Pressure Tube Research Articles

Numerical Study of the Hydride Embrittlement in Zirconium Alloy using XFEM

A numerical model for hydride embrittlement in Zirconium alloy (Zr–2.5Nb) is developed utilizing the extended finite element method (XFEM). Hydride embrittlement reduces the ductility and failure time of a metal/alloy. During hydride embrittlement, stress-directed hydrogen diffusion, metal-hydride phase transformation, mechanical deformation, and hydride precipitation occur simultaneously. The present model incorporates all these processes and is able to predict the hydrogen concentration and the hydride fraction distribution under any externally applied stress field. In this work, both the steady and transient hydrogen diffusion cases are evaluated. Further, the XFEM is utilized to develop a model of hydride embrittlement in the presence of a crack. The first step of the hydride embrittlement process is the diffusion of hydrogen. According to Fick’s law of diffusion, hydrogen diffusion is directly dependent on hydrostatic stresses and hydrogen concentration gradient under external stresses. The next step is the hydride precipitation in hydride embrittlement, where the expansion of material takes place that changes the hydrostatic stress field. Thus, studying the effect of precipitation of hydride on hydrostatic stresses is essential. Moreover, the process of hydride embrittlement is highly influenced by residual stresses in the structure. Hence, the effect of residual stress present in the zirconium alloy pressure tube (PT) is also evaluated. The results indicate that the residual tensile stresses contribute to the growth of hydride, which will reduce the material failure time.

Hydrogen diffusion in Zr-2.5Nb pressure tubes specimens between 300°C-400°C by in-situ neutron imaging experiments

Zirconium (Zr) alloys are widely used in nuclear power plants as fuel cladding and are susceptible to hydrogen (H) degradation. For long operational service, Zr-based components can suffer a mechanism known as Delayed Hydride Cracking (DHC) associated to an increase of the crack propagation velocity by the re-orientation and precipitation of Zr hydride. In this process, the H mobility has a great influence. In the present work, the isothermal diffusion of H in Zr-2.5%Nb specimens obtained from a CANDU pressure tube were studied at consecutive temperatures of 300°C, 350°C, 375°C and 400°C. H content and mobility were quantified by in-situ neutron imaging experiments performed on ANTARES, the cold neutron imaging facility of FRM II. The time evolution of the H concentration across the specimen allowed the determination of diffusion coefficients, and an assessment of the limitations of existing models commonly used to describe H diffusion.

Development of in-situ Delayed Hydride Cracking tests using neutron imaging to study the H redistribution in Zr-2.5%Nb

Delayed Hydride Cracking (DHC) is a failure mechanism that occurs in Zr alloys under certain conditions of hydrogen concentration, temperature and stress gradient. In service, hydrogen produced by corrosion reaction can be incorporated in Zr alloys and if the solid solubility is exceeded, hydrogen precipitates as zirconium hydride. The presence of a stress concentrator, such as a crack, generates the hydrogen diffusion and precipitation to the high stress zone beginning the DHC process. In this work, in-situ DHC tests in air at 250°C were performed at ANTARES, the neutron imaging facility of the FRM-II reactor. Samples of Zr2.5%Nb produced from extruded tubes and pressure tubes were studied using a stress rig specially modified to perform DHC tests in the neutron beam. H redistribution during mechanical testing was followed in-situ by registering the changes in neutron transmission. The results were compared with the images obtained by light optical microscopy after the tests. The results highlight the capabilities of neutron imaging to analyze time-dependent H distribution during DHC crack growth.

