Nuclear Containment Research Articles

Investigation of Acoustic Emission Wave Propagation Pattern with Defect Locations in a Multilayer Structure

Acoustic emission (AE) is used to monitor conditions of various structures across many industrial sectors, including containment vessel or storage tank of nuclear materials. Periodic monitoring, inspection, and analysis of structure conditions can help prevent failure and accidents. Understanding the transient elastic waves in multi-layered structures (planar or rounded types) has long been of great interest. This paper experimentally investigates changes in AE wave propagation patterns in multilayer planar structures (detecting and assessing the effect of coating layers, assumed surrogate of deposits or protective layer). Epoxy phenolic coated two mild steel plates were assembled (without any adhesion), and two piezoelectric AE sensors were placed on the coating layer. The pencil lead break (PLB) test was used to initiate the AE waves from the surface and cross-section of different layers. From wavelet transforms (WT) analysis, significant energy zone changes were observed up to the 450 kHz frequency level with PLB on the surface and cross-section of different layers. Love wave propagation on the coated plate structure resulted in wave pattern changes with PLB locations and layers. Wave duration, energy, energy ratio, and peak amplitude levels were also analysed to characterise the AE wave pattern relationship with defect location in a multilayer plate-like (planar) structure. The approaches used in this work could potentially be useful in providing a greater understanding of defects within multilayer nuclear containment structures, and also offering an alternative way to monitor corrosion related degradation of structures with insulations.

Numerical simulation and experimental study on mechanical behavior of the gusset for nuclear power plant containment

The gusset of a nuclear power plant containment, which serves as the junction between the shell and the raft foundation, is a critical area that poses a significant challenge due to its irregular shape and complex mechanical properties. Studying the stress mechanism of this area is crucial to understanding the overall structural performance of the containment. To achieve this, ABAQUS software was used to simplify the containment model and develop a gusset model, which was validated by comparing it with a fan-shaped containment model. Two 1:3 scale specimens were then designed and fabricated based on the gusset model to conduct static tests under two loading combinations. The effectiveness of the finite element model (FEM) was verified by comparing the failure mode, load–strain relationship of the rebars, and load–displacement relationship with that of the test. Through finite element parameterization analysis, the effects of reinforcement ratio and thickness on the mechanical performance of the gusset were investigated. The results showed that tripling the reinforcement ratio in rows H1 to H3 and column S1 or increasing the specimen thickness by 500 mm significantly improved the tensile damage of concrete, yield of rebars and structural displacement. Additionally, removing rebars from columns S8 to S13 had little effect on the structure. The findings of this study can guide the optimization of containment structural reinforcement and provide essential data support for FEM.

The performance enhancement of containment with fiber reinforcement following a LOCA considering different ambient conditions

The prestressed concrete containment vessels (PCCVs) will be exposed to the temperature action followed by the internal pressure increasing after a loss of coolant accident (LOCA) with safety system failure. The damage evolution of the PCCV under the effect of thermal-pressure coupling deserves in-depth study. Moreover, different containment external temperatures are seldom considered in previous studies whereas they may further pose a great challenge for the design and safety of containments. Just as important, an attempt is first made in this paper to involve steel fiber in the construction of the containment to observe the performance of the containment with steel fiber strengthening under the effect of thermal-pressure coupling. Therefore, the influence of the possible outside temperature (i.e. −30 °C, −15 °C, 0 °C, 15 °C, and 30 °C) determined from the 5 existing or planned nuclear power plant sites in China on the distribution of temperature field of containment is researched firstly. Then, the damage indices corresponding to the containment are defined to represent the actual damage performance and failure mechanism of the structure. Finally, the comparative damage performance and failure mechanism between conventional and steel fiber reinforced concrete (SFRC) containments taking into account the effect of coupling temperature and internal pressure is detailedly discussed. The results show that the SFRC is more expeditious in restraining radial and vertical displacement of containment under lower external temperatures for subsequent internal pressure. Under the ultimate internal pressure, the SFRC containment has a reduction rate of 5.37% for radial displacement and 35.52% for vertical displacement at an external temperature of −30 °C, while 2.23% and 15.33% at an external temperature of 30 °C, respectively. Furthermore, SFRC containment has better performance under the effect of coupling of temperature and internal pressure compared to conventional containment. For example, at an internal pressure of 1.2 MPa, the damage reduction rate for concrete is 94.18% at an external temperature of −30 °C and 81.37% at 30 °C; for the steel liner, the damage reduction rate is 50.22% at an external temperature of −30 °C and 87.83% at 30 °C. For the containment prestressing tendons, the damage reduction rate is around 60% under ultimate internal pressure. The study can provide important guidance for siting and design of nuclear containments and alternative retrofitting measures for nuclear containments.

