Aircraft impact analysis is needed for a safety assessment of nuclear power plants. One of the contents which should be analysed for aircraft impact is physical damage of a concrete building and this can be estimated by a numerical simulation. In order to conduct aircraft impact analysis, simulation model which validated by some experimental data needs to be established. In 1990s, impact test using actual F4 Phantom fighter was conducted at Sandia national laboratory in U.S. and a lot of important experimental data were measured. In this paper, the numerical simulation results for this F4 Phantom impact test are introduced. The relationship between the thickness of the shell of the F4 Phantom simulation model and the deceleration of this model is indicated and the differences of the deceleration between simulation and test results are discussed. In addition, the relationship between fracture strain of the shell of the F4 Phantom simulation model and the destruction mode of this model in simulation is indicated and the differences between the destruction mode of the F4 Phantom between simulation and test results are discussed. In order to evaluate the physical damage area after the aircraft impact, it is necessary to estimate the aircraft velocity after it perforates the outer concrete wall and to calculate the decrease of the kinematic energy of the aircraft by this perforation. In this paper, several aircraft impact simulations with different concrete wall thickness are conducted and the reduction in kinematic energies of an aircraft by a perforation is estimated. Using these simulation results, the necessary numbers of concrete walls until the impacting aircraft stops is discussed.

The paper describes the activities of conceptual design of tools and procedures and the virtual simulation of the Remote Handling (RH) tasks provided in the maintenance of the systems present in the Access Cell (AC) of DONES (DEMO Oriented Neutron Source) facility. In particular, the RH maintenance of the Target Assembly (TA) is critical because of its position in the most severe region of neutron irradiation, the Test Cell (TC), where the material specimen are tested to understand the degradation of the materials properties throughout the reactor operational life. The main RH maintenance activity includes the replacement of the entire TA and the cleaning of the surfaces of connection in the TC. The cleaning operation is fundamental because it allows the removal of any lithium solid deposition from the surfaces: any further deposition on the surfaces could compromise the sealing of the TA. The RH is based on the idea of a reconfigurable modular chain of devices connected to the Access Cell Mast Crane (ACMC) located in the AC. To increase the modularity and to reduce the costs of the Remote Handling System (RHS), a telescopic boom is used equipped with a Gripper Change System (GCS) that allows the use of different end effectors. To perform the tasks, a Parallel Kinematic Manipulator (PKM) and a Robotic Arm (RA) are proposed, allowing the tools to move with more degree of freedom in the AC space. The modeling of the devices and the 3D kinematic simulations maintenance operations tasks were simulated and tested in virtual reality environment, aimed at developing and validating the implemented maintenance procedures, in collaboration with the IDEAinVR Laboratory of CREATE/University of Naples Federico II, and the research center at ENEA Brasimone, Italy.

Analysis of engineering approach to the operational life forecasting for constructional elements with respect to the low-cycle fatigue is carried out. Applicability limits for a hypothesis on existence of generalized cyclic-deforming diagram in case of complex low-cycle loading (deforming) are shown. It is determined, that under condition of plane-stress state and piecewise-broken trajectories of cycle loading with stresses and deformation checking the cyclic deforming diagram is united in limits of deformations, which are not exceeded 10 values of deformation corresponding material yield point. Generalized kinematic equation of material damageability is described. The method of damageability parameter utilization for increasing of accuracy calculation of structural elements low-cycle fatigue by using the effective coefficients of stresses and deformations taking into account the damageability parameter is given.

Natural convection has been an area of intense research since it was discovered over 100 years ago. However, a phenomenological explanation for the onset of natural convection is still not available. The role of surface roughness in various thermal hydraulic phenomena is widely accepted; however, the scale of roughness is missing from Rayleigh number (Ra) which is commonly used to predict the onset of natural convection. Using the mass-spring analogy, an analytical derivation is presented to accurately characterize the thermal system and the interplay of gravity, viscosity and scale of roughness responsible for the onset of natural convection. The necessary conditions for the onset of natural convection are given in terms of a new dimensionless number β g Δ T ε H 2 α 2 which captures the effect of surface roughness. Using the mass-springs analogy, the natural frequency of the convection system is shown to depend on thermal and kinematic property of the fluid and the scale of roughness. These results are applicable to many natural system and engineering designs.