CANDU6 single channel CCP analysis on pressure tube deformation effects with multi-physics coupled calculation

Though pressure tube deformation rigorously and simultaneously affects channel coolability and neutronics behavior, the two effects have never been coupled with each other. In this study, the CUPID code is coupled with the SHAFE code to obtain the Critical Channel Power (CCP) of a designated channel which has high nominal power, a high void fraction, a high flow rate, and high pressure deformation. Even though appendage modeling is not included in this study, adequate cell-wise uniform artificial form loss was sought through iteration and imposed to conserve the header to header pressure drop by considering the inlet & outlet feeder and end fitting pressure drop. Nested loops are included in CCP calculation logic; they are the channel power loop, pin power distribution update loop and flow rate optimization loop from outer loop to inner loop, in sequence. Both CUPID and SHAFE are based on the Fortran computer language; the CUPID code is converted to the Dynamic Link Library (DLL) and the SHAFE code can load the CUPID library whenever it is required. Finally, we found that a CCP decrement for a channel with large pressure tube deformation is not that much, even for a complex deformation case. In addition, the sagging effect for a CCP decrement is significantly different from the expectation before an analysis is completed.

Improved Pulsed-Jet Cleaning of Cone Filter Cartridges Using an Annular-Slit Nozzle

There is a problem with the insufficient cleaning of the dust filter cartridge by the pulsed-jet cleaning method. This paper examines the improvement in the pulse-jet performance of cone filter cartridges achieved by using an annular-slit nozzle. Off-line pulse jet experiments were conducted to study the influence of initial compressed air pressure, jet distance, and Venturi tube on injection pressure. This paper compared the filtration performance of conventional nozzles versus slit nozzles by conducting an online dust filtration-cleaning experiment and investigating the effect of a Venturi tube. The results demonstrated that the jet produced by the slit nozzle had a bigger surface area in contact with the surrounding air, enhancing the entrainment effect, boosting the overall pulsed-jet pressure, and significantly elevating the pressure in the upper and middle section of the cartridge, particularly in the upper section. In both common nozzle and slit nozzle cases, the pulsed-jet intensity initially increased and then decreased with the jet distance. The corresponding optimal jet distances were 500 mm and 400 mm, respectively. The slit nozzle provoked an increment of 16% and 44% in the pulsed-jet intensity and uniformity (to 619 Pa and 0.19), respectively. And there was an 8% reduction in residual pressure drop, a 6% improvement in fallen dust mass, and a 7% improvement in comprehensive filtration performance. The fallen dust mass produced by a single cleaning of the slit nozzle was 0.64 kg and the comprehensive filtration performance was 0.18. While the addition of a Venturi tube improved the cleaning of the filter cartridge, it also led to a higher filtration pressure drop in the dust collector and reduced comprehensive filtration performance.

Susceptibility of Zr-2.5 (wt. %) Nb alloy to undergo nodular corrosion in water and steam environments: Effect of surface, cold work, temperature

Nodular corrosion can severely affect the service life of zirconium alloys in water cooled nuclear reactors. In this study, the susceptibility of pressure tube alloy to undergo nodular corrosion in water-steam with air surroundings at 250°C and 300°C is investigated. Accelerated corrosion has been studied in single phase steam at 500°C. The effect of specimen surface condition and cold work on nodular corrosion susceptibility has been studied in detail. The results indicate that surface condition of the specimen plays primary role in determining the propensity of the alloy to undergo nodular corrosion. Cold work has negligible effect on the initiation as well as growth of oxide nodules in Zr-2.5 (wt. %)Nb alloy. The effect of temperature is highly localized and is visible on the size of the nodules. Presence of non-uniform or discontinuous autoclave layer increases the susceptibility of pressure tube alloy to undergo nodular corrosion.

Numerical Investigation into the Temperature Profile of a PHWR Channel Containing a Disassembled Fuel Bundle During a Postulated Accident Condition

A reactor core overheats due to decay heat generated in the fuel when an effective cooling medium is unavailable, such as in a loss-of-coolant accident combined with a loss of emergency core coolant. If the heat generated is not effectively dissipated, then at extreme temperatures, the structural strength of the bundle assembly may deteriorate, leading to slumping of fuel elements onto the inner wall of the pressure tube. It is essential to examine the temperature behavior of the channel containing fuel pins in a disassembled state in order to comprehend the impact of further thermally induced deformations in the channel during postulated accident conditions. Capturing the temperature of channel components at each circumferential position from experiments is extremely difficult; thus, a modeling tool is necessary to obtain a thorough circumferential temperature profile. This paper presents a numerical study that aims to study the temperature distributions in a 1-m-long pressurized heavy water reactor (PHWR) channel containing a disassembled fuel bundle. The channel geometry and the boundary conditions implemented were obtained from the experiment. A temperature profile for each channel element at every circumferential and axial location was obtained. A thorough comparison of the predicted and the reported experimental values was performed, and it was found that the predicted temperature behavior of the channel was consistent with the experimental data. Further simulations with different fuel element configurations and decay powers may be carried out; in addition, the results obtained may be used for coupled thermal-mechanical and thermal-mechanical-chemical simulations.

