Response of reinforced concrete structures to macrocell corrosion of reinforcements. Part I: Before propagation of microcracks via an analytical approach

Assessment of the macrocell corrosion which deteriorates reinforced concrete (RC) structures have attracted the attention of many researchers during recent years. In this type of rebar corrosion, the reduction in cross-section of the rebar is significantly accelerated due to the large ratio of the cathode's area to the anode's area. In order to examine the problem, an analytical solution is proposed for prediction of the response of the RC structure from the time of steel depassivation to the stage just prior to the onset of microcrack propagation. To this end, a circular cylindrical RC member under axisymmetric macrocell corrosion of the reinforcement is considered. Both cases of the symmetric and asymmetric rebar corrosion along the length of the anode zone are studied. According to the experimentally observed data, corrosion products are modeled as a thin layer with a nonlinear stress–strain relation. The exact expressions of the elastic fields associated with the steel, and concrete media are obtained using Love's potential function. By imposing the boundary conditions, the resulting set of nonlinear equations are solved in each time step by Newton's method. The effects of the key parameters which have dominating role in the time of the onset of concrete cracking and maximum radial stress field of the concrete have been examined.     

Response of reinforced/prestressed concrete box structures to dynamically applied torsion

Inelastic dynamic response of reinforced/prestressed concrete box structures to torsion is investigated in this article. To perform this research, three items need to be critically studied. (1) The primary curve (torque-twist curve), (2) the torsional hysteretic rules and (3) a dynamic model. The inelastic dynamic response analysis of both reinforced and prestressed concrete box structures (such as: spent fuel pools, storage tanks, containments and building cores, etc.) under torsion can be completed by the proposed method. The research findings in this article can be applied to box structures in nuclear power plants.     

Local damage to reinforced concrete structures caused by impact of aircraft engine missiles Part 2. Evaluation of test results

Three sets of impact tests, small-, intermediate-, and full-scale tests, have been executed to determine local damage to reinforced concrete structures caused by the impact of aircraft engine missiles. The results of the test program showed that (1) the use of the similarity law is appropriate, (2) suitable empirical formulas exist for predicting the local damage caused by rigid missiles, (3) reduction factors may be used for evaluating the reduction in local damage due to the deformability of the engines, (4) the reinforcement ratio has no effect on local damage, and (5) the test results could be adequately predicted using nonlinear response analysis.     

Local damage to reinforced concrete structures caused by impact of aircraft engine missiles Part 1. Test program, method and results

Structural damage induced by an aircraft crashing into a reinforced concrete structure includes local damage caused by the deformable engines, and global damage caused by the entire aircraft. Local damage to the target may consist of spalling of concrete from its front face together with missile penetration into it, scabbing of concrete from its rear face, and perforation of missile through it. Until now, local damage to concrete structures has been mainly evaluated by rigid missile impact tests. Past research work regarding local damage caused by impact of deformable missiles has been limited. This paper presents the results of a series of impact tests of small-, intermediate-, and full-scale engine models into reinforced concrete panels. The purpose of the tests was to determine the local damage to a reinforced concrete structure caused by the impact of a deformable aircraft engine.     

Seismic proof test of a reinforced concrete containment vessel (RCCV): Part 3. Evaluation of seismic safety margin

In Japan, the Nuclear Power Engineering Corporation (NUPEC), sponsored by the Ministry of Economy, Trade and Industry (METI), had conducted a series of seismic reliability proving tests using full-scale or close to full-scale models to simulate an actual important equipment that is critical for seismic safety of nuclear power plants. The tests are intended to validate the seismic design and reliability with a sufficient margin even under destructive earthquakes. A series of tests was carried out on a reinforced concrete containment vessel (RCCV) for advanced boiling water reactor (ABWR) from 1992 to 1999. A large-scale high-performance shaking table at Tadotsu Engineering Laboratory, was used for this test. The test model and the results of pressure and leak tests are described in Part 1. Test procedures, input waves and the results of verification tests such as changes of stiffness, characteristic frequency and damping ratio, the failure of the model and the load–deformation relationship are described in Part 2. Part 3 reports the seismic design safety margin that was evaluated from the energy input during the failure test to a design basis earthquake. Part 4 will report simulation analysis results by a stick model with lumped masses.     

