Containment Building Research Articles

Deployment and performance of a Low-Energy-Threshold Skipper-CCD inside a nuclear reactor

Charge Coupled Devices (CCD) are being used for reactor neutrino experiments and have already demonstrated their potential in constraining new physics models. The prospect of a Skipper-CCD experiment looking for standard and beyond standard model (BSM) physics in a nuclear reactor has been evaluated for different benchmark scenarios. Here, we report the first installation of a 2-g Skipper-CCD inside the containment building of a 2 GWth nuclear power plant and analyze its performance throughout its first 18 months of operation. The sensor was successfully deployed at Atucha II, in Argentina, 12 meters away from the center of the reactor core. We discuss the challenges involved in the commissioning of the detector and present data acquired during reactor ON and reactor OFF periods, with the sensor functioning with a sub-electron readout noise of 0.17 e−. Based on an exposure of 56.8 g day reactor ON and two reactor OFF data sets with a total exposure of 118.1 g day we characterize the system and evaluate the sensitivity to CEvNS. We achieved a background rate of 33 kdru and a low threshold of 45 eVee. The ongoing efforts to improve sensitivities to CEvNS and BSM interaction are also discussed.

Advanced Sobol sensitivity analysis of a 1:4-scale prestressed concrete containment vessel using an ANN-based surrogate model

The safety of nuclear power plants can be enhanced by performing sensitivity analyses to identify and evaluate factors that significantly impact their structural integrity. In this study, an artificial neural network-based surrogate model was developed using an experimentally validated finite element model. A Sobol sensitivity analysis was performed using this model to quantitatively assess the impact of material uncertainty on the internal pressure–displacement behavior of containment buildings at various internal pressure levels. The results indicate that the compressive strength of concrete and the prestressing force in the hoop direction critically influence the behavior of containment buildings, with their importance varying according to the internal pressure level. Furthermore, the proposed surrogate model can efficiently substitute for expensive and complex finite element analyses, enabling effective exploration of the multidimensional uncertainty space and quantitative assessment of the major factors affecting structural behavior. The process presented herein facilitates the quantification of factors under various uncertainty scenarios.

Soil-structure interaction analysis of nuclear power plant structure using multi-step method in time-domain

Recently, several countries have experienced strong earthquakes, and there has been growing interest in the seismic design of structures. Particularly, consideration of soil-structure interaction (SSI) has become important when performing seismic response analysis for nuclear power plant (NPP) structures. A seismic analysis method that considers SSI in the time domain has been proposed in previous studies. This method is applicable to the seismic response of structures, because it can consider the influence of the SSI effect with the impedance function in the frequency domain and simultaneously considers the structural nonlinearity. However, it has only been applied to the structure comprising a single beam-stick. In this study, a modified method for applying it to more general structures is proposed, and its applicability is validated with an actual reactor containment building (RCB) structure comprising a three beam-sticks. Validation was performed by comparing the seismic responses with those of ACS SASSI. Using the proposed method, SSI analysis was performed with three soil profiles and the results were compared to investigate the effect of SSI on seismic responses, including the in-structure response spectrum (ISRS) and time history acceleration. For more realistic analysis, soil profile and earthquake input motion occurred in South Korea are adopted. Further, a multi-step method considering nonlinear behavior was also performed and compared with the linear case.

Seismic Fragility Assessment of Equipment in Nuclear Power Plant Structures Considering Nonlinear Hysteretic Behavior of Squat Walls

Accurate seismic risk assessment of structures necessitates precise fragility analysis. This study focuses on the seismic fragility assessment of equipment in nuclear power plant structures, particularly emphasizing the influence of nonlinear hysteretic behavior exhibited by squat walls. By developing and comparing linear and nonlinear models of reactor containment and auxiliary buildings within a nuclear power plant, this research elucidates the significant impact of nonlinear behaviors on the seismic response and subsequent fragility curves of associated equipment and systems. Utilizing a lumped-mass stick model, the study efficiently captures the characteristics of squat shear walls, facilitating the calculation of floor response spectra and fragility for nuclear facility structures. The impacts of pinching and degradation, inherent characteristics of a squat wall’s hysteresis under cyclic loading, were identified as significant. These include the shift and change in amplitude of the floor response’s peak, along with amplification in high-frequency domains. Fragility assessments, informed by seismic response analysis, indicate significant variations in fragility curve parameters based on equipment location and frequency. The findings question the effectiveness of traditional response factor methods and general regression techniques in accurately addressing the complex probabilistic seismic demands of equipment, thus highlighting the importance of using direct analysis for calculating fragility in situations with significant nonlinearity.

