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.

A modified 9-Cr alloy was developed by Oak Ridge National Laboratory, in early 1980s, to increase high temperature capabilities of ferritic steels for superheater tubing. The material improved high temperature creep properties by controlling alloying elements and microstructure. The material was added to ASME BPVC in 1983 (thru Code Case 1943) as Grade 91. Higher yield and tensile strength, in comparison to other low-Cr alloy steels (like Grade 22), allowed for fabrication of thinner component wall thickness. This in-turn reduced susceptibility to through-wall thermal stresses during transient events. Directionally this also reduced material costs. Consequently, petrochemical industries have utilized Grade 91 in applications at Heat Recovery Steam Generation units (HRSG), Steam Methane Reformers (SMR), CO2 Boilers and as convection coils in Ethane conversion units. Grade 91 material has complex microstructure and requires careful control of welding parameters to assure crack free welds that provide adequate creep ductility and retain creep strength at high temperatures. The current guidelines documented in API 582 and Technical Report 938 provide limited insights on success factors for weldability. Grade 91 material use has been growing in the recent past Petrochemicals Complex and in offshore applications, at once-throw steam generators (OTSG). The aim of this paper is to share experience on welding parameters. Guidance needs be adjusted to specific projects and repair activities.

Identifying the coupled system natural frequencies and dynamic behavior of systems in the presence of fluid-structure interaction is one of the most important issues in the engineering design of buildings, road vehicles and aircraft. This paper presents an efficient and flexible finite element procedure using fully vectorized codes for the free and forced vibration analysis of a rectangular plate in contact with fluid. The 4-node MITC plate finite element (MITC4) based on the Mindlin plate theory is used to simulate the plate, while the 8-node acoustic pressure element is used to simulate the fluid. The derived system of equations using structural displacements and fluid pressures yields a non-symmetric system of equations. Solving the generalized eigenvalue problem for the non-symmetric system is more computationally intensive compared to solving the generalized eigenvalue problem for symmetric systems. The modal expansion technique is used to reduce the model size. Then the reduced non-symmetric system is symmetrized by right eigenvectors. The Newmark method is used to solve the forced vibration problem of the coupled systems. The effect of the height of the fluid on the natural frequencies is discussed. The natural frequencies and transient responses are in good agreement with those obtained from the commercial finite element software. Moreover, the technique is proved to be effective to solve the coupled system.

The existing Class A metallic materials qualified in the ASME Section III, Division 5 rules for high temperature nuclear reactors do not have optimal corrosion resistance for some reactor coolants such as liquid lead, lead/bismuth eutectic, and molten salts which is a major constraint on long life designs. A near term solution to this limitation is the use of cladded components — overlay the Class A materials with a thin layer of some corrosion resistant material. However, this necessitates the development of a design method for cladded components without requiring long-term testing of clad materials in order to support the near-term deployment of these advanced reactor systems. In two other PVP papers [PVP2020-21469, PVP2020-21493], the development of such design method along with a complete set of design rules and clad selection criteria are presented. However, the developed design rules and clad selection criteria are based on a priori assumption of perfect bonding between the clad and base materials. In practice the clad/base debonding may occur before the end of the design life of the component. Therefore, this paper focuses on developing a general acceptance test for checking whether the mechanical integrity of the clad/ base metal interface will be retained till the end of the design life. This work proposes to perform temperature cycling tests on cladded buttons of 12.7 mm diameter and a few mm thickness. A simple analytical expression is provided to determine the temperature range for the temperature cycling tests so that the clad/base interface shear stress intensity at the edge of the cladded button mimics the maximum interface shear stress intensity experienced by the component during the design transients. The paper also provides a method to determine the clad/base interfacial shear stress from finite element structural analysis of components.

In Japan, Japan Atomic Energy Agency has developed a probabilistic fracture mechanics (PFM) analysis code, PASCAL4, for probabilistic evaluation of reactor pressure vessels (RPVs) in pressurized water reactors (PWRs) considering neutron irradiation embrittlement and pressurized thermal shock (PTS) events. Besides severe PTS events, however, transients associated with normal operations, such as the cooldown and heatup transients associated with reactor shutdown and startup, respectively, should also be considered in the integrity assessment of RPVs in both PWRs and boiling water reactors (BWRs). With regard to a heatup transient, because temperature is at its minimum, and tensile stress at its maximum on the RPV outer surface, outer surface crack and embedded crack near the RPV outer surface should be taken into account. To extend the applicability of PASCAL4, we improved the code to include analysis functions for these cracks. The improved PASCAL4 can be used to run PFM analyses of RPVs subjected to both cooldown (including PTS) and heatup transients. In this paper, improvements made to PASCAL4 are firstly described, including the incorporated stress intensity factor solutions and the corresponding calculation methods for vessel outer surface crack and embedded crack near the outer surface. Using the improved PASCAL4, PFM analysis examples for a Japanese BWR-type model RPV subjected to thermal transients including a low temperature overpressure event and a heatup transient are presented.

