Pressure Vessel Components Research Articles

Development of electron optical capabilities for manufacturing of large components by electron beam welding

Electron beam welding (EBW) is a fast process that shows promise to be used for high-productivity welding of thick-section components in the nuclear industry. Nuclear AMRC has been working for many years on developing the use of EBW for small modular reactor (SMR) applications, such as joining thick-section pressure vessel components. The major strength of EB is the single-pass process that can radically reduce fabrication times. This is also one of the major issues with EB. The stress on the welding operator is high as any error generally means failure. In chamber EB operation, such as the Pro-beam K2000 system at Nuclear AMRC, once the chamber is closed, the only interaction between the operator and the work piece is through a camera or the electron beam. Tracking features of the material or the joint line is important to gather as much information on the situation as possible. Nuclear AMRC has used the functions of the Pro-beam multimode to gain reliability during welding. The seam tracker relies on electron optical image analysis to assess the deviation of the beam from the joint line, but it is the duty of the operator to assess whether the correction should be applied. The team has demonstrated that based on the results of the seam tracker, more work is necessary before welding. It was also demonstrated that the seam tracker can be further used to align the beam/gun to the joint and assess variations in a gap in the joint or a variation of working distance for further process reliability.

Numerical Study on Fracture Behavior of Printed Circuit Heat Exchanger Core Based on Elasto-Plastic Phase Field Method

Abstract Printed circuit heat exchanger (PCHE) is a type of compact heat exchanger with high thermal efficiency, more compact, and suitable for high pressure and high temperature conditions. The safety assessment of PCHE is very important, especially for PCHE core. Because PHCE core is different from the PCHE headers which are typical pressure vessel components. However, there is no standard design method to ensure the structural integrity for PCHE core. Some failure modes of PCHE core are studied in recent years, but seldom researches have been conducted concerning the fracture behavior of PCHE core. In addition, the phase field method for fracture has developed rapidly in recent years with its advantages in predicting crack initiation, bifurcation, and merging. In this work, the fracture behavior of PCHE core with and without defects is investigated based on elasto-plastic phase field method. The model parameter of Ni-based Alloy for elasto-plastic phase field method is first calibrated by comparing the numerical results with tensile test data of diffusion-bonded PHCE core specimen. Then, the influence of loading directions, defect locations, and defect sizes on the fracture behavior of PCHE core with semicircular and semicircular-rectangular channels are analyzed. Some conclusions are discussed.

“Practical elimination” principle in the Akkuyu Nuclear Power Plant design

The paper considers the implementation of the “practical elimination” principle in the design of the Akkuyu NPP with VVER1200 reactors being under construction in Turkey. The “practical elimination” principle is defined as follows: for accident sequences or phenomena that contribute or lead to unacceptable radiological consequences for the public or the environment, it shall be shown with a high level of confidence that their occurrence is highly unlikely. “Practical elimination” is proved predominantly by results of a Level 2 probabilistic safety assessment. The implementation of the “practical elimination” principle was considered at the level of accident sequences leading to a large radioactive release. It has been shown that each accident sequence leading to an unacceptable release has the probability of occurrence below 4.45Е–8 per reactor per year, while their total probability not exceeding the value of 6.17Е–7 per reactor per year. For the phenomena inside the containment area during severe accidents, including hydrogen detonation, a large thermal explosion, direct containment heating, overpressure in the containment volume, and the containment damage at later stages due to the basement melting through, their “practical elimination” has been demonstrated. The paper also considers specific hard-to-assess scenarios of beyond design basis accidents, for which the applicability of the “practical elimination” principle is assessed as well: a major positive reactivity insertion, a rupture of the reactor pressure vessel and other largescale components, damage of fuel elements in the spent fuel pool, severe accidents with the containment bypass or containment failure, and severe accidents accompanied the means for mitigating with their consequences being unavailable. Criteria have been developed and used for the “practical elimination” assessment. A sensitivity analysis was undertaken as part of the Level 2 probabilistic safety assessment which has shown that estimated values slightly depend on the analytical assumptions, as well as on the random change in the parameters that the affect the progression of severe accidents.