Seismic proof test of a reinforced concrete containment vessel (RCCV): Part 2: Results of shaking table tests

A 1/8-scale model was constructed of a reinforced concrete containment vessel (RCCV) used in the latest advanced boiling water reactors (ABWR). Shaking table tests were conducted on it with input motions corresponding to or exceeding a design earthquake assumed for a real Nuclear Power Plant.The objectives of the tests were to verify the structural integrity and the leak-proof functional soundness of the RCCV subjected to design earthquakes, and to determine the ultimate strength and seismic margin by an excitation that led to the model's collapse. The model, the test sequence and the pressure and leak test results were addressed in Part 1. The shaking table test method, the input motions and the test results, including the transition of the model's stiffness, natural frequencies and damping factors and the effects of vertical input motions and internal pressure on the model's characteristics and behavior, the load–deformation, the ultimate strength, the failure mode of the reinforced concrete portion and the liner plate are described here. The seismic safety margin that was evaluated by the energy input during the failure test to a design basis earthquake will be described in Part 3. The analytical results of simulation using the multi-lumped mass model will be described in Part 4.     

Seismic proof test of a reinforced concrete containment vessel (RCCV): Part 1: Test model and pressure test

In Japan, the Nuclear Power Engineering Corporation (NUPEC), sponsored by the Ministry of Economy, Trade and Industry (METI), has been conducting a series of seismic reliability proving tests using full-scale or close to full-scale models to simulate actual important equipment that is critical for seismic safety of nuclear power plants. The tests are intended to validate the seismic design and reliability with a sufficient margin even under destructive earthquakes. A series of tests was carried out on a reinforced concrete containment vessel (RCCV) for advanced boiling water reactor (ABWR) from 1992 to 1999. A large-scale high-performance shaking table at Tadotsu Engineering Laboratory, the largest in the world, was used for this test. Part 1 reports the test model and the results of pressure and leak tests. Part 2 describes test procedures, input waves and the results of verification tests such as changes of stiffness, characteristic frequency and damping ratio, the failure of the model and the load deflection. Part 3 shows the seismic safety margin that was evaluated from the energy input during the failure test to a design basis earthquake. Part 4 reports simulation analysis results by a stick model with lumped masses.     

Analysis of cleavage fracture potential of martensitic stainless steel fusion structures: Part II. Fracture analysis procedures for flawed fusion structures

The potential for brittle cleavage fracture is a major concern for martensitic stainless steels which are candidates for fusion reactor structural materials. This study attempts to identify for flawed fusion structures the pertinent fracture resistance or failure parameters and the relationships between these parameters and the basic materials properties which govern cleavage fracture. Several procedures for relating test data to failure prediction, including Charpy-V-notch transition temperature referencing and two-parameter interpolation procedures, are considered; and results are discussed with respect to possible research paths for martensitic stainless steel alloy development.     

Evaluation of precipitates used in strainer head loss testing: Part II. Precipitates by in situ aluminum alloy corrosion

Vertical loop head loss tests were performed with 6061 and 1100 aluminum (Al) alloy plates immersed in borated solution at pH = 9.3 at room temperature and 60 °C. The results suggest that the potential for corrosion of an Al alloy to result in increased head loss across a glass fiber bed may depend on its microstructure, i.e., the size distribution and number density of intermetallic particles that are present in Al matrix and FeSiAl ternary compounds, as well as its Al release rate. Per unit mass of Al removed from solution, the WCAP-16530 aluminum hydroxide (Al(OH)₃) surrogate was more effective in increasing head loss than the Al(OH)₃ precipitates formed in situ by corrosion of Al alloy. However, in choosing a representative amount of surrogate for plant specific testing, consideration should be given to the potential for additional head losses due to intermetallic particles and the apparent reduction in the effective solubility of Al(OH)₃ when intermetallic particles are present.     