Transformation mechanism of C and N during rapid devolatilization of pulverized coal particles under the pressurized atmospheres

Pressurized O2/CO2 combustion technology is a new power generation technology that can achieve CO2 capture efficiently. In this study, we investigated the pressurized devolatilization of Shenhua bituminous coal in a pressurized drop tube furnace experimental system at 1273 K. The evolution mechanism of carbon/nitrogen gas–solid products during the pressurized devolatilization process was analyzed using online measurements and offline analysis. We found that increasing the pressure from 0.3 to 1.2 MPa promoted the emission of CO and inhibited emission of NO. N2O emissions were insensitive to the change of pressure, and NO2 emissions were almost zero under the experimental conditions. The introduction of CO2 decreased the production of NOx and primary sources of nitrogen-containing gas-phase products were N-5 and N-6 active groups. Therefore, pressurized O2/CO2 combustion technology can effectively reduce NOx emissions in the devolatilization of pulverized coal and impact the subsequent char-combustion process.

Green hydrogen for ammonia production – A case for the Netherlands

An integrated system is studied to supply green hydrogen feedstock for ammonia production in the Netherlands. The system is modeled to compare wind and solar resources when coupled to Alkaline Electrolysis (AEL) and Proton Exchange Membrane Electrolysis (PEMEL) technologies with a compressed hydrogen storage system. The nominal installed capacity of the electrolysis plant is around 2.3 GW, with the most suitable energy source offshore wind, and the preferred storage technology pressurized tubes. For Alkaline Electrolysis and Proton Exchange Membrane Electrolysis technologies, the levelized cost of hydrogen is 5.30 €/kg H2 and 6.03 €/kg H2, respectively.

A Composite Analogy to Study the Linear Elasticity of a Pressurized Latex Tube with Application to a Mechanical Vocal Fold Replica

Mechanical deformable vocal fold replicas are an inherent part of physical studies of the fluid–structure interaction underlying vocal folds auto-oscillation during voiced speech sound production. In this context, the current work considers the linear stress–strain characterization of a pressurized latex tube vocal fold replica. An imaging approach is developed to measure the effective low-strain linear Young’s moduli along the streamwise (49[Formula: see text]kPa) and transverse main auto-oscillation (44[Formula: see text]kPa) directions. Next, a composite analogy is proposed to model the replica’s structure as an equivalent de-homogenized multi-layer material with two, three or four layers. This way equivalent low-strain Young’s moduli of each equivalent single layer can be estimated. Both measured effective and modeled equivalent low-strain Young’s moduli are within the range up to 65 kPa associated with human vocal folds. Resulting equivalent composite representations are of interest for the future design of pressurized latex tube replicas. This is illustrated considering the influence of outer layer latex properties on the overall estimate of the effective Young’s modulus. The proposed analogy is thus efficient in contributing to the direct comparison, in terms of low-strain elastic behavior, between replicas.

Challenges to the integrity of calandria, the CANDU PHWR moderator vessel, under serious accident loads and consequences of its failure