Seismic fragility analysis of nuclear containment structure under chloride-induced reinforcing steel corrosion conditions

Chloride-induced corrosion of reinforcing steel is one of the critical degradation factors influence the life cycle performance of the containment structure. Therefore, it is of great significance to study the safety performance of containment structure under chloride-induced reinforcing steel corrosion condition. This study presents fragility evaluation of the containment structure under chloride-induced reinforcing steel corrosion condition. Nonlinear time history analysis of the containment structure under different reinforcing steel corrosion conditions is carried out considering various uncertainties including uncertainty in material strength, record to record uncertainty and uncertainty in corrosion parameters. Roof drift under different ground motion intensity levels and seismic fragility of the containment structure corresponding to different performance level with typical service life are analyzed. Results reveal that with the increase of the service life of the containment structure, the increasing rate for roof drift of the containment structure decreases, while the variability of roof drift of the containment structure changes slightly. Service life of the containment structure has different impacts on the fragility of the containment structure corresponding to different performance levels. With the increase of service life, the increasing rate for failure probability of containment structure increases.

The influence of hydrogen permeation on the protection performance of the Cu coating of nuclear waste containers

Hydrogen permeation and corrosion behaviour of Cu-coated carbon steel under simulated nuclear waste disposal conditions of the Beishan area have been investigated. The permeation current (with −500 μA·cm−2 charging current) decreased from 3778 to 141 and 47 nA.cm−2 respectively when 5 and 10 µm thick Cu coatings were electrodeposited on the steel. The XRD analysis suggests the formation of CuH during the hydrogen charging treatment, and its subsequent decomposition formed a protective CuOH/Cu2O layer on the charging surface when exposed to simulated groundwater for 44 h. No damage to the Cu coating-steel interface was observed in the cross-section analysis.

Assessing Toughness Levels for Nuclear Containment Vessels Steel to Determine the Need for Postweld Heat Treatment Based on the Master Curve Method

AbstractIn this paper, a minimum toughness assessment method of exempting postweld heat treatment (PWHT) based on the master curve method is proposed to replace the current method of determining whether to perform PWHT by material thickness. Reference temperature to prevent fracture (T0-required) for structure to be exempt from PWHT at minimum design metal temperature (MDMT) was obtained by using the master curve method under presumed stresses and flaw size. A series of exemption curves were generated for the steels used in pressure vessels with yield strength of 300∼550 MPa, and the method was corrected for low-stress conditions. Then the feasibility of this method was verified by an engineering example. A SA-738Gr.B welded joint in the example was taken as the research object, the experiment showed that the weld metal was the worst part of the fracture toughness of the joint. The results of research objects which were exempted from PWHT by the proposed method are consistent with ASME Code Case N-841, which proves that this method is feasible in assessing the exemption from PWHT.

A universal correlation development for film-wise condensation in the presence of non-condensable gases on vertical walls under turbulent free convection

Accurate prediction of steam condensation in the presence of non-condensable gases is essential for accident analysis in nuclear reactor containments. This study uses the steam–air mixture to illustrate the effect of the mixture's compressibility in analyzing steam condensation. Then, a consolidated database of steam condensation on vertical walls was established, and seven typical empirical and semi-empirical correlations were evaluated. Based on heat and mass transfer analogy and curve fitting, a new universal correlation for predicting steam condensation on vertical walls in the presence of non-condensable gases, including binary and ternary mixtures, under turbulent free convection is established, and this correlation can well handle the transition from binary to ternary mixtures using the same formula. More than 90% of predicted results compared with 1027 data points are within the ± 25% error band, and the mean absolute error is 10.6%. The similarity of helium and hydrogen in analyzing steam condensation is discussed.

