A design for flaw placement in a full-scale pipe test was developed to both detect crack initiation and measure crack growth rate upon internal pressurization of a pipe exposed to a sulfide stress cracking (SSC) environment. The objective of this work was to model different sizes of longitudinally oriented, inside diameter (ID) surface flaws and lay them out on the pipe in such a manner that (1) the flaws experience their target stress intensity factor (K) value at a chosen value of internal pressure, (2) the stress interaction between flaws is minimized, and (3) the flaw layout is optimized for detecting both crack initiation and growth using a direct-current (d-c) electric potential (EP) technique. The approach to the flaw design and layout used finite element analysis (FEA) modeling and consisted of optimizing K-profiles. First, the K-profiles were optimized by designing curved-bottom flaws such that the target K along the flaw front occurred along most of the flaw length. Then, stress interactions between the flaws were checked to confirm minimum interactions were achieved and that the proposed flaw layout around the pipe circumference was acceptable. In addition, the FE models were used to predict strains on the pipe outside surface. Finally, global (single large current supply), local (individual small current supplies) and hybrid (individual medium-sized current supplies at larger distance) approaches to the d-c EP measurements were evaluated to select which methodology would be most appropriate to detect both crack initiation and growth of the flaws. The results of the analysis show that optimizing all these design factors provides a solid basis for achieving experimental success.

The development of advanced power plants requires alloys to operate at elevated temperature and pressure for an extended period of time. It is critical to consider creep during the design process to avoid catastrophic failure. Creep rupture data are often not available for desired operating conditions. Accurate extrapolation of creep life is necessary. One of the earliest and most widely used life prediction model is the classic Larson-Miller Parametric (LM) model. Over time numerous time-temperature parametric (TTP) models have been proposed such as Manson-Haferd, Orr-Sherby-Dorn, Manson-Succop, Graham-Walles, Goldhoff-Sherby parametric models. Non-TTP models such as the Wilshire equation is available. The prediction models vary in mathematical form, and number of material constants but shares a common calibration approach. Each model is calibrated against data for every available isotherm. A recently proposed model calibration approach is the parametric numerical isothermal datum (P-NID) method that can be applied to an existing model for improved long-term extrapolation. The P-NID approach is different than the traditional approach as the data are transferred to a datum temperature followed by model calibration against the transferred data at the datum temperature. The calibrated model is then transferred back to the original temperatures. In this study, the P-NID method is applied to the LM model to perform extrapolation for Inconel 617 alloy. Creep rupture data for five isotherms ranging from 800 to 1000°C and stress levels from 9MPa to 170 MPa are used. A detail step by step procedure is provided for the application of the P-NID method to calibrate the LM model (LM-NID). The extrapolation performance of the classic LM and LM-NID models are compared. Normalized Mean Squared Error (NMSE) is used to analyze prediction accuracy. It is observed that the LM-NID model provides a realistic inflection free prediction compared to the LM model. A 10% data-cull from the lowest stress data is performed to assess the reliability of extrapolation. Based on the comparison a recommendation is provided.