This paper summarizes the technical and analytical basis for a more advanced examination of structural integrity of a CANDU PHWR Calandria vessel under certain reactor accident conditions. Of foremost interest is the early Boiling Liquid Expanding Vapour Explosions (BLEVEs) induced challenge to vessel structural integrity following a spontaneous in-core rupture of a fuel channel. A sudden, flashing jet discharge of over 100 kg/s of the 10 MPa, 300 °C saturated primary coolant into its 70 °C subcooled near atmospheric pressure moderator water inventory will cause an explosive, shock wave impact. This results from rapid nucleation and vaporization of the discharged superheated liquid. Consequential pressure pulse, pipe-whip and bundle debris impact will induce damage to adjacent channels and cause widespread failure of fuel channel calandria tube rolled joints. These thin tubes inevitably deform to create serious challenges to vessel integrity. On the other extreme of reactor accidents, is the matter of an in-vessel-retention (IVR) of high temperature molten fuel debris in the reactor vessel after a severe core damage accident, such as one caused by an unmitigated station blackout. This scenario has understandably drawn greater attention after Fukushima and potential IVR in the > 200 mm thick PWR vessels when partly submerged in water is viable.It is highly unlikely that we will see reactor debris melt retention in this 19 to 29 mm thick, stepped shell with a tapestry of welds. This is because partial shell melting, diminished strength beyond 650 °C across 2/3 of the thin shell thickness and severe thermal stresses adds to underlying residual stresses near the welded seams. The result is inevitable catastrophic off-site consequences as energetic interactions upon vessel breach of water with hot debris can cause additional steam & hydrogen explosions, containment failures, and large off-site releases.This paper presents technical arguments to illustrate the problem in accepting faulty assumptions conveniently made in licensing safety assessments to simplistically claim survival of the thin-walled welded structure under these two illustrative accidents. I propose pathways to further investigations. Therefore, the issue of Calandria vessel survival under a wide range of accident conditions must be given a second, thoughtful look. Extending boundaries of a safe operating envelope as pressure tubes age needs to be avoided. We need to be more realistic in our severe accident analysis and consequence assessments. Fukushima taught us that the consequences of continuing to rely on convenient but technically faulty rationale, along with hype of eternal safety, can be very unforgiving.

Pd/oCNT Monolithic Catalysts for Continuous-Flow Selective Hydrogenation of Cinnamaldehyde

Chemical engineering optimization from a batch process to continuous flow in liquid-phase hydrogenation brings a significant improvement in efficiency. However, its further application is limited due to the severe pressure drop and tube blockage problems in a powder-form catalyst fixed-bed reactor, especially for nanocarbon-supported catalysts. In this work, spherical monoliths containing oxygenated carbon nanotube (oCNT)-supported Pd nanoparticles (ca. average size of 2 mm) are fabricated via an in situ gelation method and applied for cinnamaldehyde (CAL) selective hydrogenation in a continuous-flow system. The simulated results by the computational fluid dynamics–discrete element method (CFD–DEM) coupled method show that the pressure drop of the monolith catalyst bed is maintained within 0.4 Pa. The Pd/oCNT monolithic catalyst exhibits excellent CAL conversion of 85.8% and hydrocinnamaldehyde (HCAL) selectivity of 93.5% within a high weight hourly space velocity (WHSV) of 0.012 s–1 at mild reaction conditions (30 °C, 3 bar). The catalyst maintains robust catalytic activity and HCAL selectivity (>93%) under varied reaction temperatures (30/60 °C), H2 partial pressures (3–10 bar), and WHSVs (0.012–0.184 s–1) and a stable reaction activity for more than 60 h time on stream, revealing the possibility in industrial hydrogenation reactions. The catalytic activity of the monolithic catalyst is determined by the surface properties of the carbon nanotubes and the chemical interactions between Pd nanoparticles and supports. The oxygenated functional groups and surface defects on oCNTs are beneficial for strong chemical interaction with Pd species, which forms abundant electron-deficient Pdδ+ species to facilitate C═C hydrogenation. This study puts forward insights into and perspectives for selective hydrogenation reactions in both electronic structure tuning of the active phase at the atomic scale and fabrication of monolithic catalysts at the macroscopic scale.

