Evaluation of Repair Welding Influence on GH3535 Superalloy Microstructure and Mechanical Properties

GH3535 supperalloy, whose grade of ASME is UNS N10003, is currently considered as a candidate material for solid-fuel and fluid-fuel molten salt reactor in china. During the development of procedures for welding GH3535 superalloy, consideration should always be given to the possibility that repair welding may be necessary. This paper presents weld repairs of GH3535 alloy rolled plates using gas tungsten arc welding with filler metal. The purpose of this work is to evaluate the low heat input process for weld repair of GH3535 alloy plates about the microstructure features and mechanical properties. The results demonstrated that sound joints without defects could be obtained after weld repairs. Due to repair thermal cycles on the original weld seam, the size of carbide precipitate became large, but repair welding is found to cause no decrease in short-term time-independent strength.

Two Case Studies in 3-D Residual Stress Prediction for Nuclear Power Applications

Residual stress prediction contributes to nuclear safety by enabling engineering estimates of component service lifetimes. Subcritical crack growth mechanisms, in particular, require residual stress assumptions in order to accurately model the degradation phenomena. In many cases encountered in nuclear power plant operations, the component geometry permits two-dimensional (i.e., axisymmetric) modeling. Two recent examples, however, required three-dimensional modeling for a complete understanding of the weld residual stress distribution in the component. This paper describes three-dimensional weld residual stress modeling for two cases: (1) branch connection welds off reactor coolant loop piping and (2) a mockup to demonstrate the effectiveness of the excavate and weld repair process.

Flaw Stability Analysis of Semi-Elliptical Surface Cracks in Canisters Under the Influence of Welding Residual Stress

Many stainless steel canisters for the dry storage of spent nuclear fuel are located in coastal regions. Because the heat treatment for relieving the welding residual stress is not required during fabrication, these canisters may be susceptible to chloride induced stress corrosion cracking due to the deliquescence of chloride-bearing marine salts or dust that enter the overpack system and deposit on the canister external surface. The NDE techniques and the associated delivery system are being developed to conduct periodic inservice inspections. The acceptance standards are needed to disposition findings should flaw-like indications be found. The instability crack lengths and depths for these flaws in the form of semi-elliptical shape near the welds are determined with R6 procedure. The cracks are subject to the canister design pressure and handling loads as well as the estimated welding residual stresses.

A Monte Carlo Implementation of the James-Ford-Jivkov Microstructurally Informed Local Approach Applied to Predict Fracture Toughness Across Irradiation Conditions

The need to predict changes in fracture toughness for materials where the tensile properties change through life, such as with irradiation, whilst accounting for geometric constraint effects, such as crack size, are clearly important. Currently one of the most likely approaches by which to develop such ability are through application of local approach models. These approaches appear to be sufficient in predicting lower shelf toughness under high constraint conditions, but may fail when attempting to predict toughness in the transition region, for low constraint geometries or for different irradiation states, when using the same parameters, making reliable predictions impossible. Cleavage toughness predictions in the transition regime are here made with a stochastic, Monte Carlo implementation of the recently proposed James-Ford-Jivkov model. This implementation is based around the creation of individual initiators following the experimentally observed distribution for specific reactor pressure vessel steel, and determining if these initiators form voids or cause cleavage failure using the model’s improved criterion for particle failure. This implementation has been presented previously in PVP2015-45905, where it was successfully applied across different constraint conditions; in the work presented here it is applied across different irradiation conditions for a second type of steel. The model predicts the fracture toughness in a large part of the transition region, demonstrates an ability to predict the irradiation shift and shows a level of scatter similar to that observed experimentally. All results presented, for a given material, are obtained without changes in the model parameters. This suggests that the model can be used predicatively for assessing toughness changes due to constraint-, irradiation- and temperature-driven plasticity changes.

