An evaluation of the fracture toughness of the heat-affected zone (HAZ), which is located under the weld overlay cladding of a reactor pressure vessel (RPV), was performed. Considering inhomogeneous microstructures of the HAZ, 0.4T-C(T) specimens were manufactured from the cladding strips locations, and Mini-C(T) specimens were fabricated from the distanced location as well as under the cladding. The reference temperature ( T o ) of specimens that were aligned with the middle section of a cladding strip (HAZ MCS ) was ∼12°C higher than that of specimens that were aligned with cladding strips at the overlap (HAZ OCS ). T o values of partial area in the HAZ were obtained using Mini-C(T) specimen. The T o values obtained near the side of the cladding were ∼13°C higher than those away from the cladding. T o values of HAZ for both 0.4T-C(T) and Mini-C(T) specimens were significantly lower than that of the base metal at a quarter thickness by 40°C–60°C. Compared to the literature data that indicated fracture toughness at the surface without overlay cladding and base metal of a quarter thickness in a pressure vessel plate, this study concluded that the welding thermal history showed no significant effect on the fracture toughness of the inner surface of RPV steel.

As one of the most promising Generation IV nuclear reactors, thorium molten salt reactor (TMSR) possesses inherent safety, simplified fuel cycle and high power generation efficiency. However, the structure material of TMSR must encounter challenges of high-temperature environment and serious molten fluoride salts corrosion. Although the UNS N10003 alloy has excellent performance in TMSR, the high cost of this alloy reduces the economy of TMSR. Although the austenitic stainless steel possesses excellent high temperature strength and lower cost, it exhibits poor corrosion resistance in molten fluoride salts. Therefore, the purpose of this work is using Ni cladding on austenitic stainless steel by gas tungsten arc welding (GTAW) for corrosion protection and investigating the effects of thermal ageing on microstructure and hardness of cladding. The elements distribution and segregation, the precipitates behavior and the microstructure evolution of interface and cladding layer under as-welded and thermal ageing conditions were studied by a series of characterization methods respectively. The hardness evolution of claddings under as-welded and thermal ageing conditions were tested and discussed.

Nuclear reactor pressure vessels must be able to withstand thermal shock due to emergency cooling during a loss of coolant accident. Demonstrating structural integrity during thermal shock is difficult due to the complex interaction between thermal stress, residual stress, and stress caused by internal pressure. Finite element and analytic approaches exist to calculate the combined stress, but validation is limited. This study describes an experiment which aims to measure stress in a slice of clad reactor pressure vessel during thermal shock using time-resolved synchrotron X-ray diffraction. A test rig was designed to subject specimens to thermal shock, whilst simultaneously enabling synchrotron X-ray diffraction measurements of strain. The specimens were extracted from a block of SA508 Grade 4N reactor pressure vessel steel clad with Alloy 82 nickel-base alloy. Surface cracks were machined in the cladding. Electric heaters heat the specimens to 350°C and then the surface of the cladding is quenched in a bath of cold water, representing thermal shock. Six specimens were subjected to thermal shock on beamline I12 at Diamond Light Source, the UK’s national synchrotron X-ray facility. Time-resolved strain was measured during thermal shock at a single point close to the crack tip at a sample rate of 30 Hz. Hence, stress intensity factor vs time was calculated assuming K-controlled near-tip stress fields. This work describes the experimental method and presents some key results from a preliminary analysis of the data.

During vacuum drying of used nuclear fuel (UNF) canisters, helium pressure is reduced to as low as 67 Pa to promote evaporation and removal of remaining water after draining process. At such low pressure, and considering the dimensions of the system, helium is mildly rarefied, which induces a thermal-resistance temperature-jump at gas–solid interfaces that contributes to the increase of cladding temperature. It is important to maintain the temperature of the cladding below roughly 400 °C to avoid radial hydride formation, which may cause cladding embrittlement during transportation and long-term storage. Direct Simulation Monte Carlo (DSMC) method is an accurate method to predict heat transfer and temperature under rarefied condition. However, it is not convenient for complex geometry like a UNF canister. Computational Fluid Dynamics (CFD) simulations are more convenient to apply but their accuracy for rarefied condition are not well established. This work seeks to validate the use of CFD simulations to model heat transfer through rarefied gas in simple two-dimensional geometry by comparing the results to the more accurate DSMC method. The geometry consists of a circular fuel rod centered inside a square cross-section enclosure filled with rarefied helium. The validated CFD model will be used later to accurately estimate the temperature of an UNF canister subjected to vacuum drying condition.