The classic four-equation drift-flux model treats the two-phase flow as a mixture to formulate mass, momentum, and energy balance equations. The dispersed phase is modeled in the mass balance equation to describe the mass transfer between the two phases. However, this model implies an assumption of the thermal equilibrium in the mixture energy balance equation. Therefore, a five-equation drift-flux model and its constitutive equations have been developed to relax the thermal equilibrium assumption in the classic four-equation drift-flux model when it is applied to thermal non-equilibrium phenomena in Light Water Reactors (LWRs). The additional energy balance equation of the dispersed phase has been introduced to address thermal non-equilibrium phenomena in LWRs in the five-equation drift-flux model. In addition, the newly added energy equation for the dispersed phase does not change the basic kinematic parameters and uses the mixture velocity instead of the dispersed phase velocity. The hyperbolicity of the drift-flux model system equations is investigated by examining if the partial differential equations have only real eigenvalues. The results show that the five-equation drift-flux model is a well-posed hyperbolic model, which can be solved by both advanced solvers using the finite element method, such as MOOSE Framework, and the traditional solvers using finite difference method, such as RELAP5 code. In developing the thermal non-equilibrium five-equation drift-flux model, it is necessary to select, improve, or develop constitutive models as closure relations for variables used in the field equations. In our current study, a literature survey has been conducted to review appropriate constitutive models and correlations for the dispersed phase mass generation rate, interfacial energy transfer, distribution parameter, and drift velocity in the five-equation drift-flux model.

Control rod hydraulic drive system (CRHDS), which is invented by INET, Tsinghua University, is a new type of internal control rod drive technology. Control rod hydraulic deceleration device (CRHDD), which consists of the plug, the hydraulic deceleration cylinder, the hydraulic buffer, etc., is one of the main components of the CRHDS. The plug is connected with the top of the control rod driving shaft and moves along with the control rod inside the hydraulic deceleration cylinder. The CRHDD performs the rod dropping deceleration function through the interworking of the plug and the deceleration cylinder which is filled with water, and reduces the rod dropping peak acceleration and the impact force acting upon the control rod to prevent the control rod cruciform blade from being deformed or damaged. The working mechanism of the CRHDD is presented and analyzed. The rod dropping performance of the CRHDS was tested experimentally under room temperature. The theoretical model of the control rod dropping process, which is composed of the three dimensional flow field analysis and flow resistance calculation of the hydraulic deceleration cylinder, the kinematics and dynamics model of the control rod, is built whose results are compared and validated by the CRHDS scram test results under room temperature. Then the model takes into account of the influence of the fuel assembly box on the control rod scram process under high temperature working conditions, and is used to analyze the influence of the key parameters, including the helical spring stiffness inside the deceleration cylinder and the working temperature on the CRHDD working performance. The research results can give guidance for the design and optimization of the CRHDD.

During the normal operation of a pebble bed gas-cooled reactor (PBR), the fuel pebbles undergo a multi-circulation on the basis of online burnup assay. In our last ICONE paper, we proposed a model to describe the relationship between online burnup assay and economy and safety of PBR. It was concluded that improvements on burnup assay accuracy could reduce fuel cost as well as the possibility that excessive burnup of fuel pebble results in unexpected radioactive discharge. However further work was expected on the burnup distribution of pebbles in and out of the core to precisely quantify the relationship. In this paper, the methodology to construct the burnup distribution of fuel pebbles in and out of the core is proposed. Firstly a model for pebble flow circulation is developed to provide a basic simulation framework. Then the irradiation history of fuel pebble could be tracked by combining pebble flow model and burn up calculation. The representative kinematic model and discrete element method (DEM) are introduced to numerically simulate the profiles of pebble flow. The classical batch-tracking methods as well as our newly-introduced DEM-tracking method are presented to perform the time-dependent analysis of pebble burnup. Overall with the burnup data obtained after going through multiple cycles, the burnup distribution of fuel pebbles in and out of the core could be reconstructed through the statistics result according to the pebble circulation model. Finally the quantification of the relationship between the pebble burnup assay and the economy and safety of the PBR would be more precise, thus providing implications on proposing reasonable requirements for accuracy of online burnup assay.

