Science and Technology of Nuclear Installations

https://www.hindawi.com/journals/stni/

List of Papers (Total 954)

The Trimming-Based Design Method for PWR Coolant Flow Distribution Device

In order to resolve the situations of nonuniform coolant flow distribution and insufficient vortex suppression, the existing Pressurized Water Reactor (PWR) usually adopts complex coolant mixing structures. However, those structures will greatly increase the complexity and maintenance cost of the system. To solve this problem, a trimming-based design method is proposed in this...

Assessment of the Radioactive Contamination of the RBMK-1500 Reactor’s Steam Pipelines and High Pressure Rings

Data analyses of radioactive contamination of the RBMK-1500 reactor’s steam pipelines (SP) and components of high pressure rings (HPR) are presented in this paper. Also, modelled results of the SP-HPR system are compared to the results of other RBMK-1500 systems at Ignalina NPP Unit 1. Characteristics of SP-HPR components, thermal-hydraulic conditions of the coolant, and system...

Validation of Modified ART Mod 2 Code through Comparison with Aerosol Deposition of Cesium Compound in Phébus FPT3 Containment Vessel

Evaluation of aerosol deposition in the containment vessel is an important step for the assessment of radioactive material release to the environment. ART Mod 2 is a calculation code that is used for evaluation of aerosol deposition in the containment vessel. The authors modified aerosol deposition models of ART Mod 2, namely, gravitational settling model, Brownian diffusion...

Validation of the COTENP Code: A Steady-State Thermal-Hydraulic Analysis Code for Nuclear Reactors with Plate Type Fuel Assemblies

This article presents the validation of the Code for Thermal-hydraulic Evaluation of Nuclear Reactors with Plate Type Fuels (COTENP), a subchannel code which performs steady-state thermal-hydraulic analysis of nuclear reactors with plate type fuel assemblies operating with the coolant at low pressure levels. The code is suitable for design analysis of research, test, and...

Validation of the COTENP Code: A Steady-State Thermal-Hydraulic Analysis Code for Nuclear Reactors with Plate Type Fuel Assemblies

This article presents the validation of the Code for Thermal-hydraulic Evaluation of Nuclear Reactors with Plate Type Fuels (COTENP), a subchannel code which performs steady-state thermal-hydraulic analysis of nuclear reactors with plate type fuel assemblies operating with the coolant at low pressure levels. The code is suitable for design analysis of research, test, and...

Effects of Diameters on Countercurrent Flow Limitation at a Square Top End in Vertical Pipes

The falling liquid flow rate under flooding conditions is limited at a square top end of a vertical pipe in the pressurizer surge line with the diameter of about 300 mm that consists of a vertical pipe, a vertical elbow, and a slightly inclined pipe with elbows. In this study, therefore, we evaluated effects of diameters on countercurrent flow limitation (CCFL) at the square top...

Innovative Design of Replacement Device for Vulnerable Parts in the Nuclear Radiation Environment

Vigorously developing nuclear power is the main development direction of current renewable energy. In the nuclear environment, in order to avoid nuclear radiation damage to maintenance personnel and improve the efficiency of nuclear reaction, it is necessary and urgent to realize automatic replacement of vulnerable parts in the electron gun. As the key equipment for the...

A Novel Nonlinear BWR Stability Indicator Based on the Sample Entropy

BWRs are thus far the simplest energy systems to transform fission energy into electrical power. However, there are still many aspects in their operation that, under certain conditions, may induce BWR unstable behavior. The default indicator to study BWR unstable behavior is the Decay Ratio (DR). However, due to the fact that BWRs show very complex responses under instability and...

Benefits of Seismic Isolation for Nuclear Structures Subjected to Severe Earthquake

The Fukushima accident has reiterated that the seismic safety is a clear necessity in the design of nuclear power plants. To overcome the weaknesses of the plant design, appropriate measures or interventions have thus to be put in place to improve the nuclear safety. In this study, seismic isolation, widely adopted for conventional constructions, is considered as retrofit measure...

Uniformity Assessment of TRISO Fuel Particle Distribution in Spherical HTGR Fuel Element Using Voronoi Tessellation and Delaunay Triangulation

Nonuniform distribution of tri-structural-isotropic (TRISO) fuel particles in a spherical fuel element (SFE) may increase the failure probability of the SFE in the high-temperature gas-cooled reactor, leading to the release of fission products. To evaluate the uniformity of the TRISO particles nondestructively, 3-dimensional cone-beam computed tomography is used to image the SFE...

Effect of Subcooling on Pool Boiling of Water from Sintered Copper Microporous Coating at Different Orientations

The subcooling effect on pool boiling heat transfer using a copper microporous coating was experimentally studied in water for subcoolings of 10 K, 20 K, and 30 K at atmospheric pressure and compared to that of a plain copper surface. A high-temperature thermally conductive microporous coating (HTCMC) was made by sintering copper powder with an average particle size of 67 μm onto...

A Conceptual Approach to Eliminate Bypass Release of Fission Products by In-Containment Relief Valve under SGTR Accident

During a hypothesized severe accident, a containment building is designed to act as a final barrier to prevent release of fission products to the environment in nuclear power plants. However, in a bypass scenario of steam generator tube rupture (SGTR), radioactive nuclides can be released to environment even if the containment is not ruptured. Thus, thorough mitigation strategies...

