Safety-related Structures Research Articles

Initial Event Analysis for Turbine Missile Events at Floating Nuclear Power Plants Using the Master Logic Diagram Method

Floating Nuclear Power Plants (FNPP) have become a popular design choice in several countries, including South Korea, China, and Denmark. Safety analysis, including probabilistic safety assessment, must be conducted for every nuclear reactor design. Among the parameters that need to be determined are initiating events. Initiating events is crucial information as it impacts further analyses, such as event and fault tree analyses. Various methods exist for deciding and initiating events, including the Master Logic Diagram (MLD). The determination of initiating events can be based on the potential impact of internal or external hazards on the FNPP. This study aims to demonstrate the application of MLD in determining initiating events in the case of a turbine missile incident on Akademik Lomonosov floating NPP. Turbine missile events are critical due to their potential to damage important safety-related structures, systems, and components in the vicinity. From the analysis, it was found that ten basic events could be candidates for initiating events. After the screening process, four initiating events were identified as the final result: reactor vessel rupture (RVR), loss of feedwater (LOFW), feedwater line break (FWLB), and loss of condenser vacuum (LOCV). Keywords: Floating NPP, safety analysis, initiating event, master logic diagram (MLD)

Hybrid simulation testing of two-storey low-aspect-ratio nuclear RC shear walls with normal- and high-strength reinforcement: Seismic performance evaluation and economic assessment

Low-aspect-ratio reinforced concrete (RC) shear walls have been commonly used in several nuclear facilities in containment and safety-related structures. Despite being a potential alternative to reduce rebar congestion and subsequently minimize complex construction activities typically associated with nuclear facilities, there has been limited experimental research on investigating the impact of using high-strength reinforcement (HSR) on the seismic performance of such walls, particularly in a multi-storey context. This lack of research is mainly due to considerable challenges imposed when testing such multi-storey nuclear RC shear walls in most laboratories. Therefore, the current study presents the experimental results of two two-storey low-aspect-ratio nuclear RC shear walls that were tested utilizing the seismic hybrid simulation testing technique. In this respect, walls W1-NSR and W2-HSR were designed using normal-strength reinforcement (NSR) and HSR, respectively, where the two test walls had comparable capacities to allow for direct comparisons. Both walls were subjected to various ground motion levels, spanning from operational to design and beyond-design earthquake scenarios. The experimental findings are then presented to include the force-displacement responses, the multi-storey effects, ductility capacities, lateral and rotational stiffnesses, rebar strains, and cracking patterns of the test walls. Subsequently, an economic assessment was carried out to quantify the total rebar weights and the corresponding construction costs of such walls. In addition, the expected seismic repair costs were determined based on a three-dimensional digital image correlation technique that provided information on the damage states of the test walls under different earthquake levels. The results show that although W1-NSR and W2-HSR attained similar force and moment capacities, W2-HSR achieved a relatively higher ductility capacity than W1-NSR. However, larger cracks were observed in W2-HSR compared to W1-NSR, which was attributed to the associated larger rebar spacing in the former relative to the latter. The economic assessment results demonstrate that using HSR minimized the rebar weights and construction costs, while both walls had similar seismic repair costs at their design and beyond-design earthquake levels. Both the seismic performance and economic assessment results presented in the current study are expected to aid future editions of relevant design standards in adopting HSR in nuclear construction practice.

Reviewing Welding Procedures—Checklists for Nuclear Power Systems

Abstract To ensure effective control of the welding process, nuclear power architecture and engineering (AE) companies routinely undertake the review of procedure qualification records (PQRs) and welding procedure specifications (WPSs) for safety-related structures, systems, and components (SSCs), as qualified by manufacturers and construction contractors. Given the substantial quantity of PQRs and WPSs necessitating evaluation, along with the multitude of variables inherent in these documents, a set of review checklists has been devised. These checklists, developed from the perspective of AE companies' welding engineers, transform welding procedure qualification rules and requirements into intuitive descriptions, assessments of correctness, or numerical comparisons of scale. AE companies' welding engineers employ these review checklists to scrutinize PQRs and WPSs, facilitating a comprehensive, accurate, and efficient review process.