Carbon Fiber Reinforced Polymer (CFRP) composites have been used for decades in various industries such as aerospace, oil and gas, mainly due to their high strength-to-weight-ratio and excellent corrosion resistance. However, the use of CFRP in nuclear industry is very limited. Recently, a new ASME Boiler and Pressure Vessel (BPV) Code Case N-871 has been accepted for internal repair of buried Class 2 and 3 nuclear safety related piping using CFRP for Service Levels A, B, C and D. The ASME BPV Code Case N-871 allows the use of both Allowable Stress Design (ASD) and Load and Resistance Factor Design (LRFD) methods to design the CFRP repair system. The USNRC has not yet accepted this code case. In 2016, a relief request was submitted to US Nuclear Regulatory Commission (US-NRC) by Surry Power Station, Virginia Electric and Power Company to perform internal repair of degraded ASME Class 2 and 3 safety related circulating and service water buried piping using LRFD method and CFRP, which was approved by US Nuclear Regulatory Commission (NRC). It is documented in the literature that CFRP materials experience property degradation when exposed to certain environments and creep behavior under sustained loading. From a limited number of tests, it has also been found that the overall strength of CFRP repair decreases with an increase in the number of plies, however, the extent of the strength reduction is not well-supported with test data. The statistical values of CFRP strength can be determined in three ways. The ASME BPV Code Case N-871 recommends using the characteristic value of CFRP strength at a 5 th percentile value with 80% confidence, whereas the A-basis (1 st percentile with 95% confidence) and B-basis (10 th percentile with 95% confidence) values of strength are used in other industries such as in aerospace applications. For a safety related nuclear application, it is therefore important to evaluate how these values compare to one another. The property degradation due to environmental exposure, creep behavior, multi-ply laminate as well as difference in various statistical analysis can be given a general term as strength reduction factor and These factors will have a direct effect on the safety margin of any CFRP repair using the ASD method and hence, it is essential to confirm these factors with an adequate number of experiments. In this work, mechanical testing and statistical analyses have been conducted to identify specific degradation mechanisms in CFRP as well as other design considerations that may affect the effective factor of safety of the repair system. These include the effect of multi-ply laminates, misalignment angle, width of the specimen as well as the characteristic values of the ultimate strength. Finally, the strength reduction factors are determined from test results followed by a discussion of how these factors affect the overall effective factor of safety of the repair system.

In order to ensure the integrity of structures such as gas turbines and nuclear power plants, the materials used should have excellent toughness. Especially in the case of nuclear piping materials applied to leak before break (LBB) design, high toughness materials are used to meet the stringent fracture toughness criteria and integrity must be verified through static J-R curve testing using the compliance method, one of the measurement techniques for fracture toughness. The measured and estimated values for the crack extension length during the test should also match, within a certain tolerance. However, in the case of materials with high toughness, rotation of the specimen becomes significant, because the test is performed until the crack open displacement (COD) is relatively large to ensure sufficient crack extension. In this case, it is not easy to satisfy these conditions due to the rotational effect on the specimen. Even though ASTM E1820 suggests a method for correcting the crack length for the rotational effect on these specimens, it has been found that there are substantial differences for high toughness materials. To solve this problem, a new crack length correction formula considering the rotation effect is proposed. Through analysis of the data from J-R curve testing with this proposed method, it was confirmed that the accuracy of crack extension length estimation is improved compared to the existing method. The proposed method well explains the variation of crack extension length due to rotation and is suitable as a correction equation for rotation of compact tension specimens.

Qualification for weld strength is typically accomplished using cross weld tensile testing. This style of testing only gives the global behavior of the welded joint and limited materials properties, such as elongation at failure and tensile strength of the material where final failure occurs. Qualification for welded structures usually requires the weldment fails in the base metal. Final failure in cross weld tensile tests in the base metal does not provide information about the actual weld metal and heat affected zone properties. There may be weaker points in the microstructure that cannot be identified in a global cross weld tensile test due to being constrained by surrounding microstructures. Additionally, the traditional cross weld tensile test does not quantify how strain accumulates and transfers in the microstructure at various loads. Using Digital Image Correlation (DIC) in combination with tensile testing, local strain of the various microstructures present across the weld was obtained for ferritic to austenitic dissimilar metal welds (DMW), as well as for a typical “matching” ferritic steel filler metal weld with a higher tensile strength than the base metal. This test also showed where and how strain accumulated and transferred during tensile loading of various welded microstructures. Local yield stresses of each region were also obtained. Obtaining such local properties provides insight into design and service limits of welded components in service.

