Nuclear Power Plant Containment Research Articles

Analyses of Jet Buoyant Flow in a Multicompartment Containment Using an Open-Source Solver

Hydrogen mixing and distribution in a nuclear power plant containment during a postulated severe accident is a phenomenon with high safety relevance for current and advanced light water reactors. The Organisation for Economic Co-operation and Development (OECD)/Nuclear Energy Agency (NEA) SESAR Thermal-Hydraulics (SETH) project was carried out to generate experimental data on basic containment phenomena with the required instrumentation to assess computational fluid dynamics (CFD) and lumped parameter codes. In this paper, we analyze the PANDA SETH Test 16 of the OECD/NEA SETH project using containmentFOAM, which is a tailored open-source CFD code package for containment analysis developed at Forschungszentrum Jülich GmbH based on OpenFOAM.® SETH Test 16 addressed steam-air mixture transport driven by jets and plumes in the multicompartment PANDA facility (two vessels and an interconnecting pipe) at the Paul Scherrer Institut in Switzerland. Steam as a wall plume was injected at 0.040 kg/s and 140°C in vessel 1, which was initially filled with dry air at 108°C. The steam-air mixture was vented out from the top of vessel 2. The system pressure was constant at 1.3 bar for the whole test duration (4000 s). To simulate SETH Test 16, the k-ω shear stress transport turbulence model was adopted with the generalized gradient diffusive hypothesis buoyancy turbulence model. The simulation represented the experimental phenomenology reasonably well. The trends for the calculated temperature distribution were similar to those of the experiments and only yielded a slight discrepancy of about 2.7%. In addition, the distribution of the steam molar fraction was well predicted above the injection tube in vessel 1. However, below the injection tube in vessel 1, some discrepancies between the simulations and the experimental results were observed. These may have been caused by the higher turbulent mass diffusivity in the physical model and the challenge of the mesh resolution in the middle region of vessel 1.

Time-Limited Aging Analysis of the Containment of Nuclear Power Plants without Monitoring Tendons

A prestressed concrete containment is the enclosure structure of a nuclear reactor that serves as the last physical barrier of a nuclear power plant (NPP) safety defense system. It plays a key role in the prestress time-limited aging analysis (TLAA) required for operating license extensions for NPPs. Considering prestress containment systems without long-term monitoring tendons in a nuclear power plant, the technical route for prestress TLAAs involves analyzing operating license extension regulations and in-service monitoring technical requirements for prestressed tendons performance. Using tendons based on theoretical calculations of prestress loss in a nuclear power plant and the minimum required value of prestress determined by numerical simulation, the theoretical predictive value can be compared with the minimum required value for prestressed steel tendons. This comparison can be used to evaluate the 20-year life extension of the prestress system and provide a reference for aging and life management of NPPs.

Simulation and calibration of radiation monitoring of nuclear power plant containment sump waste liquid

Waste liquid stored in the containment sumps of nuclear power plants may contain a variety of radionuclides. Real-time monitoring of containment sump waste liquid can ensure that accidents, such as leakage of cooling water, can be avoided. This paper presents the design of a radioactive monitoring system for waste liquid in a containment sump. The detector and the lead-shield in the measurement unit are optimized through Monte Carlo simulations. Experimental verification showed that the background count rate of the measurement chamber in the system was 418.3 cps, and the detection limit of the detection system was 3.01 Bq/L. Distinct gamma-ray characteristic peaks were also observable, demonstrating the system's ability to identify radioactive nuclides in the waste.