Sorption study of long-lived 94Nb on laterite: Effects of physicochemical parameters on sorption

Zr–Nb alloy is used as the pressure tube in pressurized heavy water reactors (PHWR). Prolonged neutron irradiation of the pressure tubes leads to the formation of a long-lived radioisotope 94Nb. Thus, the discharged pressure tubes possess huge 94Nb activity which persists for a prolonged period.If these discharged pressure tubes come in contact with ground water, 94Nb isotope may leach and migrate and this can lead to a long-term radiological impact in the environment.In the present study, we have explored the capability of laterite as a filler material for the containment and retarding the migration of 94Nb. In this regard, detailed characterization of the laterite soil was carried out using energy dispersive X-ray fluorescence (EDXRF), X-ray diffraction (XRD), fourier transform infra-red spectrometry (FTIR), total cation exchange capacity determination, zeta potential measurement and thermogravimetric analysis (TGA). The sorption study of 94Nb on laterite was carried and the effects of different physico-chemical parameters like pH, ionic strength, temperature and equilibration time were evaluated. Ionic strength, temperature and time dependent sorption studies assist to explore the probable sorption mechanism of 94Nb on laterite, which helps in understanding the migration behaviour of 94Nb in natural aquatic environment. This study suggests that laterite is a promising material in containment of 94Nb isotope owing to its good cation exchange behaviour in the acidic medium and ability to form surface complex in the neutral medium.

Machine Learning-Augmented Micro-Defect Detection on Plastic Straw

Plastic straws are well-known tools to assist human beings in drinking fluid, but most of them have micro-defects including black spot defects, head problems, pressure tube defects, and sealing wrinkles. The manual detection of these defects has drawbacks such as low efficiency, a high false detection rate, and excessive labor. This paper proposed machine vision-based detection with self-adaption and high-accuracy characteristics. A serial synthesis of algorithms including homomorphic filtering, Nobuyuki Otsu, and morphological opening operations is proposed to obtain plastic straws with binary images with good performance, and it was further found that the convolutional neural network can be designed to realize the real-time recognition of black spot defects, where the corner detection algorithm demonstrates the linear fitting of the edge point of the straw with the effective detection of sealing wrinkle defects. We also demonstrated that the multi-threshold classification algorithm is used to detect defects effectively for head problems and pressure tube defects. The detection system based on machine vision successfully overcomes shortcomings of manual inspection, which has high inspection efficiency and adaptively detects multiple defects with 96.85% accuracy. This research can effectively help straw companies achieve high-quality automated production and promotes the application of machine vision in plastic straw defects with the aid of machine learning.

Auto Sizing of CANDU Nuclear Reactor Fuel Channel Flaws from UT Scans.

The inspection of nuclear power plants is an essential process that occurs during plant outages. During this process, various systems are inspected, including the reactor's fuel channels to ensure that they are safe and reliable for the plant's operation. The inspection of Canada Deuterium Uranium (CANDU®) reactor pressure tubes, which are the core component of the fuel channels and house the reactor fuel bundles, is performed using Ultrasonic Testing (UT). Based on the current process that is followed by Canadian nuclear operators, the UT scans are manually examined by analysts to locate, measure, and characterize pressure tube flaws. This paper proposes solutions for the auto-detection and sizing of pressure tube flaws using two deterministic algorithms, the first uses segmented linear regression, while the second uses the average time of flight (ToF) within ±σ of µ. When compared against a manual analysis stream, the linear regression algorithm and the average ToF achieved an average depth difference of 0.0180 mm and 0.0206 mm, respectively. These results are very close to the depth difference of 0.0156 mm when comparing two manual streams. Therefore, the proposed algorithms can be adopted in production, which can lead to significant cost savings in terms of time and labor.

Pressure tube replica imaging system for PHWR reactors based on optical coherence tomography

This paper describes a 3D profiling system for surface profiling of replicas obtained from pressure tube (PT) of PHWR/CANDU type reactors. This system is based on Optical Coherence Tomography (OCT) technique. Fiber based Fourier domain OCT configuration was used to design and develop the system considering the potential application of imaging radioactive PT replica samples. The sample scanning assembly of the system has been custom designed considering the cylindrical nature of PT replicas. Replica of inner surface of pressure tube was obtained on a replica plate using a custom built replica tool. Three PT replicas of non-radioactive pressure tubes with various flaws were obtained in different conditions. These replicas were imaged in 3D for identification and quantification of the flaws. This 3D profiling system termed as ‘PT replica imaging system’ is a promising tool which can be useful in periodic inspection of pressure tubes for safe operation of nuclear reactor.