For seismic analyses of linear structural systems including soil-structure systems, the current practice (e.g., the U.S. Nuclear Regulatory Commission (NRC) Standard Review Plan (SRP) (NUREG-0800 [1], “Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition”) and American Society of Civil Engineers (ASCE)/Structural Engineering Institute (SEI) Standard 4-16 [2], “Seismic Analysis of Safety-Related Nuclear Structures”) allows for estimating mean seismic responses by using as few as four or five input time histories. This paper examined whether this practice can achieve a stable mean response by explicitly considering the uncertainty in the Fourier phase spectra of the input time histories and exploring how this uncertainty can affect the coefficient of variation (CV) of the in-structure response spectra (ISRS). ISRS are the response spectra of the seismic response time history at a location in the structure subjected to an input seismic time history. We found that the maximum CV ISRS across all frequencies is around 40% purely due to the uncertainty in the Fourier phase spectra for a typical range of design earthquakes for U.S. nuclear power plants. To estimate a mean ISRS within ±10% of the true mean ISRS, our analyses showed that this level of CV ISRS may require a minimum of 16 input time histories for a confidence level (CL) of 68% and 61 for a CL of 95% for soil-structure systems of low fundamental frequencies. For stiffer systems (for example, with a fundamental frequency of 5 hertz (Hz)), the maximum CV ISRS is about 30%, and thus, the minimum required number of input time histories may be reduced to 9 for a CL of 68% and 35 for a CL of 95%. In summary, the four or five time histories in the current practice may not be sufficient for estimating stable mean responses, especially for soil-structure systems with very low frequencies.

One of the most promising technologies to enable and enhance large survey capabilities for radio astronomy is the use of focal plane phased array antenna feeds, or more simply, phased array feeds (PAFs). PAFs allow for full and continuous coverage of the telescope’s field of view (FoV), and combined with cryogenic amplifiers, can result in survey speed improvements several orders of magnitude better than current multiple-feed-horn cameras. In order to locate cryogenic PAF elements and amplifiers at the telescope focal plane, a radio-frequency transparent vacuum vessel is required. Unlike typical radomes, the transmission properties must be exceptionally good when dealing with weak astronomical signals. The dome must also be sufficiently strong to carry the mechanical load on the vacuum vessel due to atmospheric pressure. Furthermore, the thermal loading on the internal cryogenic stages from the dome must be manageable for the cooling system. We have solved these problems by using a combination of welded polyethylene sheet to maintain the vacuum integrity and a closed-cell rigid foam to transfer the mechanical load to the opposite side of the vessel (a welded aluminum structure). The PAF elements and amplifiers operate at 20 K, while the foam transfers the mechanical load through an 80 K temperature stage, which also serves as a low-temperature radiation shield for the 20 K sections. The poor thermal conductivity of the foam, combined with G10-CR thermal standoffs on the opposite side, ensures the 80 K stage is sufficiently thermally isolated from room temperature conduction. The radiative loading is reduced via the usual employment of multi-layer insulation. In order to facilitate instrument maintenance and future upgrades, a modular PAF element mechanical strategy is employed. The design is such that a PAF-element-amplifier unit can be replaced without accessing the 20 K stage owing to the use of a “cryo-clamp” that uses materials with different coefficients of thermal expansion to tightly hold the units when cold. Cooling is supplied by three two-stage Gifford-McMahon cryo-coolers. This paper presents these design details for the cryostat of the ALPACA (Advanced L-Band Phased Array Camera for Arecibo), an instrument currently being designed and built for the 305 m radio telescope of the Arecibo Observatory in Puerto Rico.

The independent support motion response spectrum method (ISM) is currently used for seismic analysis to calculate the response of multiply supported piping with independent inputs of support excitations. This approach may derive considerable overestimation in the combination of group responses under the absolute sum rule of NUREG-1061 [1]. Then authors have developed an advanced method of the ISM approach named SATH ( S pectrum Method A ssisted by T ime H istory Analysis). In the SATH method, both of floor response spectra and time histories of floor acceleration are used as independent inputs of support excitations. The group responses are summed with correlation coefficients which are calculated by considering each time history of modal response by independent inputs of support excitations. In this paper, the necessity of taking the effects of correlation coefficients for the group responses into account in the ISM approach is examined. The SATH method has advantage to derive a more realistic sum rule of the group responses and applicability for the actual design.

