Containment Shell Research Articles

Structural Examination of Iridium-Based Single-Crystal Preparations

A high-temperature crystal growth experiment produced discrete singlecrystal products of Allr and IrSi. The preparation, and examination of these phases is described within. This project is part of a materials compatibility study relating to radioisotopic heat sources. These heat sources are comprised of a PuO2 fuel pellet encapsulated in an Ir alloy containment shell. Thorium is introduced as an additive within the Ir to maintain ductility. Si and P are picked up inadvertently in the fuel processing. The compatibility of the heat sources with Al is of interest because of potential interactions with the Al alloy hardware associated with the heat source environment.

Spherical steel containments of pressurized water reactors under accident conditions: Investigation program and first results

The investigations will deal with the mechanical behavior of a free standing spherical containment shell built for the latest type of a German pressurized water reactor. The diameter of the containment shell is 56 m. The wall thickness is 38 mm. The material used is the ferritic steel 15MnNi63. The investigation program includes theoretical as well as experimental activities and concerns four different accidents which are beyond the scope of the common design and licensing practice: containment behavior under quasi-static pressure increase up to containment failure; containment behavior under high transient pressures; containment vibrations due to earthquake loadings (consideration of shell imperfections); containment buckling due to earthquake loadings. First results concerning the containment behavior under quasi-static pressure increase are presented. It turns out that the mechanical failure of the containment shell is controlled by plastic instability. A computer program to describe this problem has been developed and membrane tests to check the computational methods have been carried out.

Research related to buckling design of nuclear containment

This paper discusses the recent experimental and analytical studies related to buckling design of fabricated steel shells. The effects of initial imperfections and residual stresses on buckling are under investigation. The test programs include ring and stringer stiffened as well as ring stiffened cylinders subject to combinations of axial compression and external pressure. Proposed modifications to ASME Code Case N-284, “Metal Containment Shell Buckling Design Methods,” as well as the need for additional research, are discussed.

The Development of Durable, Man-Made Containment Systems for Fuel Isolation

Progress in the Canadian program to develop high integrity containment systems for the direct disposal of irradiated fuel in a deep geologic vault is described. Possible containment lifetimes, and the factors that influence container design and materials selection, are discussed. Most effort is currently directed toward developing containers with a design lifetime of about 500 yr, in which a corrosion-resistant metal shell provides the primary barrier to radionuclide release. Two design concepts are being studied, one in which the container shell completely supports the service stresses, and one in which an internal medium provides support for the outer shell. The design details for various containers are described, together with the fabrication and testing programs that are in progress or planned. The possibility of incorporating additional independent barriers to extend the containment life, and of developing long-term containment systems based on ceramic, vitreous or thick metal containers, is discussed. Finally, the generic and applied materials programs to evaluate the fabrication and performance of containers for fuel isolation are described.

The Development of Durable, Man-Made Containment Systems for Fuel Isolation

AbstractProgress in the Canadian program to develop high integrity containment systems for the direct disposal of irradiated fuel in a deep geologic vault is described. Possible containment lifetimes, and the factors that influence container design and materials selection, are discussed. Most effort is currently directed toward developing containers with a design lifetime of about 500 yr, in which a corrosion-resistant metal shell provides the primary barrier to radionuclide release. Two design concepts are being studied, one in which the container shell completely supports the service stresses, and one in which an internal medium provides support for the outer shell. The design details for various containers are described, together with the fabrication and testing programs that are in progress or planned. The possibility of incorporating additional independent barriers to extend the containment life, and of developing long-term containment systems based on ceramic, vitreous or thick metal containers, is ...

Deterministic and stochastic earthquake response analysis of the containment shell of a nuclear power plant

Response of the containment shell of a nuclear plant to earthquake ground motion is considered. A finite element model of the structure is developed and SAP IV structural analysis program is employed for the determination of the frequencies and the corresponding mode shapes of the structure. The response of the containment shell to several past earthquakes are analyzed and the results are discussed. Stochastic models of earthquake ground acceleration are then considered and the general expressions for the power spectra, cross correlations and the mean-square responses are derived. The root mean-square of the relative displacement responses of various nodal points of the containment shell structure subjected to stationary as well as nonstationary random support motion are evaluated. The stochastically estimated maximum displacement responses are compared with those obtained from a deterministic analysis and reasonable agreements are observed.

Research needs for design of concrete containment structures

Potential failure modes of reinforced concrete containment shells are outlined, especially those associated with pressure-induced cracking and seismic forces. A summary is given of experimental and analytical research needed to evaluate tangential shear capacity and stiffness, the interaction between liner and cracked concrete, peripheral (punching) shear capacity, radial shear behavior, and nonlinear dynamic analysis approaches.