In this work, a geometrically-accurate two-dimensional (2D) computational fluid dynamic (CFD) model of a used nuclear fuel cask, that can contain up to 32 pressurized water reactor (PWR) used nuclear fuel (UNF) assemblies, is constructed. This model is similar to the TN-32 cask employed in the ongoing high-burnup (HBU) Spent Fuel Data Project lead by the Electric Power Research Institute (EPRI). This model is used to predict the peak cladding temperature under vacuum drying conditions. Due to the symmetry of the cask, only one-eighth of the cross-section is modeled. Steady-state simulations that include the temperature-jump boundary conditions at the gas-solid interfaces are performed for different heat generation rates in the fuel regions and a range of dry helium pressures, from ∼10 5 to 100 Pa. These simulations include conduction within solid-gas regions and surface-to-surface radiation across all gas regions. The peak cladding temperatures are reported for various heat generation rates and rarefaction conditions, along with the maximum allowable heat generation that brings the cladding temperatures to the radial hydride formation limit. The results showed that the decrease of helium pressure significantly increased the temperature of the cladding material compared to the atmospheric pressure condition.

Coke drums undergo cyclic operations typically in temperatures ranging from the ambient temperature to about 500°C. Generally, coke drums are fabricated from clad plates consisting of chromium-molybdenum low alloy steel as the base metal and ferritic stainless steel typically Type 410S stainless steel (TP.410S) as cladding material. TP.410S cladding material is used for protection against sulfide corrosion. Coke drums frequently display major cracking on the inside surface of their circumferential weld along the fusion line between restoration weld and the TP.410S cladding. Moreover, an array of cracks commonly referred to as “elephant skin” are typically observed proximity to the major cracks. These cracks have been reported by multiple papers, particularly in the API Technical Report 934-G, up to the present and primarily attributed to the thermal fatigue which occurs from the repeated thermal loads experienced during each operating cycle. When the existence of severe cracks is recognized, users are forced to schedule extensive repairs. The authors conducted fatigue tests employing coupons made from two kinds of cladding material, TP.410S and Alloy 625, which simulate the circumferential weld seam of coke drums. The paper summarizes the essential properties for the cladding material in order to delay cracking on the inside surface and proposes new common cladding materials which could be employed for coke drums based on test results.

Sulfidation corrosion of the carbon steel tubes at the tube-to-tubesheet joint often governs the life of waste heat boilers in sulphur recovery plants. Conventional tube joints typically have a welded joint located at the hot-side face of the tubesheet. An alternative design involves welding the tubesheet joint at the cold-side face of the tubesheet, close to the boiler feed water. The alternative design also employs stainless steel cladding on the tubesheet face and a tube-hole sleeve selectively at high-temperature locations. Finite element heat transfer analysis is used to establish the thermal profiles of the conventional and the alternative designs. From the worked example, the alternative design provided a lower metal temperature by approximately 80 °F at the joint, as compared to the conventional tube joint. Sulfidation rate prediction based on a sample gas composition using ASSET (Alloy Selection System for Elevated Temperatures) Software predicts that the alternative design can reduce the sulfidation rate by 35% because of the lower metal temperature.

The frequency and extent of vessel bulging and cracking being registered in delayed coke drums throughout the global coking industry has accelerated significantly as refinery operators reduce their cycle times. Several theoretical approaches have been developed to identify how a bulged area may lead to drum damage; however, limited information has been presented to match the theoretical predictions with actual surface damage reported by coke drum operators. The results of hundreds of laser scans spanning the last 25 years have been analyzed to correlate vessel bulging with observed surface damage. Specific categorizations of bulge profiles, and the proximity of these to circumferential weld seams (circs), have been calibrated against hundreds of real-world examples of drum damage and failure, including through wall cracking and stress cracking of the cladding, and further associated with the triggers for repair strategies implemented by industry leading refiners. Strong correlations between specific aspects of bulge profiles and the presence of surface damage were found resulting in an assessment tool that can rank and prioritize coke drum distortions on the likelihood of damage, and can serve as a useful guide for planning future coke drum maintenance.

To perform the integrity assessment of a reactor pressure vessel (RPV) related to Pressurized Thermal Shock (PTS), we model the RPV using the 3D finite element method (FEM). Accurate prediction of temperature and stress fields is determined by using 2-Phase computational fluid dynamics (CFD) simulation in combination with an appropriate finite element discretization of the RPV wall. The cladding and the ferritic low alloy steel are considered as two separated layers, which can be intersected by superficial cracks. The calculation of the stress intensity factor (SIF) in mode I is based on the linear fracture mechanics theory (LEFM) and hypothetical cracks are located in different locations to consider the most critical cases. In the present study, the sub-modeling technique is implemented to refine the mesh required by the fracture analysis in the region of interest. Three types of cracks are considered: axial, circumferential and inclined. The performed integrity assessment uses the master curve approach. The stress intensity factor in the deepest point of a surface crack was compared with the material’s fracture toughness. In previous studies the integrity of the RPV subjected to medium and small break Loss-of-Coolant Accident (MBLOCA and SBLOCA, respectively) has been evaluated, therefore the concern in this contribution is the large break of Loss-Of-Coolant Accident (LBLOCA). The combination of 3D FEM with CFD simulations allows us to study the influence of the dynamic cooling plume on the stress intensity in more detail than with the conventional one dimensional method.