Validation and Application of a Code for Three-Dimensional Analysis of Hydrogen–Steam Behavior in a Nuclear Reactor Containment during Severe Accidents

In a pressurized water reactor (PWR) during a loss of coolant accident (LOCA) or a station blackout (SBO) accident, water and steam are released into the containment building. The water vapor mixes with the atmosphere, partially condensing into droplets or condensing on the containment walls. Although a significant amount of water vapor condenses, it coexists with hydrogen generated by the reactor core oxidation. As water vapor condenses, the volume fraction of hydrogen increases, raising the risk of explosion or flame acceleration. As such, water vapor’s behavior directly affects hydrogen distribution. To conservatively evaluate hydrogen safety in a PWR during a severe accident, lumped-parameter codes have been heavily used. As a best-estimate approach for hydrogen safety analysis in a PWR containment, a turbulence-resolved CFD code called contain3D has been developed. This paper presents the validation results of the code and simulation results of hydrogen behavior affected by water vapor condensation and hydrogen removal by passive autocatalytic recombiners (PARs) in the APR1400 containment. The results provide insight into the three-dimensional behaviors of the hydrogen in the containment.

Multi-Layered Shell Model and Seismic Limit States of a Containment Building in Nuclear Power Plant Considering Deterioration and Voids

Multi-Layered Shell Model and Seismic Limit States of a Containment Building in Nuclear Power Plant Considering Deterioration and Voids

Design Change and Operational Consideration of the HVAC System during Nuclear Power Plant Decommissioning

The heating, ventilation, and air conditioning (HVAC) system plays a crucial role in ensuring the safety of workers and preventing the release of gaseous radioactive materials into the environment during the decommissioning of a nuclear power plant (NPP). To establish an HVAC operation strategy, decommissioning phases were divided into four stages, and the HVAC systems were reclassified. In addition, assumptions have been made regarding design modifications and maintenance for the reactor containment building (RCB) HVAC, fuel handling building (FHB) HVAC, and main control room (MCR) HVAC. Based on these, for RCB HVAC operation, natural ventilation and RCB purge operation during the transition period are proposed. In the decommissioning stage, recirculation operation, entire ventilation operation consisting of continuous operation and purge operation, and finally partial ventilation operation to purify local space were proposed. Moreover, during the transition period, the FHB HVAC was proposed to operate as normal NPP, and the MCR HVAC was suggested to operate with safety-related equipment removed.

Study on spent fuel heatup during spent fuel pool complete loss of coolant accident

Abstract After years of operation, nuclear power plants accumulate a substantial amount of spent fuel stored in spent fuel pools. Prior to the completion of dry storage facilities, this spent fuel remains in the spent fuel pools for several decades. The progression of accidents in spent fuel pools leading to fuel damage typically spans several days or even weeks, which may lead people to overlook the importance of spent fuel pool safety. However, accidents involving spent fuel pools have a characteristic of low probability but high consequences. Their significance has been reemphasized, especially after the Fukushima accident. This study conducts an analysis of the scenario where a spent fuel pool experiences a complete loss of water. A plant equipped with a BWR-4 reactor and a MARK-I containment structure is used as the reference plant. The analysis utilizes the MAAP5 code to assess the impact of different fuel cooling times and ventilation flow rate of secondary containment building on the fuel temperature variation, considering only air circulation for cooling. Based on the analysis results, the plant utilizes the normal ventilation system of the reactor building, which can maintain the fuel temperature below 565 °C with a cooling time of 1 year. Additionally, it was found in the study that adequate ventilation flow rates and sufficient cooling time can ensure that the fuel temperature remains below 565 °C, thereby preserving fuel integrity.