Concrete ablation depth analysis of OPR1000 NPP during top flooding ex-vessel cooling

This study assesses the integrity of OPR1000 nuclear power plant (NPP) containment during top flooding ex-vessel cooling in the event of a postulated severe accident. Maintaining containment integrity during ex-vessel cooling is critical for preventing substantial release of radioactive materials to the environment. To analyze the feasibility of top flooding ex-vessel cooling of the OPR1000 NPP, the spread of molten corium in the dry reactor cavity is first calculated, followed by calculating the concrete ablation depth according to the reactor cavity flooding delay time. It is judged whether the leak-tightness integrity of the containment during ex-vessel cooling is maintained based on whether the containment liner plate (CLP) is damaged or not. In the first step analysis, the spread of molten corium in the dry reactor cavity is calculated using computational fluid dynamics (CFD) methodology for various cases, and results indicate a uniform deposition of approximately 143 tonnes of molten corium, with a thickness of about 33 cm, meaning that the thermal load is distributed uniformly in the reactor cavity. In the second step analysis, concrete ablation depths are calculated base on the delay time for flooding the reactor cavity. For the postulated severe accident, injecting coolant into the reactor cavity one hour after breach of the reactor vessel results in a concrete ablation depth of about 0.74 m, indicating no damage to the CLP and ensuring containment integrity. The maximum allowable delay time for flooding the reactor cavity for the postulated severe accident is evaluated to be 2.24 hours maintaining the leak-tightness integrity of the OPR1000 NPP containment. The research methodology and findings presented in this study are expected to aid in assessing the feasibility of top flooding ex-vessel cooling and establishing appropriate accident management strategies for nuclear power plants like the OPR1000 NPP.

CFD and machine learning based hybrid model for passive dilution of helium in a top ventilated compartment

Abstract For hydrogen management in the containment of Nuclear Power Plants (NPPs), besides the Passive autocatalytic Recombiners (PAR), the passive dilution of lighter gas plays an important role. This could be an attractive option to optimize the containment design and to estimate the extent of dilution. Passive dilution has many other applications in nuclear industry. The experimental studies of air entrainment in the upward rising helium plume and the resulting dilution of helium gas by the Canadians in terms of Volume Flow Magnification Factor (VFMF) have been utilized for Computational Fluid Dynamics (CFD) validation. The CFD based Fire Dynamics Simulator (FDS) predicted values of VFMF found to be in good agreement with the test data. After FDS code validation, parametric study has been carried out to generate a data base of VFMF for range of hydrogen injection, side opening area and opening height. In present study various Machine Learning (ML) models are evaluated based on two-parameter relationship i.e. non dimensional hydrogen injection and VFMF using the CFD code generated database. The trained ML models were used for the predictions of the mass flow rate of gas entrainment (through opening) in the rising buoyant plume in terms of VFMF. The ML predictions were in good agreement with the predictions against test data. Multivariate Adaptive Regression Splines (MARS) based ML model found to performed best and discussed in the paper. The paper highlights details of methodology of numerical simulation, results of the CFD studies and machine learning based predictions.

Optimal design of nuclear power plants containment with tuned mass damper inerter device for earthquake loadings

Optimal design of nuclear power plants containment with tuned mass damper inerter device for earthquake loadings

Thermal-hydraulic characteristics analysis of steam–air condensation in vertical porous-wall corrugated tubes

In the passive containment cooling systems (PCCS) of Generation III nuclear power plants, the condensation performance of steam containing non-condensable gas (NCG) in condensate tubes is critical for removing heat from the nuclear power plant containment. By applying a novel porous-wall corrugated structure in the condenser tubes of this system, it was found that under certain conditions, a double vortex is formed inside and outside the porous layer, which is conductive to the transportation and condensation of steam to the condensing wall. The VOF multiphase relation and Brinkman–Forchheimerextended Darcy model in combination of local heat balance, condensation and species transport equations are employed to describe the heat and mass transfer in the porous-wall corrugated tube during condensation. The numerical analysis are conducted to investigate the effects of waveform channel, porous layer and its thicknesses, subcooling of cooling wall and inlet air mass fractions on flow and condensation under different inlet velocities, in which simulations agree with the experimental data. The results show that the condensation heat transfer is greatly influenced by the vortex distribution and air accumulation. At inflowing velocity of 1 m/s, the waveform with amplitude of 1 mm and wavelength of 18 mm can achieve the highest performance evaluation factor, φ (about 67.5 % enhancement than the smooth tubes). The ratio of waveform amplitude to wavelength needs to be chosen properly in consideration of convection between solid matrix and steam–air mixture in porous layer, and non-condensation gas distribution near cooling wall of condensation tube.