High thermal efficiency of Compact Heat Exchangers (CHX) makes them distinctly utile for application to Next Generation Nuclear Plants (NGNPs). The high temperature application and transient conditions of NGNP operation induce stresses in CHX. These induced stresses can be categorized under different classifications based on their cause and location. ASME Sec. III Div. 5 has different analysis methodologies based on failure modes, failure criteria to be assessed, and constitutive relationship considered. The primary objective of this study is to provide a description of the classification of stresses in CHX. Further, evaluation of CHX design is conducted according to simplified analysis methodologies in ASME Sec. III Div. 5: Elastic and Simplified Inelastic Analysis. These simplified analyses are performed following the submodeling technique. At the global level, the channeled core is replaced by an elastic orthotropic core for analysis. At the local level, the stresses and strains for critical regions are determined following the simplified analysis methods. The load controlled stresses are checked against HBB-3220 of ASME Sec. III Div. 5. For the Elastic Analysis Method, strains in critical sections in CHX are checked for thermomechanical cycle against the HBB-T-1320 of ASME Sec. III Div. 5 criteria. For Simplified Inelastic Analysis, critical sections are analyzed for strain limits following HBB-T-1330 of ASME Sec. III Div. 5. The analyses outcomes are compared and results are discussed.

Modifications in aged process plants may subject piping systems to fluid transient scenarios, which are not considered in the primary design calculations. Due to lack of strict requirements in ASME B31.3 the effect of this phenomenon is often excluded from piping structural integrity reassessments. Therefore, the consequences, such as severe pipe motion or even rupture failure, are discovered after modifications are completed and the system starts to function under new operational conditions. The motivation for this study emanated from several observations in offshore oil and gas piping systems, yet the results could be utilized in structural integrity assessments of any piping system subjected to pressure waves. This paper describes how to provide an approximate solution to determine maximum bending stresses in piping structures subjected to wave impulse loads without using rigorous approaches to calculate the dynamic response. This paper proposes to consider the effect of load duration in quasi-static analysis to achieve more credible results. The proposed method recommends application of lower dynamic load factors than commonly practiced values advised by design codes, for short duration loads such as shock waves. By presenting a real-life example, the results of improved and commonly practiced quasi-static analysis are compared with the site observations as well as dynamic analysis results. It is illustrated that modified quasi-static solution shows agreement with both dynamic analysis and physical behavior of the system. The contents of this study are particularly useful in structural strength re-assessments where the practicing engineer is interested in an approximated solution indicating if the design criteria is satisfied.

With the development of pressure vessel industry in China, more and more pressure vessel is widely used in the aerospace, automobile, marine project and other fields. The liquid sloshing in the device could bring adverse effect on the safety operation of the pressure vessel under the extrinsic excitation, such as the impact load results in the failure of device structure. Thus, dynamic analysis of liquid sloshing for the shipboard pressure vessel is significant. According to the little research on the design criteria and the method of the shipboard pressure vessel, this paper focuses on the research of the transient sloshing of the internal liquid in the shipboard pressure vessel. In this paper, during analyzing the liquid sloshing dynamic of the shipboard pressure vessel, an obvious cycle is got. Considering the operation condition of lateral inclination, longitudinal inclination and heave in the complex marine environment, the liquid sloshing character is obtained and compared through the liquid surface height curve and the wall sloshing pressure curve, which has great influence on the piping layout of the pressure vessel. Moreover, it is beneficial to restrain sloshing through installing baffle inside the pressure vessel. This may provide a project reference to liquid sloshing design of the shipboard pressure vessel.