Accelerated construction of projection-based reduced-order models via incremental approaches

We present an accelerated greedy strategy for training of projection-based reduced-order models for parametric steady and unsteady partial differential equations. Our approach exploits hierarchical approximate proper orthogonal decomposition to speed up the construction of the empirical test space for least-square Petrov–Galerkin formulations, a progressive construction of the empirical quadrature rule based on a warm start of the non-negative least-square algorithm, and a two-fidelity sampling strategy to reduce the number of expensive greedy iterations. We illustrate the performance of our method for two test cases: a two-dimensional compressible inviscid flow past a LS89 blade at moderate Mach number, and a three-dimensional nonlinear mechanics problem to predict the long-time structural response of the standard section of a nuclear containment building under external loading.

Mixing of Hydrogen-Steam Buoyancy Jets Released into the Upper Atmosphere of the Containment Building

This study evaluates the spatial dilution of hydrogen concentration caused by steam-hydrogen buoyancy jets rising through the open top of the steam generator compartment during a loss-of-coolant accident in the OPR1000, the Korean Standard Nuclear Power Plant. The correlation of the concentration decay rate in the plume with relatively high buoyant flux was applied to estimate the hydrogen concentration in the rise distance of the buoyant jet. The MELCOR code was used to calculate the gas composition and discharge flow rate in the ruptured cold leg during the rapid cladding oxidation to determine the volume and buoyant fluxes that affect the mixing behavior. The concentration decay rate at the plume’s center decreases as the steam-hydrogen binary buoyant jet rises. Despite the assumed initial volume flux and simplified jet nozzle geometry, the decay rate correlation can assess conservatively the diluted hydrogen in a severe accident.

Analysis of heat-loss mechanisms with various gases associated with the surface emissivity of a metal containment vessel in a water-cooled small modular reactor

In various small modular reactor (SMR) designs currently under development, the conventional concrete containment building has been replaced by a metal containment vessel (MCV). In these systems, the gap between the MCV and the reactor pressure vessel is filled with gas or vacuumed weakly, effectively suppressing conduction and convection heat transfer. However, thermal radiation remains the major mode of heat transfer during normal operation. The objective of this study was to investigate the heat-transfer mechanisms in integral pressurized water reactor (IPWR)-type SMRs under various gas-filled conditions using computational fluid dynamics. The use of thermal radiation shielding (TRS) with a much lower emissivity material than the MCV surface was also evaluated. The results showed that thermal radiation was always the dominant contributor to heat loss (48–97%), while the conjugated effects of the gas candidates on natural convection and thermal radiation varied depending on their thermal and radiative properties, including absorption coefficient. The TRS showed an excellent insulation performance, with a reduction in the total heat loss of 56–70% under the relatively low temperatures of the IPWR system, except for carbon dioxide (13%). Consequently, TRS can be utilized to enhance the thermal efficiency of SMR designs by suppressing the heat loss through the MCV.

Removal Characteristics of Ni-Ferrite Oxide Film Formed on the Stainless Steel Surface in Fiber Laser Decontamination

Every nuclear facility has a limited operating life, at the end of which they must be shut down and decommissioned. For decommissioning, the removal of radioactive contaminants and demolition should be considered. Before nuclear power plants are decommissioned, the primary system, which is the most heavily contaminated, is decontaminated first, followed by decontamination of the surfaces of containment containers and buildings. This study examined the removal characteristics of Ni-ferrite coated on the stainless steel (STS304) specimen surface using an eco-friendly laser that generates little secondary waste, and derived the optimal efficiency conditions for laser decontamination. In this study, simulated oxide film specimens were fabricated by coating the STS304 substrate using non-radioactive Ni-ferrite. The experimental equipment for laser decontamination was a Q-switched fiber laser with an average power of 100 W. The laser beam was irradiated on the surface of the specimen according to the beam overlap rate. In addition, the characteristics of the decontaminated area and removal of the coating layer were checked using digital 3D microscopy, scanning electron microscope (SEM), and X-ray diffraction (XRD). The results of laser decontamination experiment indicated the possibility of coating layer removal with fewer scans as the x- or y-axis directions beam overlap rate increased. However, the surface was discolored when excessive overlaps occurred. This study provided basic data for laser decontamination applicable to the removal of Ni-ferrite oxide films on STS304.