Analysis of fluctuations in ex-core neutron detector signal in Krško NPP during an earthquake

During an earthquake on December 29th 2020, the Krško NPP automatically shutdown due to the trigger of the negative neutron flux rate signal on the power range nuclear instrumentation. From the time course of the detector signal, it can be concluded that the fluctuation in the detector signal may have been caused by the mechanical movement of the ex-core neutron detectors or the pressure vessel components rather than the actual change in reactor power. The objective of the analysis was to evaluate the sensitivity of the neutron flux at the ex-core detector position, if the detector is moved in the radial or axial direction. In addition, the effect of the core barrel movement and core inside the baffle movement in the radial direction were analysed. The analysis is complemented by the calculation of the thermal and total neutron flux gradient in radial, axial and azimuthal directions. The Monte Carlo particle transport code MCNP was used to study the changes in the response of the ex-core detector for the above-mentioned scenarios. Power and intermediate-range detectors were analysed separately, because they are designed differently, positioned at different locations, and have different response characteristics. It was found that the movement of the power range ex-core detector has a negligible effect on the value of the thermal neutron flux in the active part of the detector. However, the radial movement of the intermediate-range detector by 5 cm results in 7%–8% change in the thermal neutron flux in the active part of the intermediate-range detector. The analysis continued with an evaluation of the effects of moving the entire core barrel on the ex-core detector response. It was estimated that the 2 mm core barrel radial oscillation results in ∼4% deviation in the power and intermediate-range detector signal. The movement of the reactor core inside baffle can contribute ∼6% deviation in the ex-core neutron detector signal. The analysis showed that the mechanical movement of ex-core neutron detectors cannot explain the fluctuations in the ex-core detector signal. However, combined core barrel and reactor core inside baffle oscillations could be a probable reason for the observed fluctuations in the ex-core detector signal during an earthquake.

Closed-form solutions for the elastic–plastic buckling design of shell structures under external pressure

This paper deals with the elastoplastic buckling analysis of shell geometries traditionally encountered in pressure vessels subjected to external pressure loading. The objective is to propose in a unified way original analytical closed-form solutions giving rise to reliable results, with a good accuracy and a wide range of applications in a straightforward and low time-consuming way, for efficient dimensioning purposes. The present study is based on the plastic bifurcation theory, classically used in the context of conservative systems, where additional terms are incorporated here due to the follower external pressure acting normally to the deformed surface of the shell still after buckling. Cylindrical shells are first considered, which certainly represent the most fundamental components of pressure vessels in practice. Owing to the external pressure also acting on the closure ends of a cylinder, the influence of the induced axial compression is investigated thereafter. It appears to be non-negligible for short cylinders in elasticity and it becomes very significant in the case of plastic buckling. Then, closed-form solutions for the critical external pressure of a complete sphere are also obtained, both in elasticity and plasticity. Finally, all these new closed-form solutions are validated against reference numerical results obtained through finite element computations, and compared to current design rules.

Numerical and Experimental Investigations of the Influence of Operation on the Technical Condition of Pressure Vessels.

The paper presents the issues related to the design and assessment of the technical condition as well as determination of the residual durability of pressure equipment. Based on a real a example, a liquid nitrogen spherical tank, we present the development and applicability of the method for assessment of the durability of the structure. In terms of the material itself, the authors analyze macroscale (structural) factors of the geometry of the real structure (by 3D scanning: material wear detection, deflections and deformations, etc.) and measured real operational loads to develop an integrated method, including material model and behavior in its operational condition, delivering a useful tool for macroscale structural analyses of the materials under complex load (mechanical, thermal, chemical, etc.). As a result, a detailed analysis of the tank is presented. The paper gives an idea of the method, its development, usefulness, and applicability of the presented approach by indication of the mutual influence of pressure vessel components (e.g., stubs, manholes) and operational loads, which may result in underestimating the strength and durability of the pressure vessels in the design process and during operation.