As part of the effort for a nuclear plant to undergo license renewal, the effect of reactor water environment on fatigue life must be addressed for limiting component locations. One method to incorporate the effects of reactor water environment into the fatigue evaluations of metal components is to apply an environmental fatigue penalty factor (F en ) to the partial usage factor obtained from the design fatigue curve for each stress cycle. Fatigue evaluations have historically been performed by assuming that temperature transient loads occur at conservatively high rates to maximize the stress response and corresponding fatigue usage values. However, with consideration of reactor water environmental effects on fatigue, transients with slower rates generally produce higher F en values that could potentially result in higher environmental fatigue usage values than transients with identical temperature changes but faster rates. A generic parametric study was performed in MRP-218 to characterize limiting transient ramp rates with respect to environmental fatigue usage for a range of piping geometry and material configurations. This paper describes the application of the parametric study results to optimize thermal hydraulic and stress response modeling assumptions with respect to transient rates and downstream effects on environmental fatigue results in both design and monitoring fatigue evaluations.

Compact Heat Exchangers (CHXs) have a large number of miniature channels inside their core, which makes them highly thermal efficient and thus, prime utile for Next Generation Nuclear Plant (NGNP) applications. The fabrication of a CHX involves diffusion, brazed or welded bonding of plates to form CHX block with a channeled core. The elevated temperature and transient conditions of NGNP operation may induce excessive strain and creep-fatigue failure in channel ligaments. The primary objective of this study is to evaluate the design of CHX for application to NGNPs, following the ASME Code Elastic Perfectly Plastic (EPP) Analysis criteria in a draft ASME Code Section III, Division 5 and using the currently available Division 5 Code Cases (N-861 and N-862). As global analysis considering channels in the core is computationally intensive, a new analysis method is evaluated. In this method, the global analysis is performed by representing the channeled core by an elastic orthotropic material core. Subsequently, at the local level, EPP analysis is performed using models that include the channels, with thermal and pressure loading conditions. An ASME Draft Code Case is under development for the construction of CHXs. The analysis results are used to assess proposed stress limits and classification for load controlled stresses. For strain limits, the analysis results are evaluated using Code Cases N-861 and N-862 against the strain limit and creep-fatigue damage using the channel level submodel analysis. The applicability of the new analysis method, and use of the analysis results for evaluation against ASME proposed limits for various regions of the CHX are presented and discussed.

Low transient pressures in piping systems are different in many ways to high transient pressures. While high pressures can obviously burst pipes or damage components, low pressures can collapse pipes, pull in environmental contaminants, bring components out of solution, or induce transient cavitation, a particular concern for hydrocarbon liquids. This paper will use examples of computer modeling to reveal how common system events such as pump trips or valve closures induce low pressure transient waves that have potential to be just as destructive as more intuitive high pressure waves. Fluid transient studies and literature often focus on high pressures, or do not clearly demonstrate how liquids with low vapor pressures (such as many hydrocarbons) can be affected. Even discerning a pipe’s negative pressure rating through codes and standards can be a challenge. It is shown that low pressure transients are a potential issue in any liquid system. It is further demonstrated that “Rule of Thumb” or typical simplified calculations are not sufficient to capture these effects, and cannot be used to properly locate and size equipment.

The safe operation of gas pipeline is related to the liveihood of people, in the process of operation, many factors will cause pipeline failure, leakage and accidents and thus impair people’s well-being and property safety. This paper puts forward the transient electromagnetic method for in-service detection of municipal buried gas pipeline, mainly introduces the principle of transient electromagnetic method and the factor analysis of pipeline failure, and comprehensively analyzes the problems existing in the application of transient electromagnetic method in in-service detection of municipal buried gas pipeline combined with two examples of gas pipeline detection. The result shows that transient electromagnetic method is practical in the detection of municipal buried gas pipeline, which can better distinguish corrosion thinning and interference, and provide reliable theoretical data basis for in-service safe operation detection and maintenance of municipal buried gas pipeline.