When cylindrical tanks installed in the ground, such as oil tanks and liquid storage tanks, receive strong seismic waves, including the long-period component, motion of the free liquid surface inside the tank called sloshing may occur. If high-amplitude sloshing occurs and the waves collide with the tank roof, it may lead to accidents such as damage of the tank roof or outflow of internal liquid of the Tank. Therefore, it is important to predict the wave height of sloshing generated by earthquake motions. Sloshing is a type of vibration of free liquid surface, and if the sloshing wave height is small, it can be approximated with a linear vibration model. In this case, the velocity-response-spectrum method using velocity potential can estimate the sloshing wave height under earthquake motions. However, if the sloshing wave height increases, the sloshing becomes nonlinear, and necessary to evaluate the wave height using other methods such as numerical analysis. Design earthquake magnitude levels in Japan tend to increase in recent years, long-period components of earthquake wave which act on the sloshing wave height also increase instead of introducing seismic isolation mechanisms. To evaluate load acting on the internal components of cylindrical tanks by nonlinear sloshing, there are few applications which quantitatively evaluated the crest impact load of nonlinear sloshing. In order to evaluate the load acting on the internal components of cylindrical tanks, the range of applicability of the fluid flow analysis method which validated the analysis accuracy of impact load acting on the roof in a simple cylindrical tank in the past study (PVP2019-93442) is extended to cylindrical tanks with internal components.

When cylindrical tanks installed in the ground, such as oil tanks and liquid storage tanks, receive strong seismic waves, including the long-period component, motion of the liquid surface inside the tank called sloshing may occur. If large-amplitude sloshing occurs and the waves collide with the tank roof, it may lead to accidents such as damage of the tank roof or outflow of internal liquid of the tank. Also, there is a possibility that the internal components in the tank may be damaged due to the fluid force generated by the flow of the sloshing. In order to evaluate the load acting on the tank roof, it is considered that the liquid surface shape and the liquid surface velocity are required as input parameters. In order to evaluate the load acting on the internal component in the tank, the flow velocity generated by sloshing is required as an input parameter. If the sloshing wave height is small, these values can be calculated based on the linear potential theory. However, when the sloshing wave height increases, the sloshing becomes nonlinear, and the difference between the nonlinear sloshing behavior and the linear sloshing behavior. Therefore, the method of evaluating nonlinear sloshing behavior is necessary to evaluate the design load of tank under the large sloshing wave height condition. In this paper, new methods of evaluating nonlinear sloshing behavior are proposed for the first-order sloshing mode of a cylindrical tank, which can evaluate the maximum nonlinear sloshing wave height, the nonlinear liquid surface shape, the liquid surface velocity, and the flow velocity. Proposed methods, which consist of simplified equations, are expected to be applied to a new sloshing load evaluation method in primary design. 1