Simulation of containment cooling with outside spray after a core meltdown

In the most severe hypothetical loss of coolant accident, the reactor core melts and falls into the containment sump, there evaporating much of the sump water and raising the pressure within the containment building. One possible method to remove the decay heat is to cool the steel containment shell with an outside spray system. To perform the structural analysis needed to confirm the integrity of the containment, the thermal profile in the containment wall must first be found. The purpose of this work is to develop a computer code to calculate this transient profile. Other aspects such as hydrogen build-up are not considered in this code. The method uses relationships for the natural convection-partial condensation phenomena occurring at the containment internal surfaces, iteratively coupled to a one-dimensional heat balance at the wall to solve for the wall temperature as a function of angular position. A differential calculation as a function of time treats the thermodynamic changes within the containment as quasi-steady state. The result is a quick-running code capable of analyzing the temperature transient for several hours following the LOCA with a few minutes of computing time.

Dynamic stresses in spherical containments with pressure suppression system during steam condensation

Abstract Different physical phenomena which control the structural integrity of the containment shell in case of a postulated failure of the primary coolant system are discussed. Detailed analyses are carried through for the chugging phase, where steam will be blown into the water pool of the pressure suppression system. As reference geometry a German boiling water reactor type 69 is used. The fluid dynamics of the water pool is described by a boundary integral equation method. For the structural dynamics of the thin spherical containment shell analytical solutions of Flugge's shell equations are obtained. The feedback of structural deformations on the fluid-dynamic loadings, i.e., the effect of fluid-structure interaction is considered. It increases the loadings significantly in comparison to calculations without this effect. The steam condensation in the water pool is treated as a parameter. As results upper limit steam condensation scenarios are presented where the resulting stresses reach the admissible values. If these upper limit scenarios cover all the steam condensation events which may occur, the dimensions of the containment shell are adequate to steam condensation loadings. (According to many experiments the upper limit scenarios cover indeed the expected condensation events.) The computer programs used in this work are carefully checked by comparisons with other analysis methods and relevant experimental findings.

A splitting method for computing coupled hydrodynamic and structural response

A numerical method is developed for application to unsteady fluid dynamics problems, in particular to the mechanics following a sudden release of high energy. Solution of the initial compressible flow phase provides input to a power-series method for the incompressible fluid motions. The system is split into spatial and time domains which lead to the convergent computation of a sequence of elliptic equations. Two sample problems are solved, the first involves an underwater explosion and the second the response of a nuclear reactor containment shell structure to a hypothetical core accident. The solutions are correlated with experimental data.

Flow of Radioactive Fluid Through the Soil Surrounding a Power-Reactor Station After a Core Meltdown Accident

Ideal gas and multiphase flow through porous soil surrounding a nuclear power generating station after a core meltdown accident are analyzed with dimensionless variables. Results are applicable to many soil conditions. The extent of radioactive fluid penetration into the soil, predicted using methods originally developed by Morrison, indicates that years are required for activity to reach the ground surface if silty clays are present but only approximately 10 h if a path of sand is included. Source pressure decay caused by mass flow into the soil is small if the normal conditions of a large containment shell volume and low soil permeability exist. 7 references. (auth)

Static and heat transfer analysis of a sphere-cone intersection in a nuclear containment vessel

Structural analyses for nuclear power plants require advanced computational techniques such as computer codes derived from thin shell theory and the finite element method. The steel containment shell of a KWU boiling water reactor, type 69, has an integrated pressure suppression system. A considerable number of loading cases, including mechanical and thermal effects, had to be investigated. Several models were used for the analysis: 1. (i) First an axisymmetric BOXSHELL-model was prepared for the overall structure. 2. (ii) Second, a sector of 3·75 degrees was idealised with ASKA. 3. (iii) Finally, stress concentrations at the sphere-cone intersection of the containment vessel were calculated using a detailed NASTRAN model; boundary conditions were made compatible with the overall BOXSHELL-model.

Two-dimensional approximations to the three-dimensional soil-structure interaction problem

The feasibility of representing a three-dimensional soil-structure interaction problem by a plane strain model, and the errors involved in such representation, were studied. By comparing the rocking and translational force-displacement relationships for a rigid circular foundation placed on an elastic half-space and the corresponding relationships for a strip footing placed on an elastic half-plane it was found that it is not possible to obtain a two-dimensional representation that will approximate both the dynamic stiffness and radiation damping over a reasonable range of frequencies. Several two-dimensional models were considered and a measure of the errors involved is presented. In general, the two-dimensional models overestimate the radiation damping associated with the three-dimensional problem. To study the effects that the use of a two-dimensional plane strain model may introduce in the solution of the soil-structure interaction problem for typical nuclear power plants, a comparison was made between the system frequencies and modal dampings obtained for three and two-dimensional models. The corresponding response at the top of the containment shell, top of the internal structure, and base slab for a particular earthquake were also compared. It was found that by properly selecting the two-dimensional model it was possible to obtain close approximations to the system frequencies. However, since the dampings associated with the low frequency modes are overestimated, the earthquake response of the structure, as obtained by the two-dimensional model, is underestimated to a significant degree.