Effects of boundary conditions on RC beams subjected to accidental thermal load in nuclear power plants

Reinforced concrete (RC) beams in nuclear safety related structures should be designed for accidental thermal gradient combined with mechanical loads. Design capacity and structural behavior of those beams are affected by constraint boundary conditions at beam ends. In design approaches widely used in nuclear power plants (NPPs), it is commonly assumed that the thermal curvature of a RC member is fully constrained while axial constraint condition is often not specified. In this paper, RC beams with various constraint boundary conditions that are subjected to accidental thermal gradient combined with mechanical loads are studied using both experimental tests and nonlinear finite element (NLFE) models. Both rotational and axial stiffness ratios at boundary and their effects on thermal moments and ultimate flexural capacity of RC beam are evaluated. Results shows that the assumption of a fully constrained rotational boundary condition is practically accurate for most RC beams in NPPs subjected to thermal gradient. On the other hand, results also show that axial constraint boundary condition has considerable influence on both thermal moment and usable design flexural strength thus neglecting its effect may lead to insufficient design of RC beam. In addition, this study established relationship between axial constrain stiffness ratio and thermal moment, as well as relationship between axial constrain stiffness ratio and usable design flexural strength for different accidental temperature gradients and reinforcement ratios. That information can be used for correction of thermal moment and usable design flexural strength in design practice.

Lateral Load-Carrying Capacity of Low-Rise Reinforced Concrete Walls in Nuclear Safety-Related Structures

Nuclear safety-related structures are crucial for ensuring the safety of nuclear facilities and preventing the leakage of radioactive materials, with the primary structural component being low-rise reinforced concrete (LRC) walls. These walls are required to carry combined in-plane axial and horizontal loads, making the accurate prediction of their lateral load-carrying capacity particularly important. In this study, six LRC walls with aspect ratios between 0.33 and 1 were tested and a model for the prediction of the lateral load-carrying capacity of LRC walls was established based on the observed failure mode and plastic limit theory. The parameter in the model was calibrated using the obtained results in this test along with a database containing 131 walls in the literature. Compared to the equations in the American standard ACI 349 and the French standard RCC-CW, the proposed equation is most suitable for assessing the lateral load-carrying capacity of LRC walls in nuclear safety-related structures. The calculated values of the proposed equation exhibit a ratio closest to 1 when compared to experimental values and possess the minimum degree of variation. The computational results reveal that the proposed equations in this study exhibit superior precision and stability.

Exploring medication safety structures and processes in nursing homes: a cross-sectional study

BackgroundMedication safety is important to limit adverse events for nursing home residents. Several factors, such as interprofessional collaboration with pharmacists and medication reviews, have been shown in the literature to influence medication safety processes.AimThis study had three main objectives: (1) To assess how facility- and unit-level organization and infrastructure are related to medication use processes; (2) To determine the extent of medication safety-relevant processes; and (3) To explore pharmacies’ and pharmacists’ involvement in nursing homes’ medication-related processes.MethodCross-sectional multicenter survey data (2018–2019) from a convenience sample of 118 Swiss nursing homes were used. Data were collected on facility and unit characteristics, pharmacy services, as well as medication safety-related structures and processes. Descriptive statistics were used.ResultsMost of the participating nursing homes (93.2%) had electronic resident health record systems that supported medication safety in various ways (e.g., medication lists, interaction checks). Electronic data exchanges with outside partners such as pharmacies or physicians were available for fewer than half (10.2–46.3%, depending on the partner). Pharmacists collaborating with nursing homes were mainly involved in logistical support. Medication reviews were reportedly conducted regularly in two-thirds of facilities.ConclusionA high proportion of Swiss nursing homes have implemented diverse processes and structures that support medication use and safety for residents; however, their collaboration with pharmacists remains relatively limited.