3D Analysis of Hydrogen Distribution and Its Mitigation Using Passive Autocatalytic Recombiners (PARs) Inside VVER-1000 Containment

Hydrogen is a flammable gas that can generate thermal and mechanical loads which could jeopardise the containment integrity upon combustion inside nuclear power plants containment. Hydrogen can be generated from various sources and disperses into the containment atmosphere, mixing with steam and air following a loss of coolant accident and its progression. Therefore, the volumetric hydrogen concentration should be examined within the containment to determine whether a flammable mixture is formed or not. Codes with 3D capabilities could serve this examination by providing detailed contours/maps of the hydrogen distribution inside containment in view of the local stratification phenomenon. In this study, a 3D VVER-1000 as-built containment model was sketched in AutoCAD and then processed into GOTHIC nuclear containment analysis code for hydrogen evaluation. The model was modified to a great extent by installing 80 passive autocatalytic recombiners and locating hydrogen sources to evaluate the performance of the hydrogen removal system inside the containment on maintaining the hydrogen concentration below the flammability limit during a large break loss of coolant accident. 2D profiles and 3D contours of volumetric hydrogen concentration with and without PARs are presented as the simulation outcome of this study. The results were validated against the results of the Final Safety Analysis Report, which also demonstrates the effectiveness of the hydrogen removal system as an engineered safety feature to keep the containment within a safe margin. Detailed 3D contours of hydrogen distribution inside containment can be employed to evaluate the local hot spots of hydrogen, rearranging and optimising the number and location of PARs to avoid the hydrogen explosion inside containment.

Two-way Eulerian-Lagrangian coupling approach in GASFLOW code for containment spray modelling

Spray droplets are considered to have significant effects on the gas mixing and depressurization in the containment of nuclear power plants during hypothetical accident scenarios. The Eulerian-Lagrangian approach, which enables the two-way mass, momentum, and energy coupling of the continuous gas and the dispersed spray droplets, is promising to model the behaviours of spray droplets in 3D simulations of the full-scale containment. Two-way heat and mass transfer models have been developed in the 3D CFD code GASFLOW. The particle group method is applied in the Lagrangian approach to track the droplets. The ordinary differential equations which couple the local heat and mass exchanges between each droplet group and the surrounding gas mixtures have been solved using the Runge-Kutta method. The GASFLOW coupling this new approach is validated by TOSQAN water spray experiments, which demonstrate the good feasibility of implementing the approach. The analysis of the impact of the spray on containment atmosphere indicates that: 1) the droplet swarm entrains and mixes the surrounding gas, breaking up the gas stratification, with turbulent diffusion dominating momentum mixing in dead zones; 2) the thermal and mass exchanges between the spray droplet swarm and gas jointly determine the atmosphere depressurization, with the evaporation of the spray-formed film/sump on heat structures of great importance for the containment pressure development; and 3) these spray thermodynamic and dynamic effects highly depend on the droplet size distribution and the spray shape.

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.

Analysis of time-varying mechanical properties of prestressed concrete containment during the tensioning process and service considering the influence of creep

The containment structure is an important barrier that ensures the safe operation of nuclear power plants. Currently, most studies on concrete creep in containment structures assumed constant stresses in the calculation of creep, which is not consistent with realistic cases or circumstances. In this study, a combination of theoretical analysis and numerical simulation was used to investigate the prestressed concrete containment (PCC) of nuclear power plants. Based on the age-adjusted effective modulus method, three widely accepted concrete creep models were improved to obtain a variable stress creep model. Using the improved creep model, the effect of concrete creep on the PCC structure during prestressed tensioning and long-term service was systematically analyzed. Then, the internal pressure bearing capacity of the PCC in long-term service was assessed, and the effects of long-term concrete creep on the mechanical properties of the PCC were evaluated.