Dissimilar Metal Welds (DMWs) are used in many industries to increase the service life of components. Typically a corrosion resistant material is bonded to the internal surface of a second material, this is often referred to as cladding. This type of composite component provides the desired mechanical properties and corrosion resistance without excessive cost. Structural integrity assessments can be used to estimate the remaining life of critical components but these methods are complicated and results are dependent on factors such as: material properties, defect size, defect location and the direction of defect propagation. The cladding process is generally conducted via welding which results in yield strength residual stresses. The interaction between primary and secondary stresses on crack driving force (CDF) are not well understood, particularly at the interface between multiple materials where the microstructure is not homogeneous. These unknown factors contribute to overly conservative structural integrity assessments. This work investigates the fracture toughness of defects orientated parallel the interface of a dissimilar metal clad weld using experimental and finite element (FE) methods. A nuclear grade austenitic clad ferritic steel was investigated. A fracture toughness test was conducted on a CT specimen and the influence of the materials boundary on crack propagation was examined using digital image correlation (DIC). DIC of the fracture toughness test shows that high amounts of strain exist both in the region immediately surrounding the crack tip and in the ferritic material at the material boundary. Post-test analysis of the specimen showed two defects initiated at the root of the notch on the outer surface of the specimen. One longer notch was found internally. Two-dimensional plane stress and plane strain FE analyses were conducted and compared to the experimental results.

Cladding structural components with a corrosion resistant material may greatly extend the design life of molten salt reactor concepts. A complete design methodology for such cladded, high temperature nuclear components will require addressing many issues: fabrication, corrosion resistance, metallurgical interaction, and the mechanical interaction of the clad and base materials under load. This work focuses on the final issue: the mechanical interaction of the base and clad under creep-fatigue conditions. Depending on the relative mechanical properties of the two materials the clad may substantially influence the long-term cyclic response of the structural system or its effect might be negligible. To quantify the effect of different clad material properties we develop an efficient method for simulating pressurized cladded components in the limiting case where the section of interest is far from structural discontinuities. Using this method we evaluate the mechanics of the clad/base system and identify different regimes of mechanical response. The focus is on situations relevant to high temperature nuclear components: thermal-cyclic Bree-type problems and similar axisymmetric structures. The insights gained from these structural studies will form the basis for developing design rules for high-temperature, nuclear, cladded components.

Preventing brittle fracture is an essential part of instituting life-cycle management strategies for fixed pressure equipment. Using fracture mechanics principles to establish permissible minimum pressurization temperature (MPT) envelopes for components is one way to mitigate the potential for unstable flaw growth. In the refining industry, heavy-walled, low-alloy hydroprocessing reactors are designed to operate at elevated temperatures and high hydrogen partial pressures. Components that operate in high-pressure hydrogen environments require special treatment and necessitate guidance that falls outside the bounds of current pressure vessel construction codes. This operating environment results in two factors that affect the MPT envelope: long-term temper embrittlement and hydrogen embrittlement. Additionally, hydrogen charging can manifest damage in two ways: fast (brittle) fracture due to a reduction in fracture toughness and slow (subcritical) hydrogen-assisted crack growth. When developing a MPT envelope for a given component, both failure modes need to be considered in addition to residual stress effects from weld overlay or cladding. MPT envelopes provide insight into the permissible pressure-temperature combinations for specific locations and for chosen reference flaw sizes. From a reliability standpoint, understanding the risk of brittle fracture associated with heavy-walled reactors for all operating scenarios is crucial. Furthermore, taking the appropriate life-cycle management steps, such as establishing MPT envelopes, coupling MPT analysis predictions with targeted inspection, optimizing process operating conditions, and developing a flaw acceptance criteria to mitigate the risk of crack propagation and ultimately, brittle fracture is essential. In this study, a fracture-mechanics based methodology is summarized that is fully documented in upcoming Welding Research Council (WRC) Bulletin 562 [1] to determine MPT envelopes for all components (in any service environment) based on fast fracture with supplemental MPT requirements based on slow fracture for equipment that operates in high-pressure hydrogen environments. Finally, a finite element analysis-based case study of a 2-1/4-Cr-1-Mo hydrotreater reactor is summarized and practical life cycle-management guidance is offered based on analysis results. This example highlights how evaluation of start-up and shut-down procedures for heavy-walled reactors has the potential to save significant time and related cost per unit shut-down cycle, while maintaining an acceptable risk tolerance against subcritical crack growth and brittle fracture.