In recent years, the nuclear industry and the Nuclear Regulatory Commission (NRC) have made a tremendous effort to assess the safety of nuclear power plants as advances in seismology have led to the perception that the potential earthquake hazard in the United States may be higher than originally assumed. The Seismic Probabilistic Risk Assessment (S-PRA) is a systematic approach used in the nuclear power plants in the U.S. to realistically quantify the seismic risk as by performing an S-PRA, the dominant contributors to seismic risk and core damage can be identified. The assessment of component fragility is a crucial task in the S-PRA and because of the conservatism in the design process imposed by stringent codes and regulations for safety related structures, structures and safety related items are capable of withstanding earthquakes larger than the Safe Shutdown Earthquake (SSE). One major aspect of conservatism in the design is neglecting the effect of Soil-Structure-Interaction (SSI), from which conservative estimates of In-Structure Response Spectra (ISRS) are calculated resulting in conservative seismic demands for plant equipment. In this paper, a typical Reactor Building is chosen for a case study by discretizing the building into a lumped mass stick model (LMSM) taking into account model eccentricities and concrete cracking for higher demand. The model is first analyzed for a fixed base condition using the free field ground motion imposed at the foundation level from which ISRS are calculated at different elevations. Computations taking into account the SSI effects are then performed using the subtraction method accounting for inertial interactions by using frequency dependent foundation impedance functions depicting the flexibility of the foundation as well as the damping associated with foundation-soil interaction. Kinematic interactions are also taken into account in the SSI analysis by using frequency dependent transfer functions relating the free-field motion to the motion that would occur at the foundation level as the presence of foundation elements in soil causes foundation motions to deviate from free-field motions as a result of ground motion incoherence and foundation embedment. Comparing the results of the seismic response analyses, the effects of the SSI is quantified on the overall seismic risk and the SSI margin is calculated. A family of realistic seismic fragility curves of the structure are then developed using common industry safety factors (capacity, ductility, response, and strength factors), and also variability estimates for randomness and uncertainty. Realistic fragility estimates for structures directly enhances the component fragilities from which enhanced values of Core Damage Frequency (CDF) and Large Energy Release Frequency (LERF) are quantified as a final S-PRA deliverable.

Supercritical Water Reactor (SCWR) proposes higher thermal efficiency and simpler plant design compared to modern Boiling Water Reactors. High pressure, temperature and power requirement in SCWR, however, escalates the cost of an experimental facility significantly. Present work, therefore, focuses on designing downscaled test facilities for stability analysis of SCWR. The facilities are conceptualized to model the European reference design of SCWR under both forced and natural circulation condition. R-134a is identified as the scaling fluids through fluid-to-fluid modeling, along with two others from literature. Similarity variables are obtained following two different approaches, starting from fundamental conservation equations. Dimensional and non-dimensional representations of important geometric, kinematic and dynamic parameters are evaluated and compared. Comparisons between two different approaches, as well as between forced and natural circulation have been presented for each scaling fluid.

Power plants, including nuclear power plants regularly employ tanks whose contents need to be kept isolated from atmospheric conditions. One way to satisfy these requirements is to provide a liner for the tank which completely fits the interior shape of the tank but floats on top of the tank contents when the tank contains fluid. As the volume in the tank changes, the liner or diaphragm accommodates the changes in volume by sliding along the tank walls. To allow free movement of the diaphragm, management of the gas volume above the fluid and behind the diaphragm is of prime importance. The work described in this paper elaborates on the conditions required to prevent the tank diaphragm from becoming damaged. To develop potential failure modes, the kinematics of the diaphragm and the interaction with the gas volume between the diaphragm and the tank fluid are considered in detail. The developed model is applied to the case of a condensate storage tank at Comanche Peak Nuclear Power Plant (CPNPP). Two physical scale models of the tank were constructed and tested to validate the model and allow the safe operation principles to be quantified for use in the operation of the condensate storage tank at CPNPP. The work allowed CPNPP to design appropriate periodic checks and maintenance activities to ensure the diaphragm will not be damaged due to tank volume changes while still ensuring the required water chemistry criteria for the tank contents can be met.

The moving direction of double seal door (DSD) of ITER remote handling transfer cask and the force of hydraulic pole will change significantly at the guide rail inflexion position (GRIP) which is a mutant site, so it is very possible to make the structure damage or the system failure at the GRIP. In this paper, the kinematics simulation and analysis of DSD were done based on special constitution restriction and working process by software ADAMS. The stress distribution of guide rail and hydraulic pole were obtained by the above simulation, at the same time the optimal GRIP was confirmed according to the force analysis result. The above-mentioned analysis process and results not only provide technical data for the optimization design and the prototype manufacture of DSD, but also provide the examples and references of kinematics analysis for other important components of ITER.

