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.     

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 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.     

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.     

Results of large-scale test of discontinuity region in a prestressed concrete containment

The test described in this paper is part of an Electric Power Research Institute (EPRI) program (Research Program RP2172-2) to provide a test-verified analytical method of estimating capacities of concrete reactor containment buildings under internal overpressurization from postulated degraded core accidents.Phase 2 of the EPRI program, on which this paper is based, includes tests of five large-scale specimens with steel liner plates. The specimens represent structural elements of prestressed concrete containment buildings. Four full-scale square wall element specimens and one specimen representing the wall/basemat junction region were tested. This paper describes results of the wall/basemat junction region test.Results of this experimental work indicate that highly localized strains in the steel liner plate caused by internal overpressurization or other accident conditions can result in liner tearing and subsequent containment leakage. It appears that this liner tearing occurs in a controller manner. Extrapolating from these test results, leakage and depressurization is more likely to occur than global failure.     

Response of reinforced concrete structures to macrocell corrosion of reinforcements. Part II: After propagation of microcracks via a numerical approach

Investigation of response of reinforced concrete (RC) structures due to axisymmetric macrocell corrosion of rebars is of concern after propagation of microcracks within the concrete medium. The geometry, boundary and interfaces conditions of the present problem are identical to those stated in part I. As seen in the companion paper, the exact solution to the boundary value problem corresponding to the uncracked steel–rust–concrete composite was possible. After appearance of the microcracks the concrete behavior becomes nonlinear anisotropic with post-cracking softening, and the associated problem is analytically intractable. Therefore, it is proposed to employ a novel meshless method, namely gradient reproducing kernel particle (GRKPM), in the cylindrical coordinates. The analytical and numerical solutions pertinent to the uncracked concrete are in good agreement. Subsequently, the effects of the parameters associated with the mechanical behavior of concrete and properties of rust on the time of surface cracking, the maximum values of consumed rebar per unit area of anode and crack width openings at the time of surface cracking, and the maximum value of radial stress at the rust–concrete interface are scrutinized in some detail.     

Strength and stiffness of reinforced concrete containments subjected to seismic loading: Research results and needs

The primary purpose of this paper is to present results of an experimental investigation on the strength and stiffness of reinforced concrete subjected to combined biaxial tension and simulated seismic forces. The test specimens represent a section of a wall of a containment structure carrying combined pressurization and seismic loading. Shear stiffness and strength, and their degradation with shear cycling, are given, along with simple expressions for predicting strength and extensional stiffness. The secondary purpose of the paper is to discuss research needs for improved prediction of the response of containment structures to seismic effects.     

The Fort St. Vrain high temperature gas-cooled reactor: VII. Prestressed concrete reactor vessel (PCRV) performance

The prestressed concrete reactor vessel (PCRV) designed and constructed by the reactor vendor, General Atomic Co., has performed satisfactorily since 1971. Some difficulties that have occurred include an inadvertent overpressure of its core support floor during construction, lengthy moisture removal times from its thermal barrier after moisture ingresses into the PCRV cavity, heat exchanger corrosion and localized high concrete temperatures adjacent to its penetrations and core barrel.     

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.     

CFD analysis of a regular sector of the ITER vacuum vessel. Part I: Flow distribution and pressure drop

The 3D steady-state Computational Fluid Dynamics (CFD) analysis of the ITER vacuum vessel (VV) regular sector #5 is presented, starting from the CATIA models and using a suite of tools from the commercial software ANSYS FLUENT^®. The peculiarity of the problem is linked to the wide range of spatial scales involved in the analysis, from the millimeter-size gaps between in-wall shielding (IWS) plates to the more than 10 m height of the VV itself. After performing several simplifications in the geometrical details, a computational mesh with ～50 million cells is generated and used to compute the steady-state pressure and flow fields from a Reynolds-Averaged Navier–Stokes model with SST k-ω turbulence closure. The coolant mass flow rate turns out to be distributed 10% through the inboard and the remaining 90% through the outboard. The toroidal and poloidal ribs present in the VV structure constitute significant barriers for the flow, giving rise to large recirculation regions. The pressure drop is mainly localized in the inlet and outlet piping.     

Analysis of LMFBR containment response using a multifield implicit continuous-fluid eulerian code (MICE)

An Eulerian computer code, MICE, for analyzing multifield-fluid flow problems in LMFBR containments is presented. The hydrodynamics of the MICE code is based upon the implicit multifield (IMF) method which includes the treatment of multifield fluids, the interpenetration of materials, the heat transfer, and the material phase changes. The finite-difference equations and the numerical techniques used in obtaining the equilibrium pressures and in calculating the fluid-structure interactions are described in detail. Sample problems are given to illustrate the capabilities of the code.     

Cermet fuel in a light water reactor: a possible way to improve safety. Part I. Fabrication and characterization

The concept of cermet (ceramic-metallic) fuel for LWRs is considered. Cermet fuel utilization allows one to reduce the fuel operating temperature and to increase reactor safety under DBAs (Design Basis Accident). A general consideration of the cermet fuel application as well as design features and fabrication techniques of cermet fuel pins are presented in the paper.     

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.     

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.     

Validation of coupled neutron kinetic/thermal-hydraulic codes. Part 2: Analysis of a VVER-440 transient (Loviisa-1)

Several three-dimensional hexagonal reactor dynamic codes have been developed for VVER type reactors and coupled with different thermal-hydraulic system codes. Under the auspices of the European Union's Phare programme these codes have been validated against real plant transients by the participants from 7 countries. Two of the collected five transients were chosen for validation of the codes. Part 1 of this article consists of validation against VVER-1000 reactor data. This second part is focussed to validation against measured data of ‘One turbo-generator load drop experiment' at the Loviisa-1 VVER-440 reactor. The experiment was performed just after plant modernisation and more measured data was available to validation than in normal operation of real plants. Good accuracy of the results was generally achieved comparable to the measurement accuracy. The confidence in the results of the different code systems has increased, and consequences of certain model changes could be evaluated.     

PARIS project: Radiolytic oxidation of molecular iodine in containment during a nuclear reactor severe accident: Part 2. Formation and destruction of iodine oxides compounds under irradiation - Experimental results modelling

In the case of a severe accident in a nuclear Light Water Reactor (LWR), the high radiation fields reached in the reactor containment building due to the release of fission products from the reactor core would induce air radiolysis. The air radiolysis products (ARP) could, in turn, oxidise gaseous molecular iodine (I₂) into aerosol-borne iodine-oxygen-nitrogen compounds, abbreviated as iodine oxides (IO_x). These reactions involve the conversion of a gaseous iodine compound resulting in a change of the iodine depletion rate from the containment atmosphere. Kinetic data were produced within the first part of PARIS project on the air radiolysis products formation and destruction. The second part of the PARIS project as presented in this paper deals with the impact of the ARP on the conversion of I₂ into IO_x. The objective was to provide a database to develop new or to validate existing kinetic models of formation and destruction of iodine oxides.The iodine tests of the PARIS project, performed at very low, realistic iodine concentrations, constitute an important database to further develop or validate empirical and mechanistic models on radiolytic I₂ oxidation. In the presence of painted surface areas or silver aerosol surface areas, radiolytic I₂ oxidation is negligible compared to I₂ adsorption on these surfaces for the conditions examined. However, radiolytic I₂ oxidation remains very efficient if surface areas are small or if they are made of the relatively non-reactive stainless steel.