The Original Net-Section-Collapse (NSC) analysis was developed in the 1970s for prediction of the maximum (failure) moment for a circumferential flaw in a pipe, and is used widely in pipe flaw assessments. A large number of past pipe tests show that deep surface cracks can break through the thickness and result in leaks; hence, the maximum moment of that surface-cracked pipe was below the maximum moment for the circumferential through-wall crack with the same length. In these cases, the applied moment has to be increased for the resulting leak to grow as a through-wall crack. Hence, load-controlled leak-before-break (LBB) fracture behavior has been experimentally observed although it is not predictable by the Original NSC analysis. Recently, Original NSC analysis for circumferential surface-cracked pipes under combined bending and axial tension were enhanced through the development of the “Apparent Net-Section Collapse” methodology to explain inconsistencies with the Original NSC. “Apparent NSC” methodology was developed considering surface-cracked pipe test data developed from external (OD) surface-cracked pipe tests conducted at room temperature (RT) with a vast majority conducted under pure bending and unpressurized conditions. Since it is undesirable to have leakage in many applications, the deficiency in the Original NSC analysis was shown experimentally, and the recently developed “Apparent NSC” methodology applied to a carefully planned matrix of pipe and elbow tests conducted on TP304 stainless steel and Alloy600 materials with different flaw dimensions (composed of short and shallow to long and deep surface cracks), in the range of normalized crack depth, a/t = 0.4 to 0.8 and crack length, 2θψ = 90° to 180°. The tests were conducted under conditions similar to a pressurized water reactor (PWR), and consistent with the International Piping Integrity Research Group (IPIRG-2) [1] test conditions, namely a temperature of 550°F (288°C) and an internal pressure of 2,250 psi. The loads corresponding to the surface-crack initiation, maximum load, and leakage events were recorded from each of the surface-cracked pipe and elbow tests. The data were used to understand the predictable nature of the “Apparent NSC” methodology and to develop an understanding of the fracture behavior of surface-cracked pipes leading to correlation of these results to LBB behavior. Further, the results were correlated between the material composition and the variation of the experimental and predicted bending stress from NSC loads to observations from the previous IPIRG-2 program, where the experimental burst loads were characterized with respect to the flow stress assumptions. The material composition such as variation in sulfur content, and the crack-initiation and crack growth based on elastic-plastic fracture mechanics were used to explain the variability of the flow stress assumption when used in a NSC/limit-load type of analysis. The investigation also showed comparison of predictions based on various flow stress (σ f ) definitions assumed using yield and ultimate stresses obtained from the tensile tests conducted on the pipe and elbow materials at 550°F (288°C) and applied to the Original NSC and “Apparent NSC” methodologies. The moment predictions using ASME elbow stress indices (B 2 , C 2 used in design) or the IPIRG-2 parameter (Ψ e c ) for the circumferentially surface-cracked elbows were also compared to the experimental maximum moments for the tested elbows.

Following the tragic events at the Fukushima Daiichi power plant in 2011, priority was given to increasing the accident tolerance of fuel systems for the current fleet of nuclear reactors. These enhanced Accident Tolerant Fuel (ATF) concepts include a wide variety of fuel and cladding materials, both as variants of the current Zircaloy-UO 2 system and also as novel fuel and cladding concepts. In addition to testing at steady-state, prototypic, conditions within a nuclear reactor, performance of these ATF concepts in off-normal and transient conditions must be evaluated. The Transient Reactor Test (TREAT) facility at Idaho National Laboratory’s (INLs) Materials and Fuels Complex (MFC) was restarted in the Fall of 2017 and is well-suited to serve this purpose. September of 2018 marked the first fueled specimen to be tested in TREAT since its restart; testing of fuel specimens has been ongoing since then. Initial fuel tests focused on the traditional Zircaloy-UO 2 fuel system in order to gain a more thorough understanding of operating characteristics of both the test vehicle system and also the interactions between the reactor and the experiment itself. These tests also served to commission new test vehicles using the well-characterized Zircaloy-UO 2 system. The Separate Effects Test Holder, SETH Capsule, is a modular capsule designed such that it can support a wide variety of specimen geometries ranging from prototypic pressurized water reactor (PWR) fuel samples, heat sink based experiments, and more. The capsule itself is an additively manufactured titanium capsule, within which the experimental specimen is loaded. The SETH Phase I series of tests included five individual SETH capsules, each with a single fuel rodlet and instrumentation to measure temperature during irradiation in TREAT. Each fuel rodlet is representative of a fuel rod in a PWR, with UO 2 in Zircaloy-4 cladding. In August of 2019, TREAT irradiated the first ATF candidate fuel, U 3 Si 2 . This marked the first transient test of an ATF concept and is part of a larger campaign that will irradiate a total of four capsules containing ATF concepts. This test campaign, SETH Phase II, built upon the previous SETH Phase I campaign with a nearly identical design except for the fuel rodlet itself. Two of the four SETH capsules contained U 3 Si 2 fuel within Zircaloy-4 cladding, and the other two capsules contained U 3 Si 2 fuel within SiC cladding. This paper reviews the design, fabrication, and assembly efforts resulting in the four qualified SETH capsules for TREAT irradiation of these ATF concepts.

