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.     

Conceptual design of helium cooling circuit for irradiation target

Small metal specimens of about 20 mm × 20 mm and 0.4 mm thick are irradiated in cyclotron facilities for radiation damage studies. Cooling of these specimens is an important factor which decides the intensity of irradiation. In this paper helium is used for the cooling of irradiation target specimen. In order to have enhanced heat removal from the specimen jet cooling is employed. The cooling scheme and the conceptual helium cooling circuit has been arrived at based on the empirical correlation available in the literature. The heat removal rate has been estimated for various jet velocities. Experiments with impinging air jets have been carried out to compare the empirical predictions. Numerical predictions have also been carried out using commercial Computational Fluid Dynamic (CFD) code. Experimental predictions are 35%–55% higher compare to empirical correlation. The empirical correlation is 30% higher compare to CFD predictions.     

Research requirements for improved design of reinforced concrete containment structures

Reinforced concrete is a competitive material for the construction of nuclear power plant containment structures. However, the designer is constrained by limited data on the behavior of certain construction details which require him to use what may be excessive rebar quantities and lead to difficult and costly construction. This paper discusses several design situations where research is recommended to increase the designer's options, to facilitate construction, and to extend the applicability of reinforced concrete to such changing containment requirements as may be imposed by an evolving nuclear technology.     

Plastic analysis of concrete containment structure

Overpressure capacity of a box type concrete containment structure is evaluated. Plastic analysis of the finite element model is performed using a quadrilateral plate element of homogeneous material. A special approach is used to represent nonlinear properties of reinforced concrete, such as concrete cracking and crushing and steel yielding. Those properties are represented by a set of idealized stress-strain curves of equivalent homogeneous sections.The analysis allows for a better estimate of the overpressure capacity of the containment structure while keeping the computer cost low by avoiding the use of the more expensive reinforced concrete brick element.     

Analysis of large-scale tests for AP-600 passive containment cooling system

All next-generation light water reactors utilize passive systems to remove heat via natural circulation and are significantly different from past and current nuclear plant designs. One unique feature of the AP-600 is its passive containment cooling system (PCCS), which is designed to maintain containment pressure below the design limit for 72 h without action by the reactor operator. During a design-basis accident (DBA), i.e., either a loss-of-coolant or a main-steam-line break accident, steam escapes and comes in contact with the much cooler containment vessel wall. Heat is transferred to the inside surface of the steel containment wall by convection and condensation of steam and through the containment steel wall by conduction. Heat is then transferred from the outside of the containment surface by heating and evaporation of a thin liquid film that is formed by applying water at the top of the containment vessel dome. Air in the annular space is heated by both convection and injection of steam from the evaporating liquid film. The heated air and vapor rise as a result of natural circulation and exit the shield building through the outlets above the containment shell. All of the analytical models that are developed for and used in the COMMIX-1D code for predicting performance of the PCCS will be described. These models cover governing conservation equations for multicomponents single-phase flow, transport equations for the k − two-equation turbulence model, auxiliary equations, liquid-film tracking model for both inside (condensate) and outside (evaporating liquid film) surfaces of the containment vessel wall, thermal coupling between flow domains inside and outside the containment vessel, and heat and mass transfer models. Various key parameters of the COMMIX-1D results and corresponding AP-600 PCCS experimental data are compared and the agreement is good. Significant findings from this study are summarized.     

Serviceability design load factors and reliability assessments for reinforced concrete containment structures

A reinforced concrete nuclear power plant containment structure is subjected to various random static and stochastic loads during its lifetime. Since these loads involve inherent randomness and other uncertainties, an appropriate probabilistic model for each load must be established in order to perform reliability analysis. The current ASME code for reinforced concrete containment structures are not based on probability concepts. The stochastic nature of natural hazard or accidental loads and the variations of material properties require a probabilistic approach for a rational assessment of structural safety and performance. The paper develops probability-based load factors for the limit state design of reinforced concrete containment structures. The purpose of constructing reinforced concrete containment structure is to protect against radioactive release, and so the use of a serviceability limit state against crack failure that can cause the emission of radioactive materials is suggested as a critical limit state for reinforced concrete containment structures. Load factors for the design of reinforced concrete containment structures are proposed and carried out the reliability assessments.     