A scaled test on the damage and vibration behavior of reinforced concrete nuclear containment subjected to a large aircraft impact

After the 9/11 terrorist attacks in the United States, the ability of nuclear power plants (NPPs) to resist the malicious impact of large commercial aircraft has become a focus in the nuclear security field. To investigate the damage and vibration characteristics of NPP containment against a large aircraft impact, an experiment of an airplane impacting a reinforced concrete (RC) nuclear containment structure model was carried out using a rocket sled launching facility. In the experiment, a 1/15 scaled RC nuclear containment structure model was designed and built according to a certain type of NPP building prototype. The total impact process was recorded by a high-speed photography system. Moreover, the corresponding test data and physical parameters of the nuclear containment model were obtained, including the damaged area of the RC nuclear containment model, acceleration time histories, reinforcement steel strain time histories, etc. It is found that the containment model exhibits local damage when the large aircraft model collides with a velocity of 142.3 m/s. However, reinforcement prevents the large aircraft model from perforating the NPP containment effectively. In this paper, the damage and failure rules and characteristics of RC nuclear containment under the impact of a large aircraft are discussed, and the impact vibration response characteristics of nuclear containment are preliminarily analyzed based on the test. The related studies and conclusions can provide a basis for NPP structural optimization and the safety assessment of large commercial aircraft collision problems and can also provide data support for finite element (FE) modelling.

Improvement of diffusion layer model for steam condensation in the presence of non-condensable gas under turbulent free convection

Steam condensation plays a crucial role in predicting the pressure behavior and hydrogen distribution in nuclear power plant containment during a postulated loss-of-coolant accident or a severe accident. The stagnant film model and diffusion layer model are two commonly used heat and transfer analogy models for predicting film-wise condensation in the presence of non-condensable gas on a vertical wall in the nuclear engineering field. This study proposed a general formula for the equivalent condensation thermal conductivity, which can be used to convert between the traditional diffusion layer model and the stagnant film model. An improved diffusion layer model is established, which can be used to analyze steam condensation in the presence of non-condensable gas on a vertical flat surface or outer surface of condenser tubes under turbulent free convection. The density difference of water vapor at the bulk and at the gas–liquid interface is the driving force for steam condensation in the improved model, the traditional diffusion layer model’s assumption, e.g., diffusion layer thickness, constant gas mixture density, and the application of modified Clausius-Clapeyron equation, are no longer used. The newly established model also divides the condensation process into liquid film heat transfer, latent heat transfer, and sensible heat transfer. The three parts of heat transfer processes are linked together by the equivalent heat flux on both sides of the gas–liquid interface. The effects of gas compressibility, suction, fog formation, and liquid film waviness have been considered in this study. The influences of gas compressibility, the thermal resistance of condensate film, and sensible heat transfer on modeling steam condensation were discussed. Compared with various sources of experimental data (619 data points), more than 97% of the model predicted results are within the ±25.0% error band, and the mean absolute relative error is 10.2%, which indicates that the newly proposed model has high accuracy and widely applicable ranges.

Development of two-dimensional hydrogen distribution model in small-scale spaces under severe accidents