Molten halide salts are being considered as working fluids for nuclear and concentrated solar power applications. High temperature molten fluoride and chloride salts are known to preferentially attack and deplete Cr in alloys, which leads to the use of high-Ni low-Cr alloys in test facilities for advanced molten salt technology. Alloy C-276 is a commercially available Ni alloy that has adequate Cr contents and is qualified to the maximum temperature of 677°C (1,251°F) in the Boiler and Pressure Vessel Code. The alloy has good corrosion resistance to acids, is resistant to stress-corrosion cracking, and has long track records of use in the chemical industry. Therefore, it has been considered as a structural material for test facilities that require operations at 700°C (1,292°F) or greater to develop high-temperature molten salt technology. To meet the requirements, predictions of the Maximum Allowable Stress above the usage temperatures permitted by the Boiler and Pressure Vessel Code were developed with experimental data as an extension to the current code design values. Analysis showed that above current Codified maximum temperature, strength of the alloy is mainly controlled by creep rupture life under the average stress, although the S c creep rate criterion is close to the F avg .S avg rupture criterion. This paper presents the intended test facilities and the design requirements, alloy selection considerations, literature review, data analysis, and proposed allowable stress extension based on some creep test data for C-276 at temperatures greater than 677°C (1,251°F). Further research activities are also briefly mentioned.

Technical requirements for petrochemical reactor steels have proliferated in the last decade. The need to increase economic benefits together with higher operating temperatures and pressures are leading to the construction of higher capacity reactors with thicker walls. Also, more and more severe and sometimes conflicting requirements in these specifications make it difficult, to derivate a steel design, e.g. in terms of chemical composition and processing parameters, for an optimized balance between quality demands and economical aspects. The design of pressure vessel for petrochemical industry is based on mechanical properties and the design method, which are given by the construction code.

Steam methane reforming is the most common method of hydrogen production relevant for plants in the petroleum upgrading, downstream refining, methanol, and ammonia industries. Owner-operators of steam methane reformer furnaces continue to make repair and replacement decisions that involve the cast outlet manifold fittings. One key part of these plans is assessment of the weldability and remaining life of the cast components. The 20Cr-32Ni-1Nb alloy casting materials typically used in the outlet manifolds are usually operated in the low end of their creep temperature range but are subject to metallurgical aging mechanisms which reduce their ductility, weldability, homogeneity, and fracture toughness. This paper covers the practices employed by several owner-users to optimize the lifecycle costs of the outlet manifold castings. These practices include but are not limited to controlled materials specifications, in-situ weldability tests, non-destructive testing in-situ and destructive testing post service, and repair practices such as annealing heat treatments. This paper also includes a limited survey of several owner-users and their fleets of reformer heaters. The details in the survey include the population of affected cast manifold components, alloy grades for the castings and welds, operating temperature ranges, number of startup and shutdown cycles, ranges of time in service, generic design details, and repair case studies. Also discussed are recent improvements in the state of the art for high temperature materials property data-gathering, as well as the structural modeling via Finite Element Methods. These new technologies are opportunities for future work to develop better strategies in the areas of condition assessment, repair planning, and remaining life prediction, taking into account the relevant parameters of installed manifold components, including: specific aging behavior of the casting chemistry, component mechanical design details, as well as the welding and heat treatment parameters during initial fabrication and subsequent maintenance activities.

One of the common assumptions done in Engineering Critical Assessment (ECA) procedures and standard design codes to assess criticality of initial defects is that the metallic material is isotropic and its plastic flow can be well represented by the von Mises yield function for large plastic deformations. This is due to the coupons used for material characterization are usually extracted from the pipe section along a specific orientation (transverse or longitudinal), and thus the material is deemed as isotropic. This factor may limit tremendously the applicability of the ECA for a more general loading case scenario, where the criticality of the initial crack would be required. In this exploratory study, a X65 seam weld pipeline steel is investigated by considering the effect of anisotropy on plasticity and fracture responses of biaxially loaded pipes having different crack lengths, diameter sizes and internal hydrostatic pressures. The results indicate that anisotropic effects become significant in thin-walled pipes than in thicker configurations regardless initial crack size and internal pressure due to the heavy wall-thickness.