Moisture Transfer Model and Simulation for Dehumidification of HTGR Core

A large number of carbon materials are adopted in high-temperature gas-cooled reactor (HTGR). These carbon materials mainly include graphite IG-110 and boron-containing carbon material (BC), both of which are typical porous materials and normally absorb moisture. In order to inhibit the chemical corrosion reaction between core internals materials and moisture, the core needs to...

Effect of Molten Corium Behavior Uncertainty on the Severe Accident Progress

Uncertainty of a severe accident code output needs to be handled reliably considering its use in safety regulation of nuclear industry. In particular, severe accident codes are utilized for probabilistic safety assessment (PSA), where the uncertainty of severe accident progress should be considered carefully due to its influence on human reliability analysis. Therefore, in this...

Theoretical Solutions for Dynamic Characteristics of Liquid Annular Seals with Herringbone Grooves on the Stator Based on Bulk-Flow Theory

Liquid annular seals are primarily used to control the leakage in high-speed turbomachinery, especially in nuclear and petrochemical pumps. In this paper, a theoretical analysis method for dynamic characteristics of liquid seals with herringbone grooves on the stator is proposed based on bulk-flow theory. Steady-state velocities and leakage rates within the upstream and...

Modelling of Severe Accident and In-Vessel Melt Retention Possibilities in BWR Type Reactor

One of the severe accident management strategies for nuclear reactors is the melted corium retention inside the reactor pressure vessel. The work presented in this article investigates the application of in-vessel retention (IVR) severe accident management strategy in a BWR reactor. The investigations were performed assuming a scenario with the large break LOCA without injection...

Modelling of Severe Accident and In-Vessel Melt Retention Possibilities in BWR Type Reactor

One of the severe accident management strategies for nuclear reactors is the melted corium retention inside the reactor pressure vessel. The work presented in this article investigates the application of in-vessel retention (IVR) severe accident management strategy in a BWR reactor. The investigations were performed assuming a scenario with the large break LOCA without injection...

Effect of Molten Corium Behavior Uncertainty on the Severe Accident Progress

Uncertainty of a severe accident code output needs to be handled reliably considering its use in safety regulation of nuclear industry. In particular, severe accident codes are utilized for probabilistic safety assessment (PSA), where the uncertainty of severe accident progress should be considered carefully due to its influence on human reliability analysis. Therefore, in this...

Theoretical Solutions for Dynamic Characteristics of Liquid Annular Seals with Herringbone Grooves on the Stator Based on Bulk-Flow Theory

Liquid annular seals are primarily used to control the leakage in high-speed turbomachinery, especially in nuclear and petrochemical pumps. In this paper, a theoretical analysis method for dynamic characteristics of liquid seals with herringbone grooves on the stator is proposed based on bulk-flow theory. Steady-state velocities and leakage rates within the upstream and...

Analysis of Spatial Discretization Error Estimators Implemented in ARES Transport Code for Equations

The discrete ordinates method (SN) is one of the mainstream methods for neutral particle transport calculations. Assessing the quality of the numerical solution and controlling the discrete error are essential parts of large-scale high-fidelity simulations of nuclear systems. Three error estimators, a two-mesh estimator, a residual-based estimator, and a dual-weighted residual...

Analysis of Spatial Discretization Error Estimators Implemented in ARES Transport Code for Equations

The discrete ordinates method (SN) is one of the mainstream methods for neutral particle transport calculations. Assessing the quality of the numerical solution and controlling the discrete error are essential parts of large-scale high-fidelity simulations of nuclear systems. Three error estimators, a two-mesh estimator, a residual-based estimator, and a dual-weighted residual...

Analysis of Loss of Essential Power System Reported in Nuclear Power Plants

The essential power supply system is important for the nuclear safety and accident mitigation of the currently operating nuclear power plants. This system provides electrical power to the essential instrumentation and control systems of the nuclear power plant when all alternate current power sources are lost. This event is known as station blackout (SBO) event. Operational...

Analysis of Steam Explosion under Conditions of Partially Flooded Cavity and Submerged Reactor Vessel

A steam explosion in a reactor cavity makes a mechanical load of the pressure pulse, which can result in a failure of the containment isolation. To prove the integrity of the containment during the ex-vessel steam explosion, the effects of water conditions on a steam explosion have to be identified, and the impulse of a steam explosion has to be exactly assessed. In this study...

Development of a Coupled Code for Steady-State Analysis of the Graphite-Moderated Channel Type Molten Salt Reactor

The molten salt reactor (MSR) is one of the six advanced reactor concepts selected by Generation IV International Forum (GIF) because of its inherent safety and the promising capabilities of TRU transmutation and Th-U breeding. In this study, a three-dimensional thermal-hydraulic model (3DTH) is developed for evaluating the steady-state performance of the graphite-moderated...

Development of a Coupled Code for Steady-State Analysis of the Graphite-Moderated Channel Type Molten Salt Reactor

The molten salt reactor (MSR) is one of the six advanced reactor concepts selected by Generation IV International Forum (GIF) because of its inherent safety and the promising capabilities of TRU transmutation and Th-U breeding. In this study, a three-dimensional thermal-hydraulic model (3DTH) is developed for evaluating the steady-state performance of the graphite-moderated...