Hydroprocessing reactors are made of heavy wall Cr-Mo steels with stainless steel cladding and installed in the most of oil refineries in the world during past half century. After long time operation, they found several material degradation and/or damages due to high pressure and high temperature hydrogen services. This paper presents typical remediation method of flaws in stainless steel cladding of reactors including flaw sizing, FFS assessment and repair procedures.

Cladding titanium-steel plate is often used in design of pressure vessel. Using large finite element software, models of cladding titanium-steel tube sheets of tubular heat exchangers are established, to investigate the effect of different tube patterns, including triangle tube pattern and square tube pattern on cladding titanium-steel tube sheets. What is more, stress distribution characteristics of cladding titanium-steel tube sheets in different tube patterns are obtained under different tube diameters. Results show that the effect of tube patterns on cladding titanium-steel tube sheets mainly concentrated in tube layout area. Compared with triangular tube pattern, square tube pattern could effectively improve the phenomenon of stress concentration of cladding titanium-steel tube sheet at the connection of base layer and cladding layer, and then improve performance of cladding titanium-steel tube sheet. However, increase of tube diameters makes an adverse effect on performance of cladding titanium-steel tube sheets in both tube patterns. Through reasonable decrease of tube diameter could enhance the performance of cladding titanium-steel tube sheet. Results provide references to study and design heat exchanger with cladding steel tube sheet.

Jagged cracks were observed in SA240 Type 405 stainless steel cladding of Inconel 625 overlay repaired coke drums. It is found that intergranular cracking is the dominant fracture mode in the fine-grained heat-affected zone (FGHAZ) of the boat specimens. The sensitization effect from the operation and welding thermal cycles leads to the depletion of Cr with the preferential precipitation of Cr-rich M 23 C 6 carbides along the grain boundaries. The cladding FGHAZ has the largest frequency of grain boundaries with higher local strain levels and the highest fraction of grain boundary Cr-rich M 23 C 6 carbides. Thermal stress distributions predicted by finite element analysis clearly show the maximum shear stress to exhibit the typical “jagged” pattern near the cracked regions. Thermal expansion coefficient and strength mismatch among the shell base metal, cladding, and overlay is believed to have caused the unique jagged maximum shear stress distribution in the cladding HAZ of Inconel 625 overlay. The magnitude of this thermal stress can reach the yield strength of the cladding at 900 °F (482 °C) service temperature, therefore, provides the driving force for the jagged cracking formation in the sensitized HAZ.

In nuclear power plants, inner surface of reactor pressure vessel, which is made of low-alloy steel, is cladded by austenitic stainless steel to improve corrosion resistance of vessel. In an evaluation of fatigue crack propagation behavior at J-welding portions of bottom mounted instrumentation in reactor pressure vessel, crack propagation crossing dissimilar materials interface between base metal of low-alloy steel and stainless steel cladding is assumed. Fatigue crack growth rate of the cladding is predicted to be captured by that of the base metal because thickness of the cladding is very thin compared that of the base metal. In this study, fatigue crack propagation analyses of cladded plates under cyclic loading were performed using XFEM, which can model crack independently of finite elements, as a preliminary study for crack propagation analyses considering dissimilar materials interface.

One potential challenge to the integrity of a reactor pressure vessel (RPV) of a pressurized water reactor is posed by a pressurized thermal shock (PTS), which is associated with severe cooling of the RPV followed by its repressurization. PTS transients lead to high tensile circumferential and axial stresses in the RPV wall. If the stress intensity factor (SIF) is large enough, a critical crack may grow. Thus, the RPV has to be assessed against cleavage fracture. In this paper, two kinds of embedded cracks, i.e. semielliptical and elliptical crack with depth of 17 mm and length of 102 mm are considered. The extended finite element method (XFEM) is used to model such postulated cracks. The embedded crack with tip in the cladding/base interface causes a high K I . This is due to the stress discontinuities at the interface between the materials. In the FAVOR (probabilistic fracture mechanics code) calculation, for such cracks the closest point to the inner surface is calculated in order to be conservative. However, due to the highly ductile cladding material, it is unlikely for the embedded crack to propagate through the cladding. Thus, it is more appropriate to consider the outer surface point of the crack front. The effect of welding residual stress and cladding/base interface residual stress on the crack driving force is studied. Surface cracks are assumed in the study of residual stresses. Results show that considering realistic welding residual stresses may increase K I by about 5 MPa·m 0.5 , while the cladding/base interface residual stress has a negligible effect on K I . The reason is that the cladding residual stress is only localized to the interface and it decreases significantly through the vessel wall.