Hydrogen distribution and combustion are important problems in nuclear reactor safety analysis under severe accident conditions.IPWR (Integrated pressurized water reactor) are very different from conventional nuclear power plants in design, structure and operating power. In particular, the nuclear power plant containment has a large space volume, and there are multiple compartments inside, while the small reactor containment has a small volume and a simple and compact internal layout. This makes the hydrogen in the containment behave differently in a severe accident. Thus, the hydrogen flow distribution characteristics of IPWR under severe accidents need further study.In this paper, a two-dimensional continuum model of hydrogen distribution in small-scale under severe accident conditions is established by using finite volume method and staggered grid scheme based on hydrodynamics calculation method. SIMPER (Semi-Implicit Method for Pressure Linked Equations Revised) algorithm is used to construct the pressure correction equation, alternate direction implicit algorithm is used to solve the algebraic equations, and block correction algorithm is used to accelerate the iterative convergence. Compared with the experimental study of hydrogen flow distribution characteristics in local compartments, the model can predict the change of hydrogen volume fraction distribution well with time in the experiment. After that, the flow distribution characteristics of hydrogen in the RPV (Reactorpressurevessel) and containment under the low-pressure core melt accident were analyzed with IPWR-IP200 as the research object. The results show that the coolant level in the lower head of the RPV is still about 0.8 m during hydrogen generation, so there is no hydrogen in the lower head area. The highest hydrogen volume fraction in the RPV is at the outlet of the core, which is about 23 %. The highest volume fraction of hydrogen in the containment is the outlet above the pressurizer relief valve, which is up to 17 %. Because of the high volume fraction of water vapor in the RPV, there is no risk of hydrogen combustion in the RPV. During hydrogen generation, there is a risk of hydrogen combustion within the containment, and the risk area is located in the upper space of the containment. After hydrogen production ceases, as water vapor continues to enter the containment, the hydrogen volume fraction gradually decreases, and the risk of hydrogen combustion will also decrease.The two-dimensional continuum model established in this paper can predict the hydrogen distribution under severe accidents, which can provide a reference for the hydrogen mitigation strategy of IPWR.

Application of CFD model for passive autocatalytic recombiners to formulate an empirical correlation for integral containment analysis

Hydrogen mitigation using Passive Autocatalytic Recombiners (PARs) has been widely accepted methodology inside reactor containment of accident struck Nuclear Power Plants. They reduce hydrogen concentration inside reactor containment by recombining it with oxygen from containment air on catalyst surfaces at ambient temperatures. Exothermic heat of reaction drives the product steam upwards, establishing natural convection around PAR, thus invoking homogenisation inside containment. CFD models resolving individual catalyst plate channels of PAR provide good insight about temperature and hydrogen recombination. But very thin catalyst plates compared to large dimensions of the enclosures involved result in intensive calculations. Hence, empirical correlations specific to PARs being modelled are often used in integral containment studies. In this work, an experimentally validated CFD model of PAR has been employed for developing an empirical correlation for Indian PAR. For this purpose, detailed parametric study involving different gas mixture variables at PAR inlet has been performed. For each case, respective values of gas mixture variables at recombiner outlet have been tabulated. The obtained data matrix has then been processed using regression analysis to obtain a set of correlations between inlet and outlet variables. The empirical correlation thus developed, can be easily plugged into commercially available CFD software.

Seismic Fragility and Risk Assessment of a Nuclear Power Plant Containment Building for Seismic Input Based on the Conditional Spectrum

A procedure for the seismic fragility assessment of nuclear power plants by applying ground motions compatible with the conditional probability distribution of a conditional spectrum (CS) is presented with a case study of a containment building. Three CSs were constructed using different control frequencies to investigate the influence of the control frequency. Horizontal component-to-component directional variability was introduced by randomly rotating the horizontal axes of the recorded ground motions. Nonlinear lumped mass stick models were constructed using variables distributed by Latin hypercube sampling to model the uncertainty. An incremental dynamic analysis was performed, and seismic fragility curves were calculated. In addition, a seismic input based on a uniform hazard response spectrum (UHRS) was applied to the seismic fragility assessment for comparison. By selecting a control frequency dominating the seismic response, the CS-based seismic input produces an enhanced ‘high confidence of low probability of failure’ capacity and lower seismic risk than the UHRS-based seismic input.

Experimental Study on Hydrogen Recombination Characteristics of a Passive Autocatalytic Recombiner during Spray Operation

During an accident, hydrogen distribution in a containment building of a nuclear power plant (NPP) and characteristics of hydrogen depletion by passive autocatalytic recombiners (PARs) differ depending on the thermal-hydraulic behaviors occurring in the containment. A spray system installed in the NPP containment to control the pressure in accident conditions may interact with PAR operations. This study intended to experimentally evaluate the hydrogen removal characteristics of a grid-type PAR when a spray was operating. For the experimental simulation of hydrogen recombination characteristics of the PAR affected by a spray operation, we used the SPARC experimental facility, which was equipped with a pressure vessel capable of controlling the wall temperature with a volume of 82 m3. To measure gas species concentrations, 14 probes each for hydrogen, oxygen, and water vapor were installed. Two tests were designed depending on the spray initiation time. The SSP3 test was an experiment to simulate an accident in which the PAR operated as hydrogen was released after the spray is activated, and the SSP4 test was an experiment to simulate an accident in which the spray began after the operation of the PAR was initiated by the hydrogen release. In the experiment, two contradictory results were obtained, which were an increased start-up delay time of the PAR by early initiation of the spray and a negligible impact on the PAR’s performance by the delayed initiation of the spray.