Configuration of piping-pressure safety valve (PPSV) is widely used but may sometimes show instabilities, which should be avoided or reduced. For exploring the cause of these instabilities, Computational Fluid Dynamics (CFD) provides a powerful tool that can be used not only to reproduce the system-level responses, but also to get the details of the local flows. However, the results of CFD simulations are usually sensitive to their model settings, thus, to ensure accuracy, effects of certain critical model settings (such as time-step size and turbulence model) on transient simulations should be determined. In this paper, a 2-D axisymmetric mesh model is developed first, which can be used to predict the dynamic responds of a PPSV system. With this model, three levels of time-steps are tested. The results indicate that the time-step of 1e−5 is a reasonable choice for transient simulations of the PPSV system. After that, a total of four two-equation turbulence models are used for transient simulations. The results indicate that the SST (Shear stress transport) k-omega model produce the closest results to the experiments. After that, the accuracy of the developed CFD model (with 1e−5s time-step and SST k-omega setting) is verified by experimental tests. The results indicate that the developed CFD model can accurately reproduce the dynamic responds of the PPSV system. Outcomes obtained in this paper can not only provide a reference for the transient CFD model development, but also verify the applicability and high accuracy of 2-D axisymmetric CFD models in PPSV dynamics prediction.

During the operation of nuclear power plants (NPPs), the thermal fatigue is one of the most severe aging factors for the components. Therefore, it is significant to monitor the fatigue degradation of components caused by the thermal variation. Moreover, the fatigue monitor result is widely used in the life extension of NPPs. From this viewpoint, the Nuclear Power Institute of China (NPIC) has developed an advanced fatigue monitor system which can classify the operational transient and calculate the cumulative fatigue usage. In classical fatigue monitor system, two methods are widely used to calculate the cumulative fatigue usage. They are the cycle based method and the stress based method. In the cycle based method, the operational transient is classified into a conservative design transient, and then the remaining times of design transient is used to describe the fatigue condition. Obviously, the cycle based method is fast, but conservative. In the stress base method, the monitored thermal data is used to calculate stress and then do the fatigue analysis. This method is more accurate than the cycle based one, but this method is very time-consuming. In present work, a hybrid fatigue analysis of cycle based and stress based method is proposed. In this method, an off-line stress based fatigue database is developed using finite element method. The cycle based method is adopted to determine the transient effects. However, according to the off-line database, the consumed transient time is no longer an integer, but is a fraction to describe the difference between the cumulative fatigue usage of design transient and the one of operational transient.

The structural integrity of reactor components is very essential for the reliable operation of all types of power plants, especially for components operating at elevated temperature where creep effects are significant and where components are subjected to high-temperature alteration and seismic transient loading conditions. In this article, a molten salt storage tank in high temperature thorium molten salt reactor (TMSR) is evaluated according to ASME-III-5-HBB high temperature reactor code. The evaluation based on 3D finite element analyses includes the load-controlled stress, the effects of ratcheting, and the interaction of creep and fatigue. The thermal and structural analysis and the application procedures of ASME-HBB rules are described in detail. Some structural modifications have been made on this molten salt storage tank to enhance the strength and reduce thermal stress. The effects of ratcheting and creep-fatigue damage under elevated temperature are investigated using elastic analysis and inelastic analysis methods for a defined representative load cycle. In addition, the strain range and the stress relaxation history calculated by elastic and inelastic methods are compared and discussed. The numerical results indicate that the elastic analysis is conservative for design and a full inelastic analysis method for estimating input for creep-fatigue damage evaluation need to be developed.

Transient surge modeling remains to be a specialized task and therefore quite unfamiliar to many engineers and designers. The process for discerning when to conduct a model can be vague. The process for how to conduct a model can be even more unclear and quite daunting. This paper serves as an introductory guide to an engineer or reviewing project manager new to surge modeling. The content within covers a brief overview of the physics behind transients, what information is required to create a computer surge model, the content within a surge report and how to interpret it. Included within this paper are several references and detailed plots derived from actual projects.

The finite element simulation of fracture propagation of BCC metals is challenging, as it needs to incorporate the brittle, ductile-brittle transition and ductile behavior presented by the fracture toughness. In this contribution, we restrict ourselves to the use of XFEM method to simulate the cleavage fracture due to initial flaws in the reactor pressure vessel of a reference design of the two-loop PWR nuclear power plant. A hypothetical large break loss of coolant accident is selected as accident scenario to obtain the loading conditions under which the crack is subjected. The thermal-mechanical calculation is performed using a finite element model of the whole RPV and the initial and boundary conditions are determined from the thermal-hydraulic simulation of the transient in TRACE. The method proposed in this contribution is based on the cohesive segment approach implemented in ABAQUS, which requires the definition of the damage properties of the material. The segment approach does not use the fracture toughness as failure criterion. Instead, it uses a traction separation law that is able to capture the brittle fracture behavior of ferritic steel. The crack propagation in XFEM uses a propagation criterion based on a cohesive damage initiation criterion and a damage evolution model. In order to implement the damage evolution model, the fracture energy release rate is directly related to the fracture toughness. The postulated crack is inserted in a submodel to reduce the computational cost of the calculation. The location of such submodel corresponds to the region of the core that suffers maximum irradiation and is subjected to high tensile stresses due to the cooling plume generated during the transient PTS cooling. The crack propagation analysis of postulated axial crack showed that its propagation happens in axial direction in those finite elements close to the inner surface because the energy release rate G I is larger than the critical energy release rate G IC . At the deepest point of the crack, the stresses in the finite element fulfil the damage initiation criterion but the crack does not propagate in radial direction (G I < G IC ).

