Longitudinal Seam Welded Piping Assessment in Refinery Reformer Units

Creep is progressive deformation of material over an extended period when exposed to elevated temperature and stresses below the yield strength. Poor Creep ductility and cracking can be a problem above 900 °F (482°C) in the HAZ of low alloy (Cr-Mo) steel. High stress areas, including supports, hangers and fittings are more vulnerable to cracking. Creep cracking has occurred in longitudinal pipe welds with excessive peaking or welds with poor quality. Numerous incidents of cracking in low alloy (Cr-Mo) steel have been reported in the power industry and in refineries with major concern in longitudinal seam welds as well as highly stressed welds in reactors-heaters interconnecting piping. This paper presents the results of an assessment performed on reactors-heaters interconnecting piping in a catalytic reformer unit with a maximum operating temperature of about 950 °F (510 °C) at 250 psig (1.7 MPa) (> 40 years in-service). Comprehensive inspection including visual, dye penetrant testing, thickness measurements and peaking measurements have been performed. Phased Array Ultrasonic Testing (PAUT) was utilized to detect crack-like defects and flaws. Detailed pipe stress analysis and finite element analyses (FEA) were also performed.

An Efficient Modelling Approach for Predicting Residual Stress in Power-Beam Welds

The prediction of welding residual stress in components is often an important input to structural integrity assessments. An efficient modelling approach was developed for predicting residual stress in power-beam welds, including validation against residual stress measurements. Specifically, sequentially coupled thermo-mechanical finite element analysis was conducted using a simplified heat source that was tuned to the observed fusion zone from a weld macrograph and thermocouple data for a series of electron beam welds in 316L austenitic stainless steel with a variety of geometries. The predicted residual stresses were compared with contour method and neutron diffraction residual stress measurements.

Evaluation of Creep Properties of Alloy 690 Steam Generator Tubes at High Temperature Using Tube Specimen

Thermally induced steam generator (SG) tube failures caused by hot gases from a damaged reactor core can result in a containment bypass event and may lead to release of fission products to the environment. A typical severe accident scenario is a station blackout (SBO) with loss of auxiliary feedwater. Alloy 690 which has increased the Cr content has been replaced for the SG tube due to its high corrosion resistance against stress corrosion cracking (SCC). However, there is lack of research on the high temperature creep rupture and life prediction model of Alloy 690. In this study, creep test was performed to estimate the high temperature creep rupture life of Alloy 690. Based on reported creep data and creep test results of Alloy 690 in this study, creep life extrapolation was carried out using Larson-Miller Parameter (LMP), Orr-Sherby-Dorn (OSD), Manson-Haferd Parameter (MHP), and Wilshire’s approach. And a hyperbolic sine (sinh) function to determine master curves in LMP, OSD and MHP methods was used for improving the creep life estimation of Alloy 690 material.

Finite Element Modeling of Hybrid Friction Diffusion Welding of Tube-Tubesheet Joints

Hybrid Friction Diffusion Bonding (HFDB) is a solid-state welding process that proved its capability of producing sound tube-tubesheet joints, but with limitations on tube thickness (up to 1mm) and tube-tubesheet materials. In the petrochemical industry, there is a great demand for the use of carbon steel shell and tube heat exchangers. To investigate the feasibility of HFDB techniques in joining thicker tube (i.e 2.1 mm) on tubesheet joint, a three-dimensional thermo-mechanical finite element model (FEM) was developed and solved using ABAQUS (commercial finite element analysis (FEA) software). The model was used to predict the temperature distribution and developed stresses during and after welding. The model considered temperature dependent material properties while Johnson-cook model was used to govern material plastic flow behavior. In this paper,19 mm (¾ in) ASTM 179 cold-drawn carbon steel tube into an ASTM A516 Grade 70 tubesheet joints was simulated. Results are validated based on temperature measurements, which was found in good agreement with experimental results. The developed model can be used to optimize processing parameters (i.e. tool rotational speed, dwell time “holding time”, and forging force.. etc) and study their effect on material flow and developed stresses.