An expanding plug wedge test method was developed to determine hoop tensile properties of irradiated fuel cladding in a hot cell [1]. Ring specimens with the plug-wedge insert were fabricated based Finite Element Analysis (FEA) design. After several design and modification iterations, the protocol was validated for M5 and Zr-4 cladding. Results show that tested ring specimens have uniform radial expansion deformation in the gage section. The test data processing procedure was developed based on the energy approach to generate hoop stress-strain correlation. α factor estimated from FEA was used to remove dissipated friction energy. With the proposed method, optimized specimen design, and data processing procedure, the protocol validates that the measured piston load, plug extension, and ring radial expansion can be effectively and accurately converted into the hoop stress-strain curve for clad tubing structural material characterization. Validation proved that the expanding plug wedge test method is robust and can be easily performed.

This paper describes further results from an ongoing study of a simplified engineering model that is intended to account for effects of clad residual stresses on the propensity for initiation of preexisting inner-surface flaws in a commercial nuclear reactor pressure vessel (RPV). The deposition of stainless steel cladding during fabrication of an RPV generates residual stresses in the cladding and the heat affected zone of the under-lying base metal. In addition to residual stress, thermal strains are generated by the differential thermal expansion (DTE) of the cladding and base material due to temperature changes during normal operation. A simplified model used in the ORNL-developed FAVOR probabilistic fracture mechanics (PFM) code accounts for the clad residual stress by incorporating a stress-free temperature (SFT) approach. At the stress-free temperature ( T s-free ), the model assumes there is no thermal strain, i.e., the thermal expansion stresses and clad residual stresses offset each other. For normal cool-down transients applied to the RPV, interactions of the latter stresses generate additional crack driving forces on shallow, internal surface-breaking flaws near the clad/base metal interface; those flaws tend to dominate the RPV failure probability computed by FAVOR. In a previous report from this study (PVP2015-45086), finite element analysis was used to compare the stresses and stress-intensity factors (SIF) during a cool-down transient for two cases: (1) the existing SFT model of FAVOR, and (2) directly applied RPV clad residual stress (CRS) distribution obtained from empirical (hole-drilling) measurements made at room temperature on an RPV that was never put into service. However, those analyses were limited in scope and focused on a single flaw orientation. In this updated study, effects of CRS on the SIF histories computed for both circumferential and axial flaw orientations subjected to a cool-down transient were determined from an extended set of finite element analyses. Specifically, comparisons were made between results from applying CRS experimental data to ABAQUS two-dimensional, inner-surface flaw models and those generated by the FAVOR SFT model. It is demonstrated that the FAVOR-recommended SFT value of 488 °F produces conservatively high values of SIF relative to the use of CRS profiles in the ABAQUS models. For the vessel and flaw geometry and transient under study, the circumferential flaw (360° continuous) required a decrease of SFT down to 390 °F to match the CRS SIF histories. For the infinite axial flaw model, a decrease down to 300 °F matched the CRS SIF histories. Future plans are described to develop more general conclusions regarding the FAVOR model.

X65 steel pipes internally clad with Alloy 625 used in subsea oil extraction are normally welded together with Alloy 625 filler metal. For pipe reeling applications, DNV-OS-F101 requires pipe girth welds to overmatch base metal yield strength with 100 MPa. Since Alloy 625 filler metal does not meet this requirement, Ni-base super alloys 718 and 282 were considered as potential welding consumables for reeling applications. The solidification behavior in weld metal of these alloys diluted with Alloy 625 pipe ID cladding was evaluated using thermodynamic simulations. The response to precipitation hardening by multiple reheat cycles was studied by producing multilayer buildups with cold metal transfer (CMT) and pulsed gas metal arc welding (GMAWp) processes. Weld buildup of Alloy 718 exhibited insufficient hardening response and yield strength, while Alloy 282 met the DNV overmatch requirement. Successful narrow groove welding of X65 pipes with Alloy 282 was performed using CMT process. Welding parameter optimization allowed resolving centerline solidification cracking and lack of fusion defects. The weld metal yield strength was lower than in the multipass buildup, which was attributed to lower number of reheats in groove welding. Meeting the overmatch requirement for yield strength in Alloy 282 groove welds requires further parameter optimization.