Pyrometallurgical reprocessing technology is currently being focused in many countries for closing actinide fuel cycle because of its favorable economic potential and an intrinsic proliferation-resistant feature due to the inherent difficulty of extracting weapons-usable plutonium. The feasibility of pyrometallurgical reprocessing has been demonstrated through many laboratory scale experiments. Hence the development of the engineering technology necessary for pyrometallurgical reprocessing is a key issue for industrial realization. The development of high-temperature transport technologies for molten salt and liquid cadmium is crucial for pyrometallurgical processing; however, there have been very few transport studies on high-temperature fluids. In this study, a salt transport test rig and a metal transport test rig were installed in an argon glove box with the aim of developing technologies for transporting molten salt and liquid cadmium at approximately 773 K. It was demonstrated that; using a centrifugal pump, molten salt at 773 K could be transported at a controlled rate from 4 to 8 dm 3 /min against a 1 m head. The transport behavior of the molten salt was found to be similar to that of water, and could be predicted from their similarity of kinematic viscosity. On the other hand, the transportation of liquid cadmium at approximately 700 K could be controlled at a rate of 0.5 to 1.6 dm 3 /min against a 1.6 m head using the centrifugal pump.

In nuclear power plants, slender tubular components are subjected to vibrations in a PHTW environment. As a result, the two contacting surfaces, tubes and their guides undergo impact at low contact pressures [1]. The components are usually made of stainless steel and it was found that the influence of the PHTW, combined with other actions (such as corrosion, erosion, squeeze film effect, third body effect and cavitation) leads to a particular wear of the material [2] [3]. Therefore, this paper aims to show that the colloidal oxides, formed on the steel surfaces in PHTW, play a principal role in the wear of the surfaces. Actually, due to the specific kinematic conditions of the contact, the flow of compacted oxides abrades the surfaces.

The nuclear power plant piping systems are often subjected to the cyclic loading conditions due to transients, seismic or other unexpected events. During these events, the probable mode of failure for piping component is fatigue-ratcheting. Earlier the design of piping subjected to seismic excitation was based on the principle of static collapse. Later on, it was postulated that the cause of failure of piping components is fatigue ratcheting and not plastic collapse. The 1995 ASME Boiler & Pressure Vessel code, Section-III, has incorporated the reverse dynamic loading and ratcheting into the code. An Analytical study is carried out to investigate the behavior of the pressurized piping system under large seismic loading. The analysis is performed using equivalent inertial forces. Chaboche nonlinear kinematic hardening model is used for ratcheting simulation. The capability of the model to simulate the ratcheting response of the piping system is of particular interest. Comparison of analysis results against test results are presented in the paper.

In the Safety analysis of Fast Breeder Reactor, assessment of Molten Fuel Coolant Interaction (MFCI) assumes importance for two aspects, namely the characterization of the debris and severity of pressure pulses generation. An attempt has been made to investigate the debris generation characteristics with molten Woods Metal (Alloy of Bi 50% Pb 25% Sn 12.5% & Cd 12.5% & melting point of 346 K) - Water simulant system. Liquid Woods metal and liquid Uranium dioxide physical properties (Density, Surface tension & Kinematic viscosity) are similar. Experimental studies were conducted for various melt temperatures covering non-boiling, convective boiling and film boiling regimes of water, to assess the debris generation resulting from both hydrodynamic and thermal interaction. Woods metal was heated to the desired temperature and poured through a hot funnel having a nozzle of 8 mm release diameter into a water column of height up to 140 cm. Experiments were repeated for different coolant temperature and melt inventory up to 5 kg. The melt entry velocity was determined from video recordings. The debris is analyzed on the basis of interface temperature, Rayleigh-Taylor and Kelvin-Helmholtz instabilities. It is observed that Kelvin-Helmholtz instability is the dominant fragmentation phenomena. Contribution due to coolant boiling resulted in more debris generation in the size less than 4 mm.

A specific tribometer equipped for electrochemical measurements in pressurized high temperature water (PHTW) is being used to study the effect of the water chemistry on the wear rate of a stainless steel. First results indicate that a slightly acidic pH increases drastically the wear rate compared to slightly alkaline pH. The influence of hydrogen partial pressure and contact kinematics is also described.