Vibration and stability of prestressed shells

Thin shells are widely used structural elements and are sometimes subjected to time-dependent loads after they have already acquired some prestress. The response of such a shell can be very different from that when the prestress is absent. The general methods of calculation of the response of the shell are discussed. The stability limit of the shell is viewed as a special result obtained in the free vibration analysis. An example of a typical containment shell of a nuclear power plant, housing the reactor, is worked out in detail.

Applications of similitude to the structural mechanics of reactors

The structural mechanics of a reactor involves the prediction of the behavior of the components of the reactor and the entire reactor system under a combination of gravitational and other force fields, electrical and magnetic fields, thermal fields and radiation fields. The nature of the materials and components is such that under the operational conditions imposed, the behavior is often nonlinear, with the result that the application of the usual techniques of mathematical analysis to predict behavior often results in lengthy computations based on assumptions that are somewhat subjective. Although computer techniques have proven very helpful in many instances, they are often limited in scope through size of available computer storage. The direct approach, that of cut-and-try design, is often not appropriate because of the possibility of release of radioactive materials in the event of a faulty design or a failure of the containment or other materials. In addition, economics, as well as safety, usually precludes construction of a complete operating unit without preliminary analysis Under these conditions of complex mathematical analysis and expensive construction the use of small scale models is suggested as a method of obtaining data from which realistic predictions may be made. In the design of small scale models the theory of similitude is used in the same way that it has been used in many fields of engineering. It is also pertinent to consider current power reactors in the 400-MW range as models of projected reactors in the 2000-MW range. Direct application of the principles of similitude leads to a series of design conditions that cannot be satisfied simultaneously. In such a situation the application of distorted model theory is mandatory. The application of distorted model theory requires the determination of the prediction factors associated with the pertinent distortion factors derived from the significant parameters that are distorted. Evaluation of the prediction factors as functions of the distortion factors may be accomplished experimentally or analytically, and the remainder of the presentation is devoted to consideration of the applicable methods and procedures for determining the prediction factors. The analytical procedures may be classed as those based on well established laws and those based on fundamental principles. Specific illustrations of the application of the principles includes fuel element containment, distortion of primary containment, performance of containment shells as a result of sudden release of energy within or from external disturbances, and thermal strains resulting from loss of coolant and subsequent sudden cooling. As the numerical examples will illustrate, distorted model theory often enables one to make practical judgments concerning which distortions are significant and which will have a trivial influence upon the results.

A new technique for reactor siting dose calculations.

The COMRADEX code was developed to calculate radiation doses from accidents postulated for advanced reactors with sophisticated engineered safeguards. The use of a direct numerical technique, designed especially for quasilinear differential equations, to solve the leakage and decay chains, permits the inclusion of several special features. These include the capability for considering up to four levels of containment, each with an independent time-varying leak rate. Provision is also included for filtration, particle coagulation and fallout, and special cleanup systems. Shielding may be introduced between the various containment shells. A full fission product inventory (plus special isotopes) is used with input release fractions. The meteorology model is that of Pasquill and Gifford. Internal doses are determined with the ICRP model. The cloud gamma dose is obtained by integrating over the anisotropic Gaussian cloud; this calculation is simplified by assuming a single air attenuation kernel for all gammas.

Structural design criteria for secondary containment structures

The basic structural design philosophy and criteria for reactor secondary containment shells is developed. Considering the several loading conditions to which the structure may be subjected, the range of allowable stresses as they relate to desired performance characteristics under various combinations of loadings is discussed. There are three general conditions under which the structure must meet appropriate stress criteria: as a long-term plant enclosure subjected to normal live and dead loads; as a protective structure for the nuclear steam supply system subjected to abnormal external loads and forces such as tornadoes, hurricanes, maximum possible earthquake, and missiles that might be generated by a tornado, turbine failure, or airplanes; and as a containment structure subjected to internal pressure and temperature as a result of a major rupture of the main coolant system.

Shipping container for unirradiated fuel elements

The article gives a description of the problems met during the design of a shipping container for unirradiated nuclear fuel elements for a pressurized water reactor: classification of shipment; package design requirements; nuclear safety; design of fuel element fixture; design of suspension system between cradle and shell; design of container shell.

Model analysis of aseismic design of a nuclear reactor building

An experimental study on the response of a nuclear reactor building under earthquake loading was done by model testing on a random shaking table and a field test by blasting in order to get some data for aseismic design. In the case in which the basement is founded on firm rock, the lobar modes with natural frequency 3 ∼ 4 c/s are predominant at the container shell and the outer concrete shield.

The influence of activity release and removal effects on the escape of fission products from a double containment system

Double reactor containment buildings with associated use of filtration and washout systems are very efficient for the retention of fission products under accident conditions. An analytical treatment of teh activity release from such a containment system is given. A numerical example estimates the influence of a second containment on the release of the particularly dangerous isotopes iodine-131, and the results are compared with the release from a reactor system enclosed only by a single containment shell.