Creep rupture limit analysis for engineering structures under high-temperature conditions

The efficient and accurate prediction of creep rupture limit poses a huge challenge for high-temperature engineering such as aerospace, nuclear and chemical industries. It is important to investigate the applicability of mainstream assessment approaches and related creep rupture failure mechanisms through theoretical and numerical views. In this study, major creep rupture assessment techniques are comparatively investigated for the first time, including the isochronous stress-strain (ISS) curve-based creep rupture limit analysis, the Omega creep damage model-based creep analysis and the direct method-based creep rupture assessment by an extended Linear Matching Method (LMM). New virtual creep test curves are generated from the Omega creep model and chosen as the unified creep source data to derive the key material parameters used for different methods. For proposing a reasonable strategy for evaluating high-temperature structures in terms of creep rupture, the balance between computational efficiency and accuracy is comprehensively analyzed. Through a practical engineering application of a high-temperature pressure vessel component, a profound insight into the techniques of creep rupture evaluation is delivered from different views. Moreover, several assessment curves are built based on a new understanding of creep rupture failure mechanism, with an effective numerical plan to validate the creep rupture boundary illustrated. It is demonstrated that the LMM direct creep rupture analysis is more suitable for calculating the structural creep rupture limit, with both monotonic and cyclic load conditions considered.

Study on Fracture Assessment for BWR Reactor Pressure Vessel Steel by GTN model

In the JSME fitness-for service code, fracture evaluation methodology based on linear elastic fracture mechanics is provided for reactor pressure vessel (RPV) components by considering brittle fracture due to neutron irradiation in the RPV beltline region. However, in order to make a rational fracture evaluation for the RPV component such as bottom head component of boiling water reactor (BWR), where irradiation embrittlement is not concerned, it is necessary to consider the applicability of the ductile fracture evaluation methodology consistent with material fracture behavior. In this study, the applicability of the damage mechanics analysis method using Gurson-Tvergaard-Needleman (GTN) model, which is a numerical analysis method for ductile failure, was investigated for RPV bottom head component. Mechanical properties tests and fracture tests with plate specimens were performed using RPV law alloy steel (LAS) material. GTN model parameters of RPV LAS material that can reproduce stress-strain curve and load-displacement curve were determined. Fracture analyses of the plate specimens were performed using determined parameters as well. The load-Crack Mouth Opening Displacement (CMOD) curves and fracture behavior of the fracture tests were compared with the analyses results. The analyses predicted the maximum loads of the tests within an error of 4% and the fracture behavior between the tests and analyses was consistent. The results in this study clearly supported the applicability of the GTN model to RPV bottom head ring component.

Stress analysis and applicability analysis of the elliptical head

As the main pressure components of pressure vessels, the mechanical performance of cylinders and heads affects the normal operation of pressure vessels. At present, no unified theoretical formula exists for the connection region between an elliptical head and the cylinder. Therefore, the authors consider the standard elliptical head as the research object. First, the theoretical stress calculation formula is deduced according to the deformation continuity equation. Second, the stress is experimentally measured using an internal-pressure thin-walled-vessel stress measurement device, and the theoretical and experimental stress values in the discontinuous region between the elliptical head and cylinder are analysed and compared to verify the accuracy and applicability of the theoretical stress calculation formula. The results show that the theoretical stress calculation formula in the discontinuous region between the elliptical head and cylinder is valid. By comparing and analysing the theoretical and experimental stress values, the accuracy and applicability of the theoretical stress calculation formula in the discontinuous region are verified. The findings can provide guidance for the stress measurement of internal-pressure vessels.

Accelerated Creep Test Qualification of Creep-Resistance Using the Wilshire–Cano–Stewart Constitutive Model and Stepped Isostress Method