Shaking Table Test for Seismic Response of Nuclear Power Plant on Non-Rock Site

In order to compare and analyze the seismic response characteristics of a safety-related nuclear structure on a non-rock site in the condition of raft and pile group foundations under unidirectional and multidirectional seismic motion input, a large-scale shaking table test of the soil-nuclear structure system was carried out in this paper. In the test, the soil was uniform silted clay, and the shear wave velocity was 213 m/s. Considering the similarity of the superstructure natural frenquency, the actual nuclear power structure was simplified to a three-story frame shear wall structure model. The annular laminated shear model box was used to take the boundary effect of soil into consideration; the seismic motions = were input in only one horizontal direction or three directions at the same time for the shaking table test, and the results were analyzed. The results of the test show that the acceleration response of the safety-related nuclear plant is affected by the directions of input seismic motion and the forms of the foundation. When the seismic motion is input simultaneously in three directions, the acceleration responses of the horizontal motion and vertical rocking of the safety-related plant are larger than those of the single-direction input. The acceleration response of the horizontal motion and vertical rocking of the safety-related structure with the pile group foundation is smaller than that with the raft foundation. The values of most frequency bands in the horizontal acceleration Fourier amplitude spectrum at the top of the pile-foundation structure are smaller than that at the top of the raft-foundation structure, while the displacement is basically the same as that of the raft-foundation structure. This is related to the relation between the frequency component of input seismic motion and the natural frequency of the structure system. Therefore, it is more reasonable to use three-dimensional seismic input in the seismic response analysis of nuclear power plants. The seismic performance of nuclear power plants can be enhanced by using pile group foundations.

Designing and implementing temporary works schemes on UK nuclear licensed sites

Nuclear projects in the UK demand high levels of assurance and substantiation during both design and construction. As a result, site licensees must demonstrate control over construction works in the vicinity of, or modifications to safety-related structures, systems and components in accordance with their nuclear site licence conditions. Furthermore, the Office for Nuclear Regulation will scrutinise designs, management controls and processes prior to permissioning the works. This paper discusses examples of additional considerations for the design and implementation of temporary works during the construction phase to demonstrate compliance with the regulators, Safety Assessment Principles and licence conditions, and to demonstrate that nuclear safety risks have been reduced to as low as reasonably practicable. References of good practice are from a confidential BAE Systems project in Cumbria, UK, titled D58, involving nuclear Class 1 concrete construction and informing of the steps taken to obtain Nuclear Safety Committee and regulatory permissioning.

HELB-induced blast wave evaluation based on mock-up tests and numerical analyses

Postulated high energy line breaks have been considered as a part of design of nuclear safety-related structures, systems and components for a long time. However, recently, technical controversy was raised over the adequacy in assumption of complex phenomena and assessment of dynamic loads. This study is to address blast wave evaluation under a typical main steam line break condition of 1,400MWe power plant. At first, a mock-up facility was developed and six tests were performed by changing rupture pressures and target distances. Corresponding computational fluid dynamics analyses were also carried out after optimization of numerical model. Finally, pressure – time histories caused by plant specific blast wave were generated in use of a theoretical correlation, which were applied to finite element analyses of neighboring steam generator and containment wall. Integrity assessment results as well as the details of tests and analyses data were provided, and their applicability was discussed.

Non-Conservatism of ASME BPVC Section III Division 5 Isochronous Stress–Strain Curves for 316H Stainless Steel at Low Stresses