A multi-segment cascade-connected transducer used as acoustic source of pressurized water-filled impedance tube

A multi-segment cascade-connected transducer used as acoustic source of pressurized water-filled impedance tube

Numerical Investigation on Heat Flux Variation in Fuel Bundle Over Fully Voided Concentric Channel of IPHWR Under Severe Accidental Scenario

Abstract A steady-state thermal analysis is performed on a fully voided channel of an Indian pressurized heavy water reactor (IPHWR) by providing variable heat flux inputs of 800 W/m2, 1500 W/m2, 2500 W/m2, 3500 W/m2, and 4500 W/m2 for evaluating the numerical results of temperature distributions in pressure tube (PT) and calandria tube (CT) at concentric scenarios using a numerical approach. The top to bottom temperature difference of PT and CT is found to be 0.57% and 3.31%, respectively, for 800 W/m2 input heat flux, and this variation decreases to 0.015% and 0.05%, respectively, for 4500 W/m2 input heat flux, thus pointing out that increasing the heat flux leads to the decreased top to bottom temperature difference. Also, the temperature distribution pattern is found to be similar for all the input heat fluxes, and for a particular heat flux, the change in circumferential temperature is found to be negligible. The boiling phenomenon starts at 2500 W/m2, and the volume fraction of moderator vapor increases as the heat flux is increased, the average density of moderators inside the enclosure decreases, and continuous variation in film thickness is also observed with respect to time for a particular heat flux value. The results also showed that the moderator acted as an effective heat sink for higher heat input rates.

Anisotropy study of hydrogen diffusion along different directions of Zr-2.5%Nb alloy pressure tube using neutron imaging

Hydrogen diffusion studies in Zr-2.5% Nb alloy pressure tube material are important inputs for modeling localized embrittlement phenomenon manifested in pressure tubes. The microstructure of the pressure tube consists of lamellar α-Zr along with the β-phase present as stingers between two alpha laths as well as fine and coarse beta globules. This two-phase microstructure along with a vast difference in diffusivities for the alpha and beta phase necessitates the need for evaluating the diffusion parameters along the three orthogonal directions of the pressure tube. This work focuses on studying the diffusivity values for hydrogen in Zr-2.5%Nb alloy along different pressure tube directions in the temperature range is 250 ˚C -450 ˚C and also correlates it to the microstructure. FEM studies have also been carried out to identify the concentration region that is unaffected by stress imposed due to the presence of a hydride layer. Neutron radiography has been used to study diffusion along the axial, circumferential and radial directions of an as-received un-irradiated pressure tube. It was observed that the diffusion along the axial and circumferential directions is similar while it is slightly lower in the case of radial direction. A theoretical approach has also been followed for evaluating the diffusivity values for the α-Zr and β-Zr phases for the same temperature range.

Mechanisms of Hydride Nucleation, Growth, Reorientation, and Embrittlement in Zirconium: A Review.

Zirconium (Zr) hydrides threaten the reliability of fuel assembly and have repeatedly induced failures in cladding tubes and pressure vessels. Thus, they attract a broad range of research interests. For example, delayed hydride cracking induced a severe fracture and failure in a Zircaloy-2 pressure tube in 1983, causing the emergency shutdown of the Pickering nuclear reactor. Hydride has high hardness and very low toughness, and it tends to aggregate toward cooler or tensile regions, which initiates localized hydride precipitation and results in delayed hydride cracking. Notably, hydride reorientation under tensile stress substantially decreases the fracture toughness and increases the ductile-to-brittle transition temperature of Zr alloys, which reduces the safety of the long-term storage of spent nuclear fuel. Therefore, improving our knowledge of Zr hydrides is useful for effectively controlling hydride embrittlement in fuel assembly. The aim of this review is to reorganize the mechanisms of hydride nucleation and growth behaviors, hydride reorientation under external stress, and hydride-induced embrittlement. We revisit important examples of progress of research in this field and emphasize the key future aspects of research on Zr hydrides.