Cermet fuel in a light water reactor: a possible way to improve safety. Part II. Testing the fuel

This paper summarizes what is done for the experimental testing of cermet fuel with various matrix materials. Low neutron absorption, high heat conductivity, good corrosion resistance in water, low chemical interaction with cladding (zirconium alloy) and UO₂ in normal and accident conditions, technological ability — are the requirements of the matrix material [1]. Suitability of the proposed solutions to the cermet fuel design with respect to these requirements was proven through a series of experiments simulating fuel operating and accidental conditions.     

Seismic proof test of a reinforced concrete containment vessel (RCCV): Part 2: Results of shaking table tests

A 1/8-scale model was constructed of a reinforced concrete containment vessel (RCCV) used in the latest advanced boiling water reactors (ABWR). Shaking table tests were conducted on it with input motions corresponding to or exceeding a design earthquake assumed for a real Nuclear Power Plant.The objectives of the tests were to verify the structural integrity and the leak-proof functional soundness of the RCCV subjected to design earthquakes, and to determine the ultimate strength and seismic margin by an excitation that led to the model's collapse. The model, the test sequence and the pressure and leak test results were addressed in Part 1. The shaking table test method, the input motions and the test results, including the transition of the model's stiffness, natural frequencies and damping factors and the effects of vertical input motions and internal pressure on the model's characteristics and behavior, the load-deformation, the ultimate strength, the failure mode of the reinforced concrete portion and the liner plate are described here. The seismic safety margin that was evaluated by the energy input during the failure test to a design basis earthquake will be described in Part 3. The analytical results of simulation using the multi-lumped mass model will be described in Part 4.     

Seismic proof test of a reinforced concrete containment vessel (RCCV): Part 1: Test model and pressure test

In Japan, the Nuclear Power Engineering Corporation (NUPEC), sponsored by the Ministry of Economy, Trade and Industry (METI), has been conducting a series of seismic reliability proving tests using full-scale or close to full-scale models to simulate actual important equipment that is critical for seismic safety of nuclear power plants. The tests are intended to validate the seismic design and reliability with a sufficient margin even under destructive earthquakes. A series of tests was carried out on a reinforced concrete containment vessel (RCCV) for advanced boiling water reactor (ABWR) from 1992 to 1999. A large-scale high-performance shaking table at Tadotsu Engineering Laboratory, the largest in the world, was used for this test. Part 1 reports the test model and the results of pressure and leak tests. Part 2 describes test procedures, input waves and the results of verification tests such as changes of stiffness, characteristic frequency and damping ratio, the failure of the model and the load deflection. Part 3 shows the seismic safety margin that was evaluated from the energy input during the failure test to a design basis earthquake. Part 4 reports simulation analysis results by a stick model with lumped masses.     

Benchmark Simulation of Turbulent Flow through a Staggered Tube Bundle to Support CFD as a Reactor Design Tool. Part II: URANS CFD Simulation

In Part II, we described the unsteady flow simulation and proposed a modification of a traditional turbulence flow model. Computational fluid dynamics (CFD) simulations of an isothermal, fully periodic flow across a tube bundle using unsteady Reynolds averaged Navier-Stokes (URANS) equations, with turbulence models such as the Reynolds stress model (RSM) were investigated at a Reynolds number of 1.8 × 10⁴, based on the tube diameter and inlet velocity. As noted in Part I, CFD simulation and experimental results were compared at five positions along (x; y) coordinates. The steady RANS simulation showed that four diverse turbulence models were efficient for predicting the Reynolds stresses, and generally, SRANS results were marginal to poor, using a consistent evaluation terminology. In the URANS simulation, we modeled the turbulent flow field in a manner similar to the approach used for large eddy simulation (LES). The time-dependent URANS results showed that the simulation reproduces the dynamic stability as characterized by transverse oscillatory flow structures in the near-wake region. In particular, the inclusion of terms accounting for the time scales associated with the production range and dissipation rate of turbulence generates unsteady statistics of the mean and fluctuation flow. In spite of this, the model implemented produces better agreement with a benchmark data set and is thus recommended.     