Abstract Future Gen IV high-temperature reactors are expected to operate above 450 °C where creep effects are significant in safety-related structures, e.g., reactor vessels. The ASME Boiler and Pressure Vessel Code (BPVC) Section III Division 5 provides the rules and methodologies for design of such high-temperature components. Of high relevance to the designer are the isochronous stress–strain curves (ISSCs) part of the rules for deformation limits in the code. The ISSCs are an important method to estimate accumulated inelastic strains at a given stress and duration at elevated temperatures. In this study, the ISSCs for 316H stainless steel in the current edition of the ASME BPVC Section III Division 5 have been reevaluated between 593 °C and 750 °C by adopting a physics-informed minimum creep rate model to reconstruct them. It is demonstrated that the current ASME Section III Division 5 minimum creep rate model underpredicts creep rates compared to experimental data at low stresses (e.g., 650 °C, <40 MPa). By employing a physics-informed minimum creep rate model which captures both diffusive- and dislocation glide/climb-controlled creep regimes, this deficiency is addressed. The ASME ISSCs for 316H stainless steel are then reconstructed by adopting this modified minimum creep rate model. It was found that the ASME ISSCs could underestimate total accumulated strains at ∼σ/σy <0.65 for durations t >1000 h by >10 times which could give rise to non-conservatism in inelastic strain. Experimental data at various temperatures confirm the findings. Potential approaches to address this non-conservatism in inelastic strain and the implications to design are discussed.

Method of Seismic Capacity Analysis of Steel Structure in Nuclear Power Plant

A steel structure in a nuclear power plant is typically constructed next to major safety-related structures. Accordingly, the structural integrity of the steel structure must be achieved until the safety-related building is damaged by external forces. Consequently, the steel structure should have seismic capacity while maintaining the structural integrity of the surrounding safety-related structure. An optimized method for the seismic capacity against the beyond design earthquake was developed to reflect this capacity concept.

Analysis of full-scale aircraft impact to reinforced concrete and steel plate reinforced concrete multiple barriers protecting nuclear power plants

This paper investigates the effectiveness of multiple barriers employed in safety-related concrete structures of nuclear power plants against aircraft impact. These barriers are constructed using reinforced concrete (RC) and steel plate reinforced concrete (SC) structural panels. The performance of such multiple barriers against the impact of a missile or an aircraft is found in many published studies. However, the influence of aircraft impact on the second barriers needs further study. To evaluate the impact resistance of multiple barriers, analysis of full-scale aircraft against RC and SC as the first panel with two different thicknesses (60 cm and 75 cm) are investigated by using finite element code LS-DYNA. The fracture process of aircraft and target panels, damage to the first panel, residual velocities, plastic strain of rebars, impact loads to second panel and energy balance histories are evaluated. The comparison of LS-DYNA analysis results indicates that the maximum force on the second rigid panels is reduced to a maximum of 61% owing to the SC panels. The SC panels experienced fewer residual velocities and impact loads thus showing better performance as compared to the conventional RC panels having the same thicknesses. Resultantly, it is possible to reduce the thickness and cost of safety-related RC structures by adopting steel plates in the design. The finite element-based methodology adopted in this study may be employed to predict the precise response of safety-related structures (barriers) against military and commercial aircraft impacts.

Seismic analysis based structure integrity assessment of steam generator in fast breeder reactor

Abstract Earthquakes, low-probability high-risk events which brings out the deficiency in the design and construction of components and buildings. Many seismic design codes given by various standards aims to reduce the damage during earthquake or to avoid serious accidents. In case of a Nuclear Power Plant, all safety related structures, systems and components are designed to with stand a very low probability of earthquake. This work gives the details of seismic analysis by Time history method for a typical Steam Generator (SG) used in sodium cooled fast breeder reactor. Steam generator of a sodium cooled reactor have sodium in shell side and water in tube side unlike water cooled nuclear reactors, where tube side and shell side is water. The failure of tubes in sodium cooled fast reactor steam generator result in leaking of sodium to water side causing sodium water reaction which is explosive in nature. The large energy release in sodium water reaction will damage the complete structure of SG and connected equipments causing shutdown of the plant. Hence it is very important to ensure the safety of the SG at all operating conditions even during event of severe earthquake. In this work the structural integrity of SG is analysed for events of earthquakes viz. Operating Basis Earthquake (OBE) and Safe Shutdown Earthquake (SSE) by performing seismic analysis by Time history method. The stresses at various critical locations are estimated and checked against limits given by ASME Section –III NC rules

Effects of Al-Si coating and Zn coating on the hydrogen uptake and embrittlement of ultra-high strength press-hardened steel