Abstract In this study, a qualification of accelerated creep-resistance of Inconel 718 is assessed using the novel Wilshire–Cano–Stewart (WCS) model and the stepped isostress method (SSM) and predictions are made to conventional creep data. Conventional creep testing is a long-term continuous process; in fact, the ASME B&PV III requires that 10,000+ h of experiments must be conducted to each heat for materials employed in boilers and/or pressure vessel components. This process is costly and not feasible for rapid development of new materials. As an alternative, accelerated creep testing techniques have been developed to reduce the time needed to characterize the creep resistance of materials. Most techniques are based upon the time-temperature-stress superposition principle that predicts minimum-creep-strain-rate (MCSR) and stress-rupture behaviors but lack the ability to predict creep deformation and consider deformation mechanisms that occur for experiments of longer duration. The SSM has been developed, which enables the prediction of creep deformation response as well as reduce the time needed for qualification of materials. The SSM approach has been successful for polymer, polymeric composites, and recently has been introduced for metals. In this study, the WCS constitutive model, calibrated to SSM test data, qualifies the creep resistance of Inconel 718 at 750 °C and predictions are compared to conventional creep testing data. The WCS model has proven to make long-term predictions for stress-rupture, MCSR, creep deformation, and damage in metallic materials. The SSM varies stress levels after time interval adding damage to the material, which can be tracked by the WCS model. The SSM data is calibrated into the model and the WCS model generates realistic predictions of stress-rupture, MSCR, damage, and creep deformation. The calibrated material constants are used to generate predictions of stress-rupture and are postaudit validated using the National Institute of Material Science database. Similarly, the MCSR predictions are compared from previous studies. Finally, the creep deformation predictions are compared with real data and is determined that the results are well in between the expected boundaries. Material characterization and mechanical properties can be determined at a faster rate and with a more cost-effective method. This is beneficial for multiple applications such as in additive manufacturing, composites, spacecraft, and industrial gas turbines.

Effect of stress stiffness on elastic-plastic strain range

Many pressure vessel and piping components have to withstand high internal pressures and are therefore thick-walled so that geometric effects such as stress stiffening need not be accounted for. However, thin-walled or moderately thick structures may be sensitive to these effects. Design Codes such as the ASME Boiler and Pressure Vessel Code usually provide little guidance on when they are to be accounted for. In the opinion of the authors, this effect deserves more attention. Therefore, the purpose of this paper is to illuminate the effect of stress stiffening by investigating some examples, with particular attention to elastic-plastic strain ranges and the plastic strain range enhancement factor Ke used for fatigue analyses.

Statistical investigation of the weld joint efficiencies in the repaired WWER pressure vessel

From a fatigue point of view, the second most important WWER (Water-Water Energetic Reactor) pressure vessel component (after closure studs) are the primary nozzles. These parts are included in the in-service inspection performed every 4 years when non-destructive testing techniques like ultrasonic, eddy-current and dye-penetration methods are applied. After the determination of the severity of fatigue crack, the progress of initial defects over time can be prevented by repair welding. Purposedetermination of values of the eight main mechanical properties of the welded nozzle after major repair, evaluation of statistical data for reliability and recommendations for the maintenance. Study designthe examination of the nozzle Du-500 revealed the presence of perpendicular crack on the outside surface in the base metal. The crack in the base metal has an angle of 72° with the horizontal plane which was repaired at normal (T = 20 °C) and elevated (T = 350 °C) temperatures by (i) automatic arc welding under the flux layer AN-42, and (ii) ceramic flux. The analysis was performed with 15 different batches of welded steel specimens and every single parameter each time was established by using 45 repetitions of the test procedure for the same type of specimen. Findingsthe study describes the specific case of defected nozzle's housing that was made of steel 15Cr2MoVA and after the crack repair along the length of nozzle diameter could potentially be further exploited. Value – the analysis of the welded specimens showed good agreement between the empirical and theoretical curves and can be used in lifetime and fatigue life predictions for WWER pressure vessels still applicable in power generation industry.

Wire-arc additive manufacturing and post-heat treatment optimization on microstructure and mechanical properties of Grade 91 steel

Wire-arc additive manufacturing (WAAM) of Grade 91 steel (P91) is demonstrated for the first time. This alloy has been widely used in pressure vessel components of thermal and nuclear power plants. The post-WAAM heat treatment process is critical to the anisotropy elimination and property improvement, which is designed and optimized with the help of the CALPHAD (CALculation of PHAse Diagrams)-based thermodynamic modeling in this work. The microstructure and mechanical properties of the as-built and post-heat-treated WAAM P91 steel are studied. The results show that the as-built P91 steel fabricated by WAAM has a much higher strength than traditional wrought material, with an ultra-fine and high-density dislocation embedded martensite lath and quantities of δ-ferrite. The heterogeneous structure resulting from the WAAM process is removed after homogenization at 1200°C for 2 h. The microstructure is optimized with the very fine martensite lath and rational prior austenite grain size (PAGS). The segregated MX observed in the matrix of the as-built material is dissolved back to the matrix during homogenization, which is beneficial to the dispersive precipitation of M23C6 and MX in the grain boundaries and matrix during the following aging. By optimizing homogenization (1200°C for 2 h) and aging (760°C for 2 h) heat treatments, the WAAM P91 steel demonstrates excellent mechanical properties with the ultimate tensile strength of 774 MPa, yield strength of 686 MPa and ductility of 19.4%, which far surpasses the mechanical properties of standard wrought P91 steel used in the industry. The strengthening mechanism is elucidated to be the synergistic effect of MX and M23C6 precipitation.