Numerical analysis of sodium-argon two-phase flow in narrow channels under severe accident conditions in pool type SFR

A two-phase flow model for sodium-argon mixture flow through the reactor vessel head penetrations under the energetic core disruptive accident in a sodium fast reactor (SFR) is presented. The model evaluates the total sodium mass in sodium-argon mixture released to the reactor containment building (RCB) through penetrations. The driving pressure at the penetration entrance is given as input. The two-phase frictional and local pressure drops, inertial and piezo-metric heads are accounted for in the model. Depending on the argon volume fraction, either a homogenous or separated flow approach is adopted to evaluate the mixture flow. Reactor-scale analysis show that the presence of argon in vessel head penetrations can significantly mitigate the total sodium release mass during the slug impact phase. Also, the two-phase frictional pressure drop plays a mitigating role and needs to be included in the model formulation to avoid overestimation of the sodium release mass. Fluid-structure interaction effects such as vessel head bending and hold-down bolt elongation have negligible influence on sodium release mass as long as the deformations are within the elastic limit. The model is conceived to be part of an integrated code system to evaluate the sodium fire and the maximum RCB pressure under severe accident conditions in a pool-type SFR.

A study on transport and plugging of sodium aerosol in leak paths of concrete blocks

In the event of a severe accident in Sodium Cooled Fast Reactors (SFR), the sodium combustion aerosols along with fission product aerosols would migrate to the environment through leak paths of the Reactor Containment Building (RCB) concrete wall under positive pressure. Understanding the characteristics of sodium aerosol transport through concrete leak paths is important as it governs the environmental source term. In this context, experiments are conducted to study the influence of various parameters like pressure, initial mass concentration, leak path diameter, humidity etc., on the transport and deposition of sodium aerosols in straight leak paths of concrete. The leak paths in concrete specimens are prepared by casting and the diameter of the leak path is measured using thermography technique. Aerosol transport experiments are conducted to measure the transported and plugged aerosol mass in the leak paths and corresponding plugging times. The values of differential pressure, aerosol concentration and relative humidity taken for the study are in the ranges 10–15 kPa, 0.65–3.04 g/m3 and 30–90% respectively. These observations are numerically simulated using 1-Dimensional transport equation. The simulated values are compared with the experimental results and reasonable agreement among them is observed. From the safety assessment view of reactor, the approach presented here is conservative as it is with straight leak paths.

Active Infrared Thermography applied for concrete structures inspection in Nuclear Power Plants

Due to the new requirements that are being demanded of the Nuclear Power Plants for their License Renewal (LR), Life Extension (LE) and Long Term Operation (LTO), it is necessary to verify the integrity of several civil structures, such as the containment building, used fuel building and cooling towers, among others. In general, the objectives of Non-Destructive Test methods (NDT) for the inspection of concrete structures in nuclear power plants can be arranged into: location of the reinforcement and tendons, determination of the state of the concrete matrix, determination of the corrosion of the metallic armor and detection of cracks, voids, inclusions and delaminations in the matrix of the structure. For this, Tecnatom, based on the experience acquired in the field of NDT, has developed a set of techniques for determining the structures condition including surface hardness testing, ultrasound (transverse wave pulse-echo and transmission speed), electromagnetic methods (Ground Penetrating Radar - GPR) and non-contact techniques such as Infrared Thermography (IRT). The method and technique best suited for a given inspection has to be selected depending on the objective sought, the typical defectology of the component, the characteristics of the structure and the conditions of the inspection itself. In this work it is presented an example of the development and in-field application of the IRT method for the inspection of a concrete structure in a nuclear power plant to detect possible delaminations caused by the corrosion of the metallic reinforcement.

Benchmark VERCORS 2022: blind prediction of time-dependent behavior of concrete containment building with low and high-fidelity models

The VERCORS program aims to acquire an extensive experimental dataset collected to provide a solid basis for numerical modeling of concrete containment buildings (CCBs). In order to cover the entire life-span of a real containment, the measurements are done on 3× smaller mock-up which leads to 9-fold acceleration of all processes related to drying. The goal for the participants of the third VERCORS benchmark was to predict the behaviour of the CCB based on standard laboratory measurements on VERCORS concrete. The previous paper [1] presented the calibration procedure of material models for moisture transport and time-dependent behavior of concrete and summarized the results obtained with a computationally efficient low-fidelity model (LFM). The present paper compares the responses of the LFM and a high-fidelity model (HFM) with a detailed geometry of the entire containment and presents a comparison with the experimental data collected over the last 8 years on the VERCORS mockup.