SWGs (Spiral Wound Gaskets) are well known as a most used gasket type in bolted pipe flange connections all over the world. Recently, the connections with SWGs have been used under more severe conditions such as higher temperature and pressure, and in addition, the connections have been used in the more long-term application. Thus, it is necessary for plant owners (gasket users) to know the long-term characteristics of the connections with SWGs from a standpoint of integrity in the connections. In this study, the objective is to establish a long-term estimation method of sealing performance for bolted pipe flange connections with SWGs under elevated temperature. The long-term characteristics of pipe flange connections with SWGs are estimated using FEM calculations in which the fundamental mechanical characteristics of SWGs such as compression property under changed temperature is considered, thermal expansion behavior, creep relaxation and sealing performance are taken into considerations. For verification of FEM calculations, the experiments are carried out for the pipe flange connections with SWGs of which the nominal size is ASME class 300 2inch under elevated temperature and internal pressure. The gasket used is chosen as SWG with flexible graphite filler. The change in axial bolt forces and an amount of leakage are measured and the measured results are fairly coincided with the FEM calculation results. In addition, the contact gasket stress in the connection with SWG is shown in 72 months. The FEM calculations are performed as heat conduction problem in transient state.

In order to improve the safety, reliability, and life of diverse structures, the development of effective methodologies for structural health monitoring is critical. Among damage detection techniques, guided ultrasonic Lamb waves are particularly suitable for damage detection applications for plate-like and shell-like structures, such as aircraft wing-box structures, heat exchanger tubing, stiffened panels, and nuclear steam generator tubing, due to their sensitivity to damage. Computational models can play a critical role to study wave propagation for monitoring structural health and develop a technique to detect structural damage. Due to complexity of guided wave behavior, efficient and accurate computation tools are essential to study the mechanisms that account for coupling, dispersion, and interaction with damage. In this study, a numerical technique is presented for guided waves propagation in metallic structure by employing co-simulation using ABAQUS Standard module and ABAQUS Explicit module simultaneously to simulate transient wave propagation from an PZT actuator into a metallic plate. The present co-simulation analysis couples multiphysics (piezoelectric) analysis with transient dynamics (wave propagation) analysis. A numerical test is conducted using a PZT actuator for exciting planar Lamb waves and a sensor for acquiring wave signals. The signals achieved from defected and pristine models by FEA are then compared to identify and detect damage in the structure.

Nuclear Regulatory Commission (NRC) guidance for Subsequent License Renewal (SLR) applicants is provided in the NUREG-2191. Specifically, Section X.M1 of the NUREG-2191 includes guidance for aging management programs (AMP) which has two aspects related to fatigue analyses, one that verifies the continued acceptability of existing analyses through cycle counting and the other that provides periodically updated evaluations of the fatigue analyses to demonstrate that they continue to meet the appropriate limits. In addition, the NUREG-2191 provides requirements to account for reactor water environment by deriving cumulative fatigue usage (CUF) including environmental effects (CUF en ) in component fatigue analyses for a set of sample critical components for the plant outlined in NUREG/CR-6260. Furthermore each applicant is required to determine plant-specific component locations in the reactor coolant pressure boundary that may be more limiting than those considered in NUREG/CR–6260. This paper presents a methodology to identify plant-specific primary equipment component locations that are potentially more limiting than the locations identified in NUREG/CR-6260, through a detailed review and ranking of the analyses of record (AORs) for a plant. The ranking approach implemented in this methodology assesses the level of technical rigor required to derive the CUF values and provides a means to appropriately compare CUF values within the same transient section. This paper also illustrates how this methodology was applied for a recent license renewal application.