Probabilistic seismic performance assessment of nuclear containment structure considering statistical uncertainty

Statistical uncertainty is inherently involved in probabilistic seismic performance assessment of containment structure due to heavy computational efforts of nonlinear dynamic analysis. Aiming at the statistical uncertainty involved in seismic fragility and seismic risk evaluation of containment structure, this study proposes a method for quantifying the statistical uncertainty in probabilistic seismic performance assessment of the containment structure. Taking a three story reinforced concrete frame as an example, effectiveness and efficiency of the proposed method for confidence interval estimation of seismic fragility curve is verified. Then, the proposed method is applied to fragility analysis and seismic risk evaluation of the containment structure. Roof drift of the containment structure under different ground motion intensity levels is discussed. Confidence interval for seismic fragility curve and seismic risk of the containment structure considering statistical uncertainty is quantified by the proposed method. The results highlighted the necessity for quantifying the impact of statistical uncertainty involved in probabilistic seismic performance evaluation of the containment structure.

Comparative analysis on application conditions of indium (III) oxide-reinforced glasses in nuclear waste management and source transportation: A Monte Carlo simulation study

This study's primary objective is to provide the preliminary findings of novel research on the design of Indium (III) oxide-reinforced glass container that were thoroughly developed for the purpose of a nuclear material container for transportation and waste management applications. The shielding characteristics of an Indium (III) oxide-reinforced glass container with a certain elemental composition against the 60Co radioisotope was thoroughly evaluated. The energy deposition in the air surrounding the designed portable glass containers is measured using MCNPX general-purpose Monte Carlo code. Simulation studies were carried out using Lenovo-P620 workstation and the number of tracks was defined as 108 in each simulation phase. According to results, the indium oxide-doped C6 (TZI8) container exhibits superior protective properties compared to other conventional container materials such as 0.5Bitumen-0.5 Cement, Pb Glass composite, Steel-Magnetite concrete. In addition to its superiority in terms of nuclear safety, it is proposed that the source's simultaneous observation and monitoring, as well as the C6 (TZI8) glass structure's transparency, be underlined as significant advantages. High-density glasses, which may replace undesirable materials such as concrete and lead, provide several advantages in terms of production ease, non-toxic properties, and resource monitoring. In conclusion, the use of Indium (III) oxide-reinforced glass with its high transparency and outstanding protection properties may be a substantial choice in places where concrete is required to ensure the safety of nuclear materials.

Ultimate strength assessment of penetrated nuclear reactor containment vessel subjected to external pressure and thermal loads

For nuclear-power ships or submarines, the mechanical penetration assemblies (MPAs) installed on the nuclear reactor containment are designed to withstand extreme loadings caused by high pressures and temperatures. The ultimate strength of the MPAs should be adequately predicted for the safety of reactor containment. An analysis approach has been developed to assess the ultimate strength of MPAs under external pressure and thermal load. Based on the equivalent method, the analysis method is validated by the theoretical formula and experimental data of the cylindrical shells. Then, the influences of geometry, environmental temperature and thermal stress on the ultimate strength of the MPAs are discussed. An ultimate strength assessment relationship of the MPA subjected to external pressure and thermal loads is also proposed based on the analysis results. The proposed approach and suggested relationship can drastically facilitate future MPA design from the point of view of ultimate strength.

Standardized models for cracking due to restraint of imposed strains—The state of the art

AbstractCrack control throughout the service life of a structure has always been a challenging task for engineers and asset owners, particularly those who are involved with the design and construction of structures with increased durability requirements, such as liquid retaining structures, nuclear containment buildings or structures in direct contact with ground, but also other externally restrained structures. For these reasons, several standardized methods exist throughout the global engineering community which encompass differences between them. This work is a state‐of‐the‐art review of the methods for crack width control due to imposed strains recommended by different regulatory documents from Europe, USA, Australia and Japan, and provide a reasonable degree of guidance with respect to the assumptions, possibilities and restrictions of each method. This work resonates with the scope of the RILEM TC 287‐CCS: Early age and long‐term crack width analysis in RC structures.