Crack assessment for pipe components of a nuclear power plant or oil/gas pipeline is one of the essential procedures to ensure safe operation services. To assess cracked pipes, J -integral has been considered as a theoretically robust and useful elastic-plastic fracture parameter, so that the estimations of J -integral for various pipe geometries, material properties and loading conditions are highly needed. For this reason, many engineering predictive solutions for J -estimations based on finite element (FE) analyses have been developed. Generally, many engineering predictive solutions have been suggested as a tabular-form or closed-form. Among them, the closed-form solution is more preferred than a tabular-form solution for its convenience when many lots of interpolation are required to use it. However, the accuracy of the closed-form solution tends to be significantly reduced as the number of design parameters increases. Moreover, since there is no strict rule to define the form of functions as well, the accuracy of the closed-form solution is inevitably dependent on the rule of thumb. Therefore, it is highly required to suggest a new approach for J -estimation of cracked pipes with various geometries, material properties and loading conditions. In this paper, we propose an efficient approach based on a machine learning technique to estimate J -integral for surface cracked pipes with various geometric sizes and material properties under axial displacement loading condition. Firstly, parametric FE analysis studies were systematically performed to produce the coefficients representing the engineering J -estimation for the corresponding cracked pipe. Secondly, artificial neural network (ANN) models based on deep multilayer perceptron technique were trained based on FE results. The five input neurons (pipe geometries and material properties) and the two output neurons (the coefficients representing the engineering J -estimation) were considered. Lastly, the accuracy of the trained ANN model was studied by comparing to that of the closed-form solution from multi-variable regressions.

Following the ASME codes, the design of pipelines and pressure vessels for transportation or storage of high-pressure hydrogen gas requires measurements of fatigue crack growth rates at design pressure. However, performing tests in high pressure hydrogen gas can be very costly as only a few laboratories have the unique capabilities. Recently, Code Case 2938 was accepted in ASME Boiler and Pressure Vessel Code (BPVC) VIII-3 allowing for design curves to be used in lieu of performing fatigue crack growth rate (da/dN vs. ΔK) and fracture threshold (K IH ) testing in hydrogen gas. The design curves were based on data generated at 100 MPa H 2 on SA-372 and SA-723 grade steels; however, the data used to generate the design curves are limited to measurements of ΔK values greater than 6 MPa m 1/2 . The design curves can be extrapolated to lower ΔK (< 6 MPa m 1/2 ), but the threshold stress intensity factor (ΔK th ) has not been measured in hydrogen gas. In this work, decreasing ΔK tests were performed at select hydrogen pressures to explore threshold (ΔK th ) for ferritic-based structural steels (e.g. pipelines and pressure vessels). The results were compared to decreasing ΔK tests in air, showing that the fatigue crack growth rates in hydrogen gas appear to yield similar or even slightly lower da/dN values compared to the curves in air at low ΔK values when tests were performed at stress ratios of 0.5 and 0.7. Correction for crack closure was implemented, which resulted in better agreement with the design curves and provide an upper bound throughout the entire ΔK range, even as the crack growth rates approach ΔK th . This work gives further evidence of the utility of the design curves described in Code Case 2938 of the ASME BPVC VIII-3 for construction of high pressure hydrogen vessels.