Effects of fuel type on the safety characteristics of a sodium cooled fast reactor. Part II: Simulation results

The objective of this investigation is to examine the impact of the fuel type on the inherent safety characteristics of Liquid Metal Fast Reactors (LMFRs). To perform this study, the responses to various transient conditions are examined for metallic, oxide and nitride cores of a baseline LMFR. GE-Hitachi’s Super Power Reactor Innovative Small Module (S-PRISM) was chosen as the baseline LMFR. In Part I of this paper, the background on S-PRISM’s metal and oxide cores are described and the redesign of a new nitride fueled S-PRISM core were introduced. Reactivity feedback and power profile data necessary for transient simulations with RELAP5-3D/ATHENA (RELAP5-3D, 2009) code are also presented and discussed. In this Part II of our paper, we present the results of accident simulations and a comparison between the metal, oxide and nitride cores based on their performance during the selected accident scenarios. Loss of Flow, Loss of Heat Sink, Loss of Power and inadvertent control rod withdrawal accidents were simulated for each core at beginning (BOC), middle (MOC) and end of a fuel cycle (EOC). The simulations were stopped at the initiation of melting of fuel or cladding. The results showed that in most of the transients the metal core came closer to its melting temperature while the strong reactivity feedbacks of the oxide and nitride cores limited their fuel temperature increases. Overall, the oxide and nitride cores had similar performance with respect to their inherent safety characteristics.     

A second order turbulence model based on a Reynolds stress approach for two-phase boiling flow. Part 1: Application to the ASU-annular channel case

High-thermal performance PWR (pressurized water reactor) spacer grids require both low pressure loss and high critical heat flux (CHF) properties. Numerical investigations on the effect of angles and position of mixing vanes and to understand in more details the main physical phenomena (wall boiling, entrainment of bubbles in the wakes, recondensation) are required.In the field of fuel assembly analysis or design by means of CFD codes, the overwhelming majority of the studies are carried out using two-equation eddy viscosity models (EVM), especially the standard K-? model, while the use of Reynolds Stress Transport Models (RSTM) remains exceptional.But extensive testing and application over the past three decades have revealed a number of shortcomings and deficiencies in eddy viscosity models. In fact, the K-? model is totally blind to rotation effects and the swirling flows can be regarded as a special case of fluid rotation. This aspect is crucial for the simulation of a hot channel in a fuel assembly. In fact, the mixing vanes of the spacer grids generate a swirl in the coolant water, to enhance the heat transfer from the rods to the coolant in the hot channels and to limit boiling.First, we started to evaluate computational fluid dynamics results against the AGATE-mixing experiment: single-phase liquid water tests, with Laser-Doppler liquid velocity measurements upstream and downstream of mixing blades. The comparison of computed and experimental azimuthal (circular component in a horizontal plane) liquid velocity downstream of a mixing vane for the AGATE-mixing test shows that the rotating flow is qualitatively well reproduced by CFD calculations but azimuthal liquid velocity is underestimated with the K-? model.Before comparing performance of EVM and RSTM models on fuel assembly geometry, we performed calculations with a simpler geometry, the ASU-annular channel case. A wall function model dedicated to boiling flows is also proposed.     