Pressure-induced damage monitoring in prestressed concrete of nuclear containment wall segments using acoustic emission technique – Application to VeRCoRs containment building

This paper reports benchmark findings of an experimental study of pressure-induced damage monitoring in a nuclear containment building (NCB) reproducing realistic conditions of internal pressure increase accidents. The leak-tightness of NCBs is assessed every 10 years during integrated leak-rate tests (IRLT). For double-wall containments, which have no metallic liners, the leak-tightness is strongly influenced by the degree of cracking of concrete and opening of the cracks. This paper is focused on assessing the effectiveness of the Acoustic Emission (AE) technique to evaluate the damage evolution of a reinforced concrete containment wall under pressure loading test. Piezoelectric accelerometers were mounted on the internal containment wall of the VeRCoRs mockup, which constitutes the containment building of a Nuclear Power Plant in a 1:3 scale. The Akaike criterion (AIC) as an AE hit picking method and the adaptive meshing algorithm (AMA) as an AE events localization algorithm have been applied for damage and cracking monitoring. A benchmark analysis based on the distribution of AE events and energy is presented for the detection of damages and the monitoring of crack propagation in correlation with the known locations of main cracks. The potential and limitations of the AE technique for in situ damage sources estimation are discussed upon two experimental campaigns (2017 and 2019), and the main conclusions of the benchmark survey are given. The results show that the AE technique is adequate for damage evaluation in large-scale concrete structures. It provides important information about the integrity of the structure and the assessment of damage evolution with time.

Prediction of long-term prestress loss for prestressed concrete cylinder structures using machine learning

The long-term prestress loss caused by shrinkage and creep of concrete and stress relaxation of prestressed tendons has significant effects on the sealability and safety of prestressed concrete cylinder structures such as nuclear reactor containments and liquified natural gas (LNG) tanks. By utilizing machine learning (ML) techniques, this study aims to establish an intelligent approach for the long-term prestress loss prediction of concrete cylinder structures. Firstly, based on the Infrastructure Technology Institute of Northwestern University (NU-ITI) database of concrete shrinkage and creep performance, the explicit expressions are presented for concrete shrinkage and creep function using genetic programming (GP); Moreover, the concrete constitutive model is incorporated into a general finite-element software package based on the ABAQUS UMAT platform. Then finite element analysis (FEA) models are established and calibrated based on the existing long-term prestress loss tests of prestressed concrete beams. In addition to the experimental results in the literature, the numerical results of the FEA model are also used to form the database of the long-term prestress losses for concrete cylinder structures. Finally, three prediction models of long-term prestress loss are proposed by utilizing the artificial neural network (ANN), one-dimensional convolutional neural network (1D CNN) and genetic programming (GP). Compared with the measured results of nuclear containments in practical engineering, the ML based prediction models are demonstrated to be accurate and efficient in evaluating the long-term prestress loss for prestressed concrete cylinder structures.

Large-eddy simulation on gas mixing induced by the high-buoyancy flow in the CIGMAfacility

The hydrogen behavior in a nuclear containment vessel is a significant issue when discussing the potential of hydrogen combustion during a severe accident. After the Fukushima-Daiichi accident in Japan, we have investigated in-depth the hydrogen transport mechanisms by utilizing experimental and numerical approaches. Computational fluid dynamics is a powerful tool for better understanding the transport behavior of gas mixtures, including hydrogen. This paper describes a Large-eddy simulation of gas mixing driven by a high-buoyancy flow. We focused on the interaction behavior of heat and mass transfers driven by the horizontal high-buoyant flow during density stratification. For validation, the experimental data of the Containment InteGral effects Measurement Apparatus (CIGMA) facility were used. With a high-power heater for the gas-injection line in the CIGMA facility, a high-temperature flow of approximately 390 °C was injected into the test vessel. By using the CIGMA facility, we can extend the experimental data to the high-temperature region. The phenomenological discussion in this paper helps understand the heat and mass transfer induced by the high-buoyancy flow in the containment vessel during a severe accident.

