization kinetics, investigations were carried out to establish the sensitization behaviour of four heats of 316L SS containing (i) 0.07%N and 0.035%C, (ii) 0.120%N and 0.030%C, (iii) 0.150%N and 0.025%C and (iv) 0.22%N and 0.035%C. These stainless steels were subjected to heat treatments in the temperature range of 823-1023 K for various durations ranging from 1 h to 500 h. Using ASTM standard A262 Practice A and E tests, time-temperature-sensitization diagrams were constructed and from these diagrams, critical cooling rate above which there is no risk of sensitization was calculated. The data established in this work can be used to select optimum heat treatment parameters during heat treatments of fabricated components for fast reactors.     

Laser cladding of Colmonoy 6 powder on AISI316L austenitic stainless steel

Stainless steels are widely used in nuclear power plant due to their good corrosion resistance, but their wear resistance is relatively low. Therefore, it is very important to improve this property by surface treatment. This paper investigates cladding Colmonoy 6 powder on AISI316L austenitic stainless steel by CO₂ laser. It is found that preheating is necessary for preventing cracking in the laser cladding procedure and 450 °C is the proper preheating temperature. The effects of laser power, traveling speed, defocusing distance, powder feed rate on the bead height, bead width, penetration depth and dilution are investigated. The friction and wear test results show that the friction coefficient of specimens with laser cladding is lower than that of specimens without laser cladding, and the wear resistance of specimens has been increased 53 times after laser cladding, which reveals that laser cladding layer plays roles on wear resistance. The microstructures of laser cladding layer are composed of Ni-rich austenitic, boride and carbide.