Developing “Design by Analysis” Methodology for Windows for Pressure Vessels for Human Occupancy

Abstract The ASME pressure vessels for human occupancy (PVHO) codes and standards are engineering standards developed to provide a reliable design method for pressure vessel windows. This empirical method is based primarily on years of government-sponsored testing and development and does not directly use engineering theory. This empirical algorithm makes it challenging to revise without additional large-scale physical testing. The industries using the PVHO code need a way to incorporate advances in material science, manufacturing technology, and overall engineering advances without spending years in code case review. Verification and validation techniques, coupled with stochastic finite element analysis (FEA) to address operational variables, can be the basis for a “design by analysis” method to complement the existing testing requirements to produce a full engineering package consistent with other pressure vessel and pressure vessel component design. A design method sufficiently reliable for PVHO could be used in other applications.

Safety assessment of pressure vessel components for Reaktor Daya Eksperimental

The 10 MW Indonesia’s Reaktor Daya Eksperimental (RDE) which is to be constructed in Serpong Nuclear Zone, Puspiptek and was firstly developed in 2014 is expected to be operational in 2022/2023. The RDE is expected to be a landmark to Indonesia for a nuclear power technology provider in the imminent futurity. Till the end of 2018, the RDE is focused on the basic design including its reactor pressure vessel (RPV). This research paper presents mechanical stresses assessment on basic design of RPV structure components for RDE in ensuring the safety design targets of RDE which are no dangerous radioactive release to the workers, populations, and environments even in the worst accident. Results are presented from the structural stresses assessment under following conditions such as thermal loading, operating pressure, and mechanical loads. The geometric models and dimensions of pressure vessel components within the basic design calculations had been performed by using computer code RPV_RDE.exe in the previous work. These components were designed in accordance with ASME code specification for SA516-70 carbon steel plate. In this work, finite element analysis is applied to assess the occurred structure mechanical stresses upon those RPV components. This work recommends that the RPV components of RDE are safe in the basic design calculations by analysis approach as described in model simulations.

The Development of Reinforcement Pad Design for Pressure Vessel

This paper reviews research from difference researchers on pressure vessel component particularly reinforcement pad or repad design. Present study includes the history of pressure vessel and background of famous pressure vessel code American Society of Mechanical Engineers Boiler and Pressure Vessel Code establishment. Purpose of present research is to study the development repad design and the application repad on pressure vessels. Literatures from other researches on various repad design carried out by experimental and finite element analysis were discussed in present study.

Buckling Analysis of Circular Shells for RDE Reactor Pressure Vessel by Finite Element Study and Analytical Model

One of the basic design considerations for circular shells of Reactor Pressure Vessel (RPV) components of Reaktor Daya Eksperimental (RDE) is buckling mode shape of geometric structure imperfections. Geometric structure is prone to a large number of imperfections due to the manufacturing difficulties. Moreover, determination of critical loads of circular shell structure can be properly approximated only through randomness in the geometry. As at the basic design phase this is possible and expensive to measure the geometric structure imperfections in situ or in laboratory, a simulation of computer code should be provided. In this paper, a contribution is made to numerical simulation by finite element study and analytical model by Monte Carlo technique approach of such circular shells for RPV components. For this reason, computer code SolidWorks is used in the implementation finite element analysis and Fortran code is developed for analytical model analysis. The main objective of this investigation is to provide a comprehensive analysis of unstable structure due to the buckled geometric imperfection in the design calculations involved. A method of solution attempts to use the linearized equilibrium equations to interpret geometric imperfections as structure of circular shells associated with the RPV components of RDE. A number of calculation results carried out on a computer code are presented to illustrate the design of RPV components that are safe from the buckled shape failure. It is found that both of the computer code results are very similar with greatest difference value of vibrational amplitude of 0.00699 % and smallest value for buckling load factor of 12.51 on upper circular shell under 750 °C conditions.