Press-hardened steel (PHS), used for automotive safety-related structure parts, is sensitive to hydrogen embrittlement due to its martensitic microstructure. Hydrogen is introduced in PHS during the hot press forming (HPF) process, by an atmospheric corrosion process. In this study, the hydrogen embrittlement behavior of uncoated, aluminized, and galvanized PHSs was investigated. The Al-10%Si coating promoted the absorption of diffusible hydrogen at elevated temperature during the HPF while the reacted coating layer prevented the absorbed hydrogen from out-diffusing through the reacted coating surface layer at room temperature. Therefore, the aluminized PHS showed a greater sensitivity to both the hydrogen uptake and the resultant embrittlement, as compared to the uncoated and galvanized PHSs. Use of galvanized PHS for HPF application reduces the risk of hydrogen embrittlement, since the Zn coating effectively prevents the hydrogen uptake. The greater embrittlement resistance of the galvanized PHS is possibly due to the inhibition of the hydrogen generation reaction by the surface ZnO oxide layer and the low rate of hydrogen transport through the liquid Zn phase.

Structural assessment of radiation damage in light water power reactor concrete biological shield walls

An important topic for long-term operation of nuclear plants is aging of plant concrete structures. The containment building, biological shielding, and support concrete are examples of concrete structures that are of primary importance in the operation of a nuclear plant. These and other safety-related structures must be capable of maintaining structural capability for the operating life of the plant. Demonstration of the satisfactory condition of the concrete structures is required for safe operation, particularly when plant operation beyond 60 years is considered.The concrete biological shield (CBS) wall is particularly important because of its shielding and structural functions and that it is irradiated due to its proximity to the reactor pressure vessel. Demonstrating that the concrete remains structurally able to perform its intended function is vital in effective aging management.Neutron irradiation above certain thresholds can cause loss of tensile and compressive strength of the concrete and volumetric expansion. If the irradiation level exceeds the threshold, both loss of strength and volumetric expansion should be evaluated. Methods to reliably evaluate the effects of the loss of strength and volumetric expansion are required.

The multi-axial strength performance of composited structural B-C-W members subjected to shear forces

This paper presents a new method to compute the shear strength of composited structural B-C-W members. These B-C-W members, defined as concrete-filled steel box beams, columns and shear walls, consist of a slender rectangular steel plate box filled with concrete and inserted steel plates connecting the two long-side steel plates. These structural elements are intended to be used in structural members of super-tall buildings and nuclear safety-related structures. The concrete confined by the steel plate acts to be in a multi-axial stressed state: therefore, its shear strength was calculated on the basis of a concrete\'s failure criterion model. The shear strength of the steel plates on the long sides of the structural element was computed using the von Mises plastic strength theory without taking into account the buckling of the steel plate. The spacing and strength of the inserted plates to induce plate yielding before buckling was determined using elastic plate theory. Therefore, a predictive method to compute the shear strength of composited structural B-C-W members without considering the shear span ratio was obtained. A coefficient considering the influence of the shear span ratio was introduced into the formula to compute the anti-lateral bearing capacity of composited structural B-C-W members. Comparisons were made between the numerical results and the test results along with this method to predict the anti-lateral bearing capacity of concrete-filled steel box walls. Nonlinear static analysis of concrete-filled steel box walls was also conducted by using ABAQUS and the results agreed well with the experimental data.

In-structure response spectra considering nonlinearity of RCC structures: Experiments and analysis

Many devastating earthquakes exceeding the design basis levels have occurred in the past. Safety assessment of equipments and piping systems of industrial/nuclear safety related structures subjected to earthquakes beyond design basis levels necessitates consideration of the structural nonlinearity. The In-structure or floor response spectra (FRS) are the input required for qualification of these systems and hence it is essential to generate this input considering structural nonlinearity. This paper addresses the experimental and analytical work on generation of FRS for nonlinear Reinforced Cement Concrete (RCC) structure. Push over and shake table experiments are performed on three storeyed RCC framed structure. A simplified analytical approach is proposed to derive earthquake-induced In-structure response spectra for the structure deforming beyond elastic limits. In this approach, initially, the performance of the structure is evaluated using pushover analysis. Floor spectra of the structure are then obtained considering the degraded stiffness and increased damping evaluated at each performance level iteratively from pushover curve. The floor spectra of the structure generated by the proposed method are validated using experimentally simulated results from shake table tests.