Quantification of the Effective Detectable Period for Concrete Voids of CLP by Lock-In Thermography

This study is to inspect the voids between the concrete containment building and the containment liner plate (CLP) in the light-water reactor type nuclear power plant with lock-in thermography (LIT) inspection technology. For that, a finite element method (FEM) model containing concrete voids was created, and the thermal distribution change of the CLP surface was simulated through numerical analysis simulation of various LIT inspection conditions and converted with real-time thermography data. For the simulated temperature distribution image and the amplitude and phase images calculated by the four-point method, the signal-to-noise ratio (SNR) is analyzed based on the sound area and void areas. As a result, the difference in SNR according to the size of voids was remarkable, and the effective detectable period (EDP), which was common to each inspection condition, was derived. Furthermore, a CLP concrete mockup identical to the model shape is produced, and the thermal image of the EDP is analyzed through the experiment with the same analysis technique, and the results are compared. Although there are some differences between the numerical analysis conditions and the experimental environments, the deduction and utilization of EDP through FEM simulation are considered useful approaches to applying LIT to inspect concrete voids on the back of the CLP.

Study on seismic safety evaluation of crossover pipeline connecting containment and turbine building in three-dimensional isolated nuclear power plant

Isolation design enhances the safety of nuclear islands during earthquakes. Magnification of isolation layer displacement increases the damage risk of the nuclear pipeline crossing the isolation gap. A simplified safety evaluation procedure for pipelines based on elbow displacement limits and damage indices was proposed. An analytical model of the nuclear elbow was first developed for unidirectional and multi-directional displacement loading. The failure point is determined based on the elbow force–displacement curve and adjusted to obtain a safe displacement limit. Subsequently, a 3DBNL model with a horizontal isolation frequency of 0.5 Hz and a vertical isolation frequency of 1.25 Hz was proposed based on the BNL model. Nonlinear time history analysis was performed to obtain the dynamic responses such as nuclear island acceleration, isolation layer displacement, and pipe relative displacement. Finally, the damage index was calculated as the ratio of the peak relative displacement to the elbow displacement limit. The damage index and the distribution interval under multi-directional seismic input were calculated separately to assess pipeline safety. The results indicate that the three-dimensional isolated nuclear island remains safe under the horizontal 0.6 g and vertical 0.4 g seismic input, while the pipeline failed under the horizontal 0.3 g and vertical 0.2 g seismic input.

A Study on Robust Optimal Sensor Placement for Real-time Monitoring of Containment Buildings in Nuclear Power Plants

A Study on Robust Optimal Sensor Placement for Real-time Monitoring of Containment Buildings in Nuclear Power Plants

A numerical approach for assessing internal pressure capacity at liner failure in the expanded free-field of the prestressed concrete containment vessel

Since containment building is the major shielding structure to ensure safety of nuclear power plant, the structural behavior and ultimate pressure capacity of containments must be studied in depth. This paper addresses ambiguous issue of determining free-field position for liner failure by suggesting an expanded free-field region and comparing internal pressure capacities obtained by test data, conservative assumption and suggested free-field region. For this purpose, a practical approach to determine the free-field position for the evaluation of liner tearing is carried out. The maximum principal strain histories versus internal pressure capacities among different free-field positions at various azimuths and elevations are compared with those at the equipment hatch as a conservative assumption. The comparison shows that there are considerable differences in the internal pressure capacity at liner failure within the expanded free-field region compared to the vicinity of the equipment hatch. Additionally, this study proposes an approximate correlation with conservative factors by considering the expanded free-field ranges and material characteristics to determine realistic failure criteria for liner. The applicability of the proposed correlation is demonstrated by comparing the internal pressure capacities of full-scale containment buildings following liner failure criteria according to RG 1.216 and an approximate correlation.