A knowledge-based material selection system for interactive pressure vessel design

Continual introduction of new materials and improvements in existing materials increase the variety of materials that can be used for pressure vessel components. Among wide variety of materials, the most suitable one must be selected for a component by matching its functional requirements with various available materials’ specifications. This study proposes an interactive knowledge-based decision support system for selecting the most suitable material for a given pressure vessel component and its working environment. The developed decision support system, namely Pressure Vessel SELection (PVSEL), consists of two separate phases. In the first elimination phase of PVSEL, the user obtains a feasible set of alternative materials by answering various questions and providing lower-limit values at materials’ critical specifications. PVSEL, then, uses a ranking phase which uses ELECTRE, TOPSIS and VIKOR methods to rank the feasible materials. In the second phase, each alternative material’s ranking is determined by combining its performance values at weighted critical specifications (selection criteria), which are considered as important in meeting the functional requirements of the component. Usage of PVSEL is illustrated in the paper and the results show that the proposed PVSEL is an effective selection tool and provides meaningful results for the designers.

Effect of multiple passes on Lüders/yield plateaus, microstructure and tensile behaviour of narrow-gap thick-section weld plates

During nuclear reactor pressure vessel manufacturing, conventional arc welding processes are often employed involving multiple passes to join thick sections. Such thick-section welding may cause variable through-thickness metallurgical and mechanical properties due to variable heat input, particularly in the heat-affected zone (HAZ) and weld metal. Structural steels for pressure vessel components can be selected based on properties such as the Luders plateau and yield/tensile (YS/UTS) ratio in order to assure extended service lives. In the present work, through-thickness tensile properties of narrow-gap (NG) welds performed on 78-mm-thick SA533 plate welded according to ASME IX with 58 passes of gas tungsten arc welding (NG-GTAW) and 38 passes of narrow-gap submerged arc welding (NG-SAW) are examined and compared. The Luders plateau strength, length and velocity were determined along with yield, YS/UTS ratio, ultimate tensile strength and elongation. These properties were analysed in connection with metallurgical and microstructural properties for both welding approaches and established the correlation among them in both cases with the mechanical properties and Luder behaviour. Multiple thermal cycles have tempered and refined the microstructure and hardness in weld and HAZ, which pretends to be a heat treatment of weldment and promotes the formation of Luders plateau. The YS/UTS ratio and Luders plateau behaviour has been evidenced to be influenced by large numbers of thermal cycles. Additionally, the heat input is also vital concern as it is observed that the Luders plateau stress is more in GTAW process than the SAW. However, Luders plateau length is significant in SAW than GTAW process. The implication of these variations on structural integrity aspects and the correlation of mechanical-microstructural properties have been justified in the present work.

Predicting Leak Rate Through Valve Stem Packing in Nuclear Applications

Leaking valves have forced shutdown in many nuclear power plants. The myth of zero leakage or adequate sealing must give way to more realistic maximum leak rate criterion in design of nuclear bolted flange joints and valve packed stuffing boxes. It is well established that the predicting leakage in these pressure vessel components is a major engineering challenge to designers. This is particularly true in nuclear valves due to different working conditions and material variations. The prediction of the leak rate through packing rings is not a straightforward task to achieve. This work presents a study on the ability of microchannel flow models to predict leak rates through packing rings made of flexible graphite. A methodology based on experimental characterization of packing material porosity parameters is developed to predict leak rates at different compression stress levels. Three different models are compared to predict leakage; the diffusive and second-order flow models are derived from Naiver–Stokes equations and incorporate the boundary conditions of an intermediate flow regime to cover the wide range of leak rate levels and the lattice model is based on porous media of packing rings as packing bed (Dp). The flow porosity parameters (N, R) of the microchannels assumed to simulate the leak paths present in the packing are obtained experimentally. The predicted leak rates from different gases (He, N2, and Ar) are compared to those measured experimentally in which the set of packing rings is mainly subjected to different gland stresses and pressures.