Fracture and Deformation Behavior in Slow-Strain-Rate Tensile Testing of Cu–Ni Alloy With Internal Hydrogen

The effect of hydrogen on the deformation and fracture behavior in pure Cu, pure Ni and Cu–Ni alloy was studied via tensile tests of H-charged, smooth and circumferentially-notched specimens at room temperature (RT) and 77 K. Hydrogen-diffusion properties were determined by the desorption method. To obtain a uniform hydrogen concentration in the H-charged specimens, specimens were exposed to 100-MPa hydrogen gas at 543 K for 200 h, based on the determined hydrogen diffusivity. In tensile tests of smooth pure Ni and Cu–Ni alloy specimens at RT, common hydrogen effects were detected, namely, an increase in yield and flow stresses — a hardening effect; and a ductility loss that was accompanied by a change in fracture surface from ductile to brittle feature — an embrittling effect. With regard to the embrittling effect, the pure Ni and Cu–Ni alloy showed different fracture-surface morphologies at RT; the pure Ni showed an intergranular (IG) surface and the Cu–Ni alloy surface was flat. However, a number of IG cracks were detected beneath the fracture surfaces on the smooth Cu-Ni alloy. The tensile tests of the H-charged smooth specimens at 77 K yielded an IG surface for the pure Ni and a ductile fracture surface with dimples in the Cu–Ni alloy. In contrast, tensile tests of the H-charged, notched specimens at RT demonstrated clear IG fractures for the pure Ni and Cu–Ni alloy. These facts indicate that IG cracking was the first step in the embrittling process for the pure Ni and Cu–Ni alloy, and IG cracking was accompanied by a large plastic deformation that formed the flat surface (unclear IG surface) for the smooth Cu–Ni alloy. Considering that the HE of both pure Ni and Cu–Ni alloy was related to IG cracking, possible mechanisms were discussed and tensile tests performed at 77 K suggested two possibilities: (I) interaction between hydrogen-moving dislocation is more important in the HE process of the Cu-Ni alloy compared to the pure Ni; (II) hydrogen transportation towards grain boundaries are required to cause the IG fracture in the Cu-Ni alloy.

Assessment of Structural Integrity on Irradiated Steel Structure: Focusing on Long Column Type Reactor Pressure Vessel Supports

The structural steel in a nuclear facility experiences significant degradation due to the accumulated neutron irradiation. Particularly, the long-column type reactor pressure vessel supports have been focused since they resist considerable loading to maintain the primary coolant system in their position and experience high neutron irradiation in low temperature, which is an unfavorable condition for the fracture toughness. This study implemented the API 579-1/ASME FFS-1, fitness-for-service (FFS) method to consider both irradiated mechanical properties and multiple loading cases. A three-dimensional (3D) finite element model of long column type reactor pressure vessel support was built for the linear analysis. The metallurgical properties of reactor pressure vessel support for assessment were estimated by empirical equations. This study provides the structural margin of long-column type reactor pressure vessel support by levels of the loads and levels of the neutron fluence.

An Update on the Investigation of Fracture Toughness Properties of the High Flux Reactor Vessel From Surveillance Test Campaign in 2017

The reactor vessel of the High Flux Reactor (HFR) in Petten has been fabricated from Al 5154-O alloy grade with a maximum Mg content of 3.5 wt. %. The vessel experiences large amount of neutron fluences (notably at hot spot), of the order of 1027 n/m2, during its operational life. Substantial damage to the material’s microstructure and mechanical properties can occur at these high fluence conditions. To this end, a dedicated surveillance program: SURP (SURveillance Program) is executed to understand, predict and measure the influence of neutron radiation damage on the mechanical properties of the vessel material. In the SURP program, test specimens fabricated from representative HFR vessel material are continuously irradiated in two specially designed experimental rigs. A number of surveillance specimens are periodically extracted and tested to evaluate the changes in fracture toughness properties of the vessel as function neutron fluence. The surveillance testing results of test campaigns performed until 2015 were published previously in [1, 2]. The current paper presents fracture toughness and SEM results from the recent surveillance campaign performed in 2017. The fracture toughness specimen tested in this campaign received a thermal neutron fluence of 13.56 x1026 n/m2, which is ∼8.9 × 1025 n/m2 more than the thermal fluence received by the specimen tested in SURP 2015 campaign. These results from this campaign have shown no change in the fracture toughness from the values measured in the previous SURP campaign. The SEM observations are performed to study the fracture surface, to measure (by WDS) the transmutation Si formed near crack tip and to investigate various inclusions in the microstructure. SEM fracture surface investigation revealed a tortuous (bumpy) fracture surface constituting micro-scale dimples over majority of the fracture area. Islands of cleavage facets and secondary cracks have been observed as well. EDS analysis of various inclusions in the microstructure revealed presence of Fe rich inclusions and Mg-Si rich precipitates. Additionally, inclusions rich in Al-Mg-Cr-Ti were identified. Finally, changes in mechanical properties of Al 5154-O alloy with an increase in neutron fluence (or transmutation Si) are discussed in correlation with SEM microstructure and fracture morphology observed in SEM. TEM investigation of precipitate microstructure is ongoing and those results will be published in future.