Characterization of Radiation Fields for Assessing Concrete Degradation in Biological Shields of NPPs

Life extensions of nuclear power plants (NPPs) to 60 years of operation and the possibility of subsequent license renewal to 80 years have renewed interest in long-term material degradation in NPPs. Large irreplaceable sections of most nuclear generating stations are constructed from concrete, including safety-related structures such as biological shields and containment buildings; therefore, concrete degradation is being considered with particular focus on radiation-induced effects. Based on the projected neutron fluence values (E > 0.1 MeV) in the concrete biological shields of the US pressurized water reactor fleet and the currently available data on radiation effects on concrete, some decrease in mechanical properties of concrete cannot be ruled out during extended operation beyond 60 years. An expansion of the irradiated concrete database is desirable to ensure reliable risk assessment for extended operation of nuclear power plants.

Improvement of impact-resistance of a nuclear containment building using fiber reinforced concrete

Since the act of terrorism that occurred in the USA on September 11, 2001, the protection of nuclear power plants against large commercial aircraft crashes has been an emerging issue. Besides the verification of the safety of nuclear power plants in operation or in design, efficient methods for improving the impact-resistance of these structures have been investigated. Fiber reinforced concrete (FRC) has been generally accepted as an effective material for this purpose. In particular, FRC has been developed to improve the tensile behavior of concrete such as tensile strength, ductility and toughness. One of the main fields of application of FRC can be found in blast-protective or blast-resistant concrete structures. It is expected, therefore, that safety-related structures in a nuclear power plant can also be effectively protected from external blast, aircraft crash, etc. by applying FRC. In order to analytically verify the effect on structural behavior of applying FRC, the particular material properties of FRC should be incorporated into the material modeling of a structural analysis program. This study investigates the mathematical modeling of FRC, which represents various aspects of material behavior. Two numerical examples are provided to show the improved impact-resistance of a nuclear containment building that is expected when applying FRC in comparison with ordinary concrete. The analysis results show that the displacement decreases by 43–67% while the impact-resistance increases by 40–82%, depending on a fiber type.

Inelastic seismic behavior of post-installed anchors for nuclear safety related structures: Generation of experimental database

Post installed (PI) anchors are often employed for connections between concrete structure and components or systems in nuclear power plants (NPP) and related facilities. Standardized practices for nuclear related structures demand stringent criteria, which an anchor has to satisfy in order to qualify for use in NPP related structures. In NPP and related facilities, the structure–component interaction in the event of an earthquake depends on the inelastic behavior of the concrete structure, the component system and also the anchorage system that connects them.For analysis, anchorages are usually assumed to be rigid. Under seismic actions, however, it is known that anchors may undergo significant plastic displacement and strength degradation. Analysis of structure–component interaction under seismic loads calls for numerical models simulating inelastic behavior of anchorage systems. A testing program covering different seismic loading scenarios in a reasonably conservative manner is required to establish a basis for generating numerical models of anchorage systems. Currently there is a general lack of modeling techniques to consider the inelastic behavior of anchorages in structure–component interaction under seismic loads.In this work, in view of establishing a basis for development of numerical models simulating the inelastic behavior of anchors, seismic tests on two different undercut anchors qualified for their use in NPP related structures were carried out. The test program was primarily based on the DIBt-KKW-Leitfaden (2010) guidelines for testing of fastening for use in NPP. The testing program consisted of monotonic and cyclic tests in non-cracked and cracked concrete specimens with different crack widths. In addition, crack cycling tests at design tension load, and simultaneous tension and crack cycling tests (in-phase and out-of-phase) were also conducted. The present paper deals with the details of the tests performed and the results obtained. The development of numerical models simulating anchor inelastic behavior based on the results given in the present paper will be presented in another paper.