Development of a specialised calibration and infra-red emissivity correction to enable the temperature monitoring of intermediate level nuclear waste using thermal imaging

The temperature measurement of intermediate level nuclear waste containers is vital to monitor thermal stability and track temperature trends over time. The use of thermal imaging for this monitoring has numerous advantages compared to using a single spot radiation thermometer, and Sellafield wish to employ this approach for their ‘3 m3 box’ intermediate level waste containers. In order to achieve this objective with confidence a custom temperature calibration is required for low uncertainty temperature measurement. Another key consideration for practical non-contact temperature measurement is the emissivity of the surface being measured at the relevant infrared wavelengths, as such the emissivity of the Sellafield 3 m3 box materials was measured and then corrected for. The implementation of a calibrated instrument and emissivity correction were validated using temperature controlled, 1:10 scale model 3 m3 box containers within a climatic chamber. The surface temperature of the scale containers was determined to an uncertainty of ± 6.5 °C (k = 2.1) for the container walls, ± 3.5 °C (k = 2.1) for the container lid and ± 1.5 °C (k = 2.1) for the container vents. The measurement techniques used to determine the temperature of the 1:10 scale 3 m3 box containers were therefore successful and there is a clear development path to monitoring the containers in storage using thermal imaging.

Development of new AE technology to detect and characterize leak on nuclear shielding building from inside during integrated leakage rate test

French Projects of Investment for the Future, called “Research for Nuclear Safety and Radiation Protection” have been initiated to further research on the causes, the management, the impact of the observed nuclear accidents and to propose and validate solutions to limit the risk and the consequences. In this context the "Non Destructive Evaluation of nuclear plants containment” project” (ENDE) with eight partners (six research institutes and two industrials) supported by the “National Agency of Research”, proposes a methodology for the Non Destructive Evaluation of the containment capacity. The two major functions to be assess are strength and leak tightness. Under the frame of the VeRCoRs project EDF has developed and build a containment mock-up on the 1/3 scale (diameter 20m, height 20m, wall thickness 40cm) which has been used for some partners of ENDE project. Mistras has been involved in characterisation of concrete cracking during pneumatic proof tests either in ENDE and VeRCoRs programs. Since some leaks occurs in the area of monitoring we use AE to characterized them. The mockup has been used to develop and characterized an instrumentation and technology for leak detection with a major challenge leak detection has to be performed from this inside of the containment building with all the contingences of operating of full scale operating nuclear plant. This paper describe the steps of development and validation of the methodology up to the application this measurement on site help the leak characterisation during pneumatic ten year Integrated Leakage Rate Test at 4,2 bar. EWGAE 35, Ljubljana, Slovenia, 13th – 16th Sep. www.ewg

Bayesian inference with correction of model bias for Thermo-Hydro-Mechanical models of large concrete structures

Drying and creep are mainly driving the continuous strain evolution in time of aging large prestressed concrete structures, and may jeopardize the safety of their environment. Therefore, it is crucial to dispose of an accurate assessment of structures’ long-term strain level so as to evaluate their integrity. In the framework of probabilistic modeling of the delayed mechanical analysis of aging concrete structures, Bayesian inference allows to update uncertainties related to input parameters of computational models through the assimilation of noisy monitoring data. In this paper, a Bayesian inference methodology aiming at updating uncertain parameters of computational models for large concrete structures is proposed. This methodology combines a powerful surrogate modeling technique named PC-PCE (Principal Component Polynomial Chaos Expansions) with the so-called BUS (Bayesian Updating with Structural reliability methods) framework, in order to efficiently draw samples from posterior distributions for a sensibly reduced computational cost. In particular, the methodology is based on a framework to account for model uncertainties and biases, which are usually disregarded in the existing literature related to large concrete structures, and it enables correction predictions through Bayesian inference. The proposed approach is illustrated through an application to a large aging prestressed concrete structure, namely a 1:3 scale mock-up of a Nuclear Containment Building. In this context, a thermo-hydro-mechanical computational model with uncertain parameters is adopted to model the time evolution of strains of the structure. Results emphasize that the proposed approach performs Bayesian updating for a reduced computational cost. The proposed Bayesian inference approach also enables the identification of model biases, and correction of strain predictions.