Characterizing Successful Robotic Insertion and Removal From a Dry Storage Cask Using Peg-Like Jamming and Wedging Analysis

This paper describes the derivation and experimental validation of geometric equations that govern insertion and extraction of a robotic inspection system that operates in gaps around vertical dry storage casks. During insertion, a robotic system may become jammed due to unbalanced forces acting on the robot, or wedged due to over-sized robot geometry. The robot must be removable by a tether in the event of power loss. Assuming simplified geometry and a quasi-static approach, the problem is modeled using a two-dimensional representation in which the robot is assumed to be rigid with equal weight distribution and a constant friction coefficient between surfaces. Equilibrium equations are derived from a modified peg-insertion formulation, allowing calculation of the maximum size of the robot and angle of insertion as a function of inspection gap geometry and friction. Experimentation tested the derived relationships using varying robot dimensions in a 1:1 scale mock-up of the overpack-to-canister gap space of a nuclear dry storage container. Experimental data confirmed that the modifications of the typical peg-insertion predicted successful insertion and extraction better than unmodified equations. The error between the model and experimentation had a mean and standard deviation of 4.4 and +/− 0.53 degrees.

3D Residual Stress Simulation of an Excavate and Weld Repair Mockup

This paper presents predictions of weld residual stresses in a mockup with a partial arc excavate and weld repair (EWR) utilizing finite element analysis (FEA). The partial arc EWR is a mitigation option to address stress corrosion cracking (SCC) in nuclear power plant piping systems. The mockup is a dissimilar metal weld (DMW) consisting of an SA-508 Class 3 low alloy steel forging buttered with Alloy 182 welded to a Type 316L stainless steel plate with Alloy 82/182 weld metal. This material configuration represents a typical DMW of original construction in a pressurized water reactor (PWR). After simulating the original construction piping joint, the outer half of the DMW is excavated and repaired with Alloy 52M weld metal to simulate a partial arc EWR. The FEA performed simulates the EWR weld bead sequence and applies three-dimensional (3D) modeling to evaluate the weld residual stresses. Bi-directional weld residual stresses are also assessed for impacts on the original construction DMW. The FEA predicted residual stresses follow expected trends and compare favorably to the results of experimental measurements performed on the mockup. The 3D FEA process presented herein represents a validated method to evaluate weld residual stresses as required by ASME Code Case N-847 for implementing a partial arc EWR, which is currently being considered via letter ballot at ASME BPV Standards Committee XI.

Tensile Characterization of Narrow Gap Bimetallic Weld Ferritic-Austenitic Steel With a New Measurement System for Tensile Testing

Ferritic steel 2 ¼ Cr is a candidate material for future pressure component in nuclear fields. In order to validate this choice, it is necessary, firstly to verify that it is able to withstand the planned environmental and operating conditions, and secondly to check if it is covered by the existing design codes, concerning its procurement, fabrication, welding, examination methods and mechanical design rules. A large R&D program on 2 ¼ Cr steel has been undertaken at CEA and Areva in order to characterize the behavior of this material and of its welded junctions. In this frame, a new measurement system for tensile testing was developed in the LISN laboratory of the CEA (French atomic commission), in order to characterize the local behavior of the material during a whole tensile testing. Indeed, with the conventional measurement system (typically an extensometer), the local behavior of the material can only be determinate during the stable step of the testing. So, usually the behavior of the material during the necking step of the step is unknown. This new measurement is based on the use of some laser micrometers which allow measuring the minimum diameter of the specimen and the curvature radius during the necking phase with a great precision. Thanks to the Bridgman formula, we can evaluate the local behavior of the material until the failure of the specimen. This new system was used to characterize the tensile propriety of a bimetallic welded junction of 2 ¼ Cr steel and austenitic stainless steel 316L(N) realized with inconel filler metal. These works lead to propose a tensile curve for each materials of the welded junction at room temperature and the effect of postweld heat treatment.