Limit analysis and design of containment vessels

In the introduction, the theory of plastic analysis of shells is briefly recalled. Minimum-volume design for assigned load factor at plastic collapse is then considered and optimality criteria are derived for plates and shells of continuously varying or piecewise-constant thickness.In the first part, containers made of metal are examined. Analytical and numerical limit analysis solutions and corresponding experimental results are considered for various types of vessels, including intersecting shells. Attention is given to experimental post-yield behavior. Some tests up to fracture are discussed. New theoretical and experimental results of limit analysis of stiffened cylindrical vessels are presented, in which reinforcing rings are treated as discrete structural element (no smearing out) and due account is taken of their strong curvature. Cases of collapse by instability under internal pressure are pointed out. Minimum-volume design of circular plates and cylindrical shells is then formulated and various examples are presented of sandwich and solid metal structures. Containers of piecewise-constant thickness are given particular attention. Available experimental evidence on minimum-volume design of plates and shells is reviewed and commented upon.The second part deals with reinforced concrete vessels. Cylindrical containers are studied, from both points of view of limit analysis and of limit design with minimum volume of reinforcement. The practical use of the latter solutions is discussed.A third part reviews other loading cases (including cyclic and impact loads) and gives indications on corresponding theories, formulations and solution methods.The last part is devoted to a discussion of the limitations of the methods presented, within the frame of the “limit states” design philosophy, which is first briefly recalled. Considerations on further research in the field conclude the paper.     

Construction of a concrete containment model

This paper discusses the features and construction of a reinforced-concrete containment model that has been built at Sandia National Laboratories in Albuquerque, New Mexico. The model Light-Water-Reactor (LWR) containment building was designed and built to the American Society of Mechanical Engineers (ASME) code by United Engineers and Constructors, Inc. The containment model will be tested to failure to determine its response to static internal overpressurization. The results from testing the heavily instrumented containment will be used to assess the capability of analytical methods for predicting the performance of containments subject to severe accident loads as part of the US Nuclear Regulatory Commission's program on containment integrity.The scaled dimensions of the cylindrical wall and hemispherical dome are typical of a full-size containment. Features representative of a prototypical containment and included in the heavily reinforced model are equipment hatches, personnel airlocks, several small piping penetrations, and a thin steel liner attached to the concrete by headed studs.     

Concept study for a combined reinforced concrete containment

A variety of different types of steel and concrete containments have been designed and constructed in the past. Most of the concrete containments had been pre-stressed, offering the advantage of small displacements and a certain leak-tightness of the concrete itself. However, considerable stresses in concrete as well as in the tendons have to be maintained during the whole lifetime of the plant in order to guarantee the required pre-stressing. The long-time behaviour and the ductility in the case of beyond-design-load cases must be verified. Contrary to a pre-stressed containment a reinforced containment will only be significantly loaded during test conditions or when needed in case of an accident. It offers additional margins which can be used especially for dynamic loads such as impacts or for beyond-design events.The aim of this paper is to show the feasibility of a so-called combined containment which means a containment capable of resisting both severe internal accidents and external hazards, mainly the aircraft crash impact as considered in the design of nuclear power plants in Germany.The concept is based on a lined reinforced containment without pre-stressing. The mechanical resistance function is provided by the reinforced concrete and the leak-tightness function is provided by a so-called composite liner made of non-metallic materials. Some results of tests performed at Siemens laboratories and at the University of Karlsruhe which show the capability of a composite liner to bridge over cracks at the concrete surface will be presented in the paper.The study shows that the combined reinforced concrete containment with a composite liner offers a robust concept with high flexibility with respect to load requirements, beyond-design events and geometrical shaping (arrangement of openings, an integration of adjacent structures). The concept may be further optimized by partial pre-stressing at areas of high concentration of stresses such as at transition zones or at disturbances around large openings.     

Capability of a prestressed concrete containment for internal pressure load

This paper presents a study for a PWR prestressed concrete containment which determines a realistic lower bound internal pressure where no structural failure is anticipated. The paper indicates the analytical method used, the actual material properties investigated, and the failure criteria selected for the material stresses and strains.     

Refined cracked concrete analysis of concrete containment structures subject to operational and environmental loadings

Recent commercial nuclear power plant containment concepts involve the use of large reinforced concrete structures to form pressure boundaries. Where these structures are not provided with an integral steel liner, excessive cracking of the concrete under loads could result in the loss of the pressure boundary integrity with the risk of over-pressurization of other structures. Cracking of concrete is a local phenomenon and considerable detail must be included in any analytical model to obtain sufficiently refined results for the prediction of crack size and propagation. This imposes severe limitations on the overall size of structures or structural components for which detailed cracking analysis can be considered directly. To overcome this restriction, a two step procedure was developed in which linear analyses were performed to obtain the gross response, and nonlinear cracking analyses were performed for selected portions of the structure to evaluate local cracking in detail. Through iteration, compatibility of behavior between the linear and nonlinear analyses was achieved with the gross response being used to extrapolate the local cracking results to predict cracking over the entire structure. This paper discusses the analysis procedures for the detailed evaluation of cracking in large reinforced concrete structures and components. Analyses performed for an actual unlined reinforced concrete containment structure using these procedures are discussed and results are presented.     