CFD analysis of a regular sector of the ITER vacuum vessel. Part II: Thermal-hydraulic effects of the nuclear heat load

The 3D Computational Fluid Dynamic (CFD) steady state analysis of the regular sector #5 of the ITER vacuum vessel (VV) is presented in these two companion papers using the commercial software ANSYS-FLUENT®. The pure hydraulic analysis, concentrating on flow field and pressure drop, is presented in Part I. This Part II focuses on the thermal-hydraulic analysis of the effects of the nuclear heat load. Being the VV classified as safety important component, an accurate thermal-hydraulic analysis is mandatory to assess the capability of the water coolant to adequately remove the nuclear heat load on the VV. Based on the recent re-evaluation of the nuclear heat load, the steady state conjugate heat transfer problem is solved in both the solid and fluid domains. Hot spots turn out to be located on the surface of the inter-modular keys and blanket support housings, with the computed peak temperature in the sector reaching ～290 °C. The computed temperature of the wetted surfaces is well below the coolant saturation temperature and the temperature increase of the water coolant at the outlet of the sector is of only a few °C. In the high nuclear heat load regions the computed heat transfer coefficient typically stays above the 500 W/m² K target.     

Vibrations of probe used for the defect detection of helical heating tubes in a fast breeder reactor: Part 2. Vibration analysis and numerical simulations

A detection of a defect of a helical heating tube installed in the fast breeder reactor “Monju” in Japan is done by a feeding of an eddy current testing (ECT) probe with magnetic sensor into the tube. An undesirable vibration of the ECT probe always happens under a certain condition and makes the inspection difficult. Several characteristics of the vibration have been made clear by some experiments using a mock-up, but the essential factor of the vibration is still unclear. In this paper, a numerical simulation of the vibration is implemented on the assumption that the vibration is caused by Coulomb friction. An analytical model, which is obtained as a lumped mass model, is a large-scale non-linear vibration system and many computational costs are ordinarily required to carry out the simulations. The Transfer Influence Coefficient Method is applied so that the simulation is efficiently carried out. The results of simulation qualitatively agree well with the experimental results. It confirms the validity of the assumption that the vibration is caused by Coulomb friction.     

A one dimensional model for the prediction of stratification in horizontal pipes subjected to fluid temperature transient at inlet: Part II. Thermal model

Various thermal interactions that are of interest during stratification in horizontal pipes are analyzed in this paper. The interactions studied are those that occur between a slow and a fast moving stream or between the fluid and the pipe wall. A lumped capacity approach in the radial direction is used to model the fluid-wall and fluid-fluid interactions. However the internal resistance of the stationary layer is approximately accounted for in the model. The application of the model to sodium flowing in a 15.2 cm ID stainless steel pipe shows that fluid-fluid interaction can reduce the magnitude of maximum stratification by about 18% as compared to the case in which interaction is not considered or is of little consequence (e.g. for water). This implies that stratification predictions based on the hydrodynamical model alone tend to be conservative. It is also shown that the magnitude of the stratification based on the pipe wall temperature can be higher than that predicted from the hydrodynamic model.     

The effective convectivity model for simulation of melt pool heat transfer in a light water reactor pressure vessel lower head. Part II: Model assessment and application

The paper reports detailed assessments and representative application of the effective convectivity model (ECM) developed and described in the companion paper (Tran and Dinh, submitted for publication). The ECM capability to accurately predict energy splitting and heat flux profiles in volumetrically heated liquid pools of different geometries over a range of conditions related to accident progression is examined and benchmarked against both experimental data and CFD results. Augmented with models for phase changes in binary mixture, the resulting PECM (phase-change ECM) is validated against a non-eutectic heat transfer experiment. The PECM tool is then applied to predict thermal loads imposed on the reactor vessel wall and Control Rod Guide Tubes (CRGTs) during core debris heatup and melting in the BWR lower plenum. The reactor-scale simulations demonstrate the PECM's high computational performance, particularly needed to analyze processes during long transients of severe accidents. The analysis provides additional arguments to support an outstanding potential of using the CRGT cooling as a severe accident management measure to delay the vessel failure and increase the likelihood of in-vessel core melt retention in the BWR.