Introduction of the Scenario of How to Use Low Alloy Steels Safely in High Pressurized Hydrogen Gas

In 2012, the Japanese regulation for selecting SUS316 austenitic stainless steel with a specific Ni equivalent (SUS316 and SUS316L can be used in the temperature ranges between −45 and 250 °C for a Ni equivalent of ≧28.5%, between −10 and 250 °C for a Ni equivalent of ≧ 27.4%, and between 20 and 250 °C for a Ni equivalent of ≧ 26.3%) as an appropriate material available in hydrogen refueling stations (HRSs) that provide 70 MPa fueling to fuel cell vehicles (FCVs) was updated with the support of NEDO (New Energy and Industrial Technology Development Organization) Program Phase 1 [1].

Characterisation of Microstructure and Properties of a Transition Weld

Transition welds represent a challenge for the assessment of structural integrity of nuclear plant due to the complexity of the microstructure, properties and local stress state. This paper presents the initial findings of a study aimed at characterising the local microstructure and properties of a transition weld between SA508-4N ferritic steel and SS316LN austenitic stainless steel using a nickel-base filler of Alloy 82. The local microstructures and local composition of the material interfaces are characterised using backscattered electron imaging and Energy-dispersive X-ray spectroscopy. The ferritic steel shows significant grain refinement in the heat affected zone compared to the base metal. This refinement is also observed in the heat affected zone of the austenitic stainless steel although not as significant. Micro-hardness testing has also been incorporated to provide an indication of the influence of local microstructure on flow properties across the weld region. The results indicate a hardness range of between 180–340HV across the weld with the highest value in the heat affected zone of the ferritic steel and the lowest in the austenitic stainless steel. Yield and flow properties derived from flat transweld tensile tests incorporating digital image correlation are related to the micro-hardness results and microstructural characterisation, and an initial assessment of the fracture mechanism performed using fractography.

Damage Characterization of Heat Resistant Steels After Long Service in Reforming Plants

ASTM A 297 grade HP steels are widely employed for radiant tubes in reforming furnaces: this class of heat resistant alloys shows high creep and corrosion resistance, ensuring good performances in extreme pressure and temperature conditions. The typical microstructure of such materials is an austenitic matrix surrounded by a network of interdendritic carbides, which contain chromium and other carbide forming elements, namely Nb, Ti, W, Zr and Y. During long service life, these high strength materials may suffer aging or even severe damage, especially when process conditions allow coke deposition, or maintenance procedures are not carried properly. Service aging can be summarized, for HP steels, in terms of microstructure degradation: coalescence and coarsening of interdendritic precipitates, precipitation of secondary carbides in the austenite matrix and transformation of niobium-rich carbides in the G-phase silicide are the typical phenomena occurring on the microstructure of these alloys during service. Carburization can also occur in radiant tubes, since their inner wall side is exposed to hydrocarbon-rich process fluids: carbon diffuses into the metal matrix, causing massive precipitation of chromium-rich carbides. The alloy corrosion resistance is then reduced, resulting in surface attack, cracks development and a general wastage of the material. Furthermore, the high temperatures, which tubes are exposed to, can also induce creep, especially if a local tube overheating occurs: cavities and microcracks, mainly localized at precipitates, are the typical evidences of creep damage on HP steels. The present work is aimed on the damage characterization of several radiant tubes in HP alloys, after long term service aging in reforming plants. We employed optical and electron microscopy, EDX elements mapping and mechanical tests, in order to characterize and evaluate the various damages affecting the alloys. Microstructure evolution has been detected in all the analyzed tubes, but we found that such a phenomenon was strictly influenced by the chemical composition of each alloy, so that in presence of small amounts of titanium and tungsten, the chemical evolution of the secondary phases was appreciably contained. Creep also was observed in all the investigated tubes and its extent was found to be related to both alloy composition and process conditions. These latter have assumed to be the main driving factor for carburization, since we observed that slight differences in temperature, pressure, chemical composition of the process fluid and tube maintenance dramatically conditioned the performances of each tube. Massive precipitation and material degradation, in fact, were found in some cases, but, on the other side, no appreciable evidence of carburization damage was observed on other cases.