Aircraft-structure interaction study for the assessment of the reaction-time response of nuclear containment structure

A systematic finite element (FE) investigation has been presented in this study to basically examine the two basic constraints (deformation and curvature) in the conventional reaction-time response approach for studying the behavior of a nuclear containment structure. A coupled finite element investigation has been carried out wherein the reaction-time response of the Boeing 707–320 and Boeing 747–400 has been obtained against the rigid flat and curved targets of diameter 42 and 100 m in order to study the possible influence of curvature on the reaction-time response curve. The aircrafts were subsequently considered to hit the deformable containment of Boiling Water Reactor Nuclear Power Plant (BWR-NPP) of diameter 42 m. The effect of target deformability thus obtained in the reaction-time response has been compared with that of the one obtained against the rigid target of the same curvature. Further, the numerical as well as analytical approaches have been employed to evaluate the exact impact area at the interface of the aircraft and flat rigid target as a function of time and incidence velocity. Finally, the behaviour of BWR-NPP has been studied against Boeing 707–320 and Boeing 747–400 using the coupled approach, traditional average contact area approach and area trifurcation (fuselage, first set engines and second set engines) approach. The response of the containment thus obtained against the three approaches has been compared.

Study of an improved and novel venturi scrubber configuration for removal of radioactive gases from NPP containment air during severe accident

Owing to the rising concerns about the safety of nuclear power plants (NPP), it is essential to study the venturi scrubber in detail, which is a key component of the filtered containment venting system (FCVS). FCVS alleviates the pressure in containment by filtering and venting out the contaminated air. The main objective of this research was to perform a CFD investigation of different configurations of a circular, non-submerged, self-priming venturi scrubber to estimate and improve the performance in the removal of elemental iodine from the air. For benchmarking, a mass transfer model which is based on two-film theory was selected and validated by experimental data where an alkaline solution was considered as the scrubbing solution. This mass transfer model was modified and implemented on a unique formation of two self-priming venturi scrubbers in series. Euler-Euler method was used for two-phase modeling and the realizable K−ε model was used for capturing the turbulence. The obtained results showed a remarkable improvement in the removal of radioactive iodine from the air using a series combination of venturi scrubbers. The removal efficiency was improved at every single data point.

Vibration Monitoring of Nuclear Power Plant Containment Buildings During the Integrated Leakage Rate Test for Structural Condition Assessment

Abstract This study investigates the use of the operational vibrations produced during the Integrated Leak Rate Test of nuclear power plant containment buildings for further informing on its mechanical condition. The experiment is performed on a 1:3-scale containment building mock-up. The results show that meaningful vibrations were generated during the pressurization test. Different features were extracted from the vibration signals and analyzed as a function of the internal pressure. Experimental modal analysis was performed and demonstrated that several frequency peaks generated during the pressurization cycle effectively corresponded to the eigenmodes of the containment building. The identified operational frequency modes exhibited remarkable hysteretic dependencies on the internal pressure. The latter was phenomenologically described through a simplified two-dimensional (2D) finite element model of the vessel. Besides, a surrogate statistical model based on the Principal Component Analysis of the vibration data was proposed as a baseline and so detect abnormal behavior. Then, different synthetic damage scenarios were created by subtlety altering the recorded signals and ultimately substantiate the capability of the statistical model to detect these odd signals. Finally, conclusions were drawn regarding the possibility of using mechanical vibrations for assisting in the licensing process of nuclear power plants and monitor the structural health condition of in-service containment buildings.