As part of nuclear power plant ageing management, the increased probability of a need of repair welding must be taken into account along with the increase of plant lifetime. An essential prerequisite for successful and safe repair welding is that the applied welding procedures have been properly validated and qualified prior to their use. For instance, if no post-weld heat treatment can be performed and the desired tempering effect has to be based on temper-bead technique, a user needs to scan among several available repair welding procedures. A decision has to be made which of the procedures provides the maximum desired tempering effect with the case in question. This research is a part of a larger experimental effort developing repair welding techniques, and is a part of the Finnish Nuclear Power Plant Safety Research Programme SAFIR2022. The currently studied experimental repair welding case is a low-alloy steel mock-up with an austenitic cladding. Repair welding is assumed to represent a ‘worst-case’ scenario where a postulated linear crack-like defect exists beneath the cladding and might extend across the interface into the reactor pressure vessel steel side. This postulated defect will be removed by machining, and the thereby machined groove will be filled by repair welding using a nickel-base super alloy 52M filler metal by cold metal transfer-gas metal arc welding with a robotic arm. In this paper, different repair welding techniques and alternatives are shortly surveyed based on existing literature. Overall, published documentation was sparse. While only few studies were considered relevant in terms of established links to actual repair cases of under-cladding defects in reactor pressure vessels, others were mainly for modelling and simulation purposes without e.g. cladding groove preparation or the use of irradiation-embrittled material. Most of these procedures were based on the use of nickel-base alloy filler metal in the combination with temper-bead welding technique, with the aim at omitting both preheating and post-weld heat treatment. The main challenge in the repair weld design is to optimise all relevant welding parameters, including the thermal efficiency of temper-bead welding, in order to obtain a sound, defect-free weld with controlled reactor pressure vessel steel heat affected zone maximum hardness. In the simulations presented in the paper, the goal was to compute the resulting deformations, strains and stresses induced by the repair process and make a-priori estimates of the effectiveness of different repair techniques based on the numerical predictions. The numerical analyses allow the comparison of the procedures and enable selecting the one with most efficient combination of weld thermal cycles in terms of tempering and normalisation effects. The prediction of prevailing residual stresses is also important when further application of the component is considered. The paper is followed by Part II, in which the topics of experimental evaluation and material characterization of the repair weld are presented.

Aging management of the existing fleet of nuclear power plants is becoming an increasingly important topic, especially as many units are approaching their design lifetimes or are entering long-term operation. As these plants continue to age, there is an increased probability for the need of repairs due to extended exposure to a harsh environment. It is paramount that qualified and validated solutions are readily available. A repair method for a postulated through cladding crack into the low alloy steel of a nuclear power plant’s reactor pressure vessel has been investigated in this study. This paper is part of larger study that evaluates the current possibilities of such repair welds. The present paper documents the weld-trials and method selection. A parallel paper describes numerical simulations and optimization of weld parameters. The presented weld-trial represents a case where a postulated crack has been excavated and repaired using a nickel base Alloy 52M filler metal by gas metal arc welding-cold metal transfer with a robotic arm. A SA235 structural steel has been used as a base material in this weld-trial. No pre-heating or post-weld heat treatment will be applied, as it would be nearly impossible to apply these treatments in a reactor pressure vessel repair situation. While Alloy 52M presents good material properties, in terms of resistance to environmentally assisted degradation mechanisms, such as primary water stress corrosion cracking, it is notoriously difficult to weld. Some difficulties and challenges during welding include a sluggish weld puddle, formation of titanium and/or aluminium oxides and its susceptibility to lack of fusion defects and weld metal cracking, such as ductility dip cracking and solidification cracking. Moreover, gas metal arc welding-cold metal transfer is not traditionally used in the nuclear industry. Nonetheless, it presents some interesting advantages, specifically concerning heat input requirements and automation possibilities, as compared to traditional welding methods. The mechanical properties, in terms of indentation hardness, and microstructure of a weld-trial sample have been evaluated in this study. The fusion boundary and heat affected zone were the main areas of focus when evaluating the mechanical and microstructural properties. Detailed microstructural characterization using electron backscatter diffraction and nanoindentation were performed across the weld interface. Based on these results, the gas metal arc welding cold metal transfer is seen as a potential high-quality weld method for reactor pressure vessel repair cases.