Ductility and Fatigue Strength Loss of a Hydrogen-Charged 316L Austenitic Stainless Steel

The susceptibility of 316L-type austenite stainless steel to hydrogen was quantified by means of SSRT results and low-cycle fatigue life measurement. Both tests were conducted in the air condition after being charged with high-pressure hydrogen gas of 10 MPa and a temperature of 300°C for 120 hours. In addition, SSRT tests in gaseous hydrogen at a pressure of 10 MPa were also performed and compared to the tests conducted in hydrogen pre-charged and as-received conditions. The 0.2% yield strength and tensile strength did not show there to be a considerable difference between hydrogen pre-charging and the as-received conditions, whereas the gaseous hydrogen condition revealed a remarkable degradation in tensile properties, especially in terms of fracture elongation. In the case of fatigue life test, a considerable influence of hydrogen pre-charging in fatigue life properties was observed in the high strain amplitude regime whereas the measured values in the low strain deformation region are consistently comparable to that in the as-received condition. Fatigue limit was not affected by hydrogen pre-charging.

Using Additive Manufacturing for Making Light Water Reactor Components

Making light water reactor (LWR) components using additive manufacturing (AM) provides a high degree of design freedom to create complex near net shape geometries, with significant reduction in the deployment time. Since most of the current AM research focus on the refining of the fabrication variables, little information exists on the actual corrosion behavior (including general corrosion, localized corrosion and environmentally assisted cracking - EAC) of AM components. The most frequent operational material failure modes in LWR is EAC and debris fretting of fuel rods, therefore the objective of this work was to evaluate the environmental performance of AM type 316L SS in relation to the traditionally melted and forged wrought material. Stress corrosion cracking results show similar behavior between AM and wrought 316L SS in high temperature water. Ambient temperature electrochemical testing showed that the AM 316L SS was slightly more resistant to localized corrosion than the traditional wrought material and that AM material did not suffer sensitization during fabrication or heat treatment processes.

Thermal Aging Assessment and Microstructural Investigations of Alloy 52 Dissimilar Metal Welds for Nuclear Components

Dissimilar Metal Welds (DMW) are commonly found between the ferritic low alloy steel heavy section components and the austenitic stainless steel piping sections in nuclear power plants. In the EPR™ design which is the latest FRAMATOME Pressurized water reactor (PWR) these DMW involve a narrow gap technology with no buttering, and only one bead per layer of a nickel base alloy weld filler metal (Alloy 52). In order to assess the thermal aging performance of this relatively new narrow gap DMW design, a significant internal R&D program was launched some years ago. Several representative mock-ups were thoroughly characterized in the initial condition as well as in the thermal aged condition, up to 50,000 hours aging at 350°C. The characterisations were focused on the fusion line between the ferritic low alloy steel (LAS) and the nickel base alloy since a particular microstructure is present in this area, especially in the carbon depleted area of the Heat Affected Zone (HAZ) which is often regarded as the weak zone of the weld joint. Metallography, hardness, nanohardness, chemical analyses, and Atom Probe Tomography, as well as fracture toughness tests were carried out on different specimens in different thermal aging conditions. The results show that the fracture toughness behaviour in the ductile-brittle domain of the low alloy steel carbon depleted HAZ at the interface with the alloy 52 weld metal of the DMWs is excellent, even for a thermal ageing equivalent to 60 years at service temperature. This was found in spite of the carbon depleted zone of the HAZ, the variations of hardness, chemical composition, particularly the carbon gradients, and the thermal aging effect induced by phosphorous segregation at grain boundaries.