Preliminary study on a new self-sealing strategy for nuclear reactor containment buildings using phase change materials

Nuclear containment buildings are designed as the final barriers to prevent the release of radioactive materials. A containment building consists of a containment liner plate (CLP) and a thick concrete wall. Concrete inevitably deteriorates due to a variety of factors, including the continuously applied load and plastic shrinkage. This continued deterioration could cause cracks in the concrete, which may extend through the entire concrete wall. This would ultimately compromise the ability of a containment building to serve as a final barrier between a nuclear power plant and its surroundings. As a passive measure to repair cracks, research on self-healing concrete has been conducted. In general, self-healing concrete is made by adding a healing agent or bacteria to a concrete mix. However, the previous self-healing concepts cannot be easily applied to current nuclear containment buildings since changes in the chemical composition of concrete mixes cause the changes in the mechanical properties of concrete; thus, a different containment design that matches the changed properties is required. Therefore, in this study, a new self-sealing strategy based on a phase change material (PCM) is introduced, in which the PCM seals a crack physically rather than chemically. To investigate the feasibility of this new strategy, numerical simulations with a thermal analysis were conducted. In addition, experiments were carried out to validate the results of the numerical simulations. The numerical simulations show that the PCMs can seal cracks with maximum sizes of 1.3 mm and 1.8 mm when the inlet temperatures are 60 °C and 55 °C, respectively. Finally, based on the results of successful PCM sealing, even with several conservative assumptions that make it difficult for the PCM to seal a crack, it is concluded that the new self-sealing strategy is indeed feasible.

Characterization of precipitates susceptible to clogging of sump filters after a Loss-of-Coolant-Accident in PWRs

During a Loss Of Coolant Accident involving a primary circuit failure in a nuclear power plant, debris, generated by the impact of the steam jet, is transported to the hot liquid phase formed at the bottom of the nuclear containment. Water is recirculated after passing through filters. Chemicals are released and may precipitate on the filters, increasing the pressure drop. This paper investigates the characterization of precipitate formation in a soda/borate buffer, due to the presence of calcium, silicon, and zinc, released from mineral fibers, concrete particles, or the corrosion of metal surfaces. Experiments run at various temperatures and chemical compositions are presented, with the objective of identifying the precipitates and determining the critical thresholds of concentration and temperature beyond which the risk of chemical clogging disappears. The experimental results are compared to those calculated with the CHESS software, intended to calculate the speciation at thermodynamic equilibrium. The results show the formation of amorphous precipitates: silica, calcium borates, borosilicates and zinc silicate. Precipitation is temperature dependent, with the highest masses obtained between 50 and 60 °C. The comparison of experimental and calculated results indicates a good overall consistency. This work demonstrates the importance of considering chemical clogging in this context.

Experimental investigation of shear-extension coupling effect in anisotropic reinforced concrete membrane elements

Performance based analysis under seismic loads using the finite element method for wall-type reinforced concrete (RC) members in buildings and in important structures like liquid retaining structures, nuclear containment structures, offshore concrete gravity structures etc., necessitates the understanding of the non-linear behaviour of the constituent membrane elements. The current orthotropic formulation of the softened membrane model (SMM) can be strictly used only when the reinforcement is symmetric to the principal axes of applied stresses. When the reinforcement is asymmetric, shear strain is generated due to the normal stresses in the principal axes of applied stresses, which is referred to as shear-extension coupling. An anisotropic formulation is required to capture the generated shear strain. The current study quantifies the shear strain due to asymmetry in reinforcement, by testing panels under biaxial tension-compression using a large-scale panel testing facility. A model for the shear strain is proposed based on the tests data. The paper presents the experimental programme, important test results and the modelling of shear strain. Expression developed for the shear strain can be incorporated in the solution algorithm of the SMM for improved prediction of the shear behaviour of a membrane element. This further aids in accurate prediction of the seismic performance of the important structures mentioned earlier.