Conceptual design for the superconducting magnet system of a pulsed DEMO reactor

A methodology has been developed to consistently investigate, taking into account main reactor components, possible magnet solutions for a pulsed fusion reactor aiming at a large solenoid flux swing duration within the 2–3 h range. In a conceptual approach, investigations are carried out in the equatorial plane, taking into account the radial extension of the blanket-shielding zone, of the toroidal field magnet system inner leg and of the central solenoid for estimation of the pulsed swing.Design criteria are presented for the radial extension of the superconducting magnets, which is mostly driven by the structures (casings and conductor jacket). Typical available cable current densities are presented as a function of the magnetic field and of the temperature margin.The magnet design criteria have been integrated into SYCOMORE, a code for reactor modeling presently in development at CEA/IRFM in Cadarache, using the tools of the EFDA Integrated Tokamak Modeling task force.Possible solutions are investigated for a 2 GW fusion power reactor with different aspect ratios.The final adjustment of the DEMO pulsed reactor parameters will have to be consistently done, considering all reactor components, when the final goals of the machine will be completely clarified.     

Study on the performance of an open-loop passive containment cooling system

The containment is an ultimate and important barrier to keep the radioactivity from release. The integrity of the containment is crucial to control the consequences of either loss of coolant accident or main steam line break accident. A passive containment cooling system concept designed to remove the heat by natural circulation means is proposed, which is composed of a series of heat exchangers, long connecting pipes with relative large diameter, valves, and a water tank. The performance of the system is numerically simulated and the self-developed codes are validated by the experimental data. The influences of several key parameters are investigated on the performance of the system from different aspects. The results confirm that four distinct operating stages could be experienced as follows: startup stage, single-phase quasi-steady stage, flashing speed up transient stage, and flashing dominated quasi-steady operating stage. Furthermore, the mechanisms of the ways through which the parameters influence the behaviors of the proposed system are thus analyzed. Moreover, the feasibility of the system is also commented on the basis of the numerical results.     

Allowable stresses to limit shearing deformation in reinforced concrete containment design

An extensive program of the U.S. Nuclear Regulatory Commission (NRC) to study reinforced concrete containment wall behavior has been completed for orthogonal reinforcement. The transfer of shear caused by the action of seismic load has been studied sufficiently to recommend the seismic shear design and allowable shear stresses. However, the recommendations made in this paper are not the NRC position for the design.     

Impact research for containment design

Engaged for many years in research work concerning the safety and integrity of nuclear containments, the first author has performed numerous theoretical and experimental investigations at this institute. Airplane crashes on nuclear power plants, as well as containment attacks by detonation and missiles generated by bursting vessels have been studied with respect to practical design. Also, a series of fundamental researches has been done to evaluate constitutive laws for shockwaves in concrete and constitutive relations for concrete with regard to strain rate effects. Further investigations have focused on friction phenomena for projectiles impinging on concrete.     

Conceptual design modifications of the cooling system of MNSR reactor to increase its maximum continuous operation time

To increase the maximum daily operation time of Miniature Neutron Source Reactor (MNSR) reactor several conceptual thermal hydraulic design modifications have been investigated aiming at the improvement of reactor cooling conditions to limit the increase of average core temperature. For this purpose an integrated full-scale thermal hydraulic-neutronics model using the advanced code ATHLET has been developed, tested and verified. The selected design modifications rely upon introducing auxiliary cooling systems operating in four different modes to cool pool water or reactor water using heat exchanger located either inside or outside of reactor pool. The simulation results show that the increase of continuous reactor operation time varies between 1 and 8 additional operation hours. The optimal results are achieved for the second and the fourth options that use external heat exchanger. The second option enables the extending of continuous operation time up to 10 h and the fourth up to 15 h, both at nominal reactor power and under the assumption of initial excess reactivity corresponding to the fresh reactor core. The analysis included the evaluation of xenon poisoning effect on the increase of operation time. It has been shown that its remarkable effect starts after the first 3 operation hours and increases continuously after that. For the best cooling options, where the average core temperature is being fixed at certain value resulting in complete elimination of reactivity feedback of cooling temperature, xenon effect becomes the exclusive limiting effect during the later operation phase. The analysis discuss also general aspects of technical realization for the different cooling options in relation with the specific features of MNSR and the preliminary engineering safety measures and operational radiological protection that have to be taken. The performed analysis and the achieved results during this work would make valuable contribution for updating the Safety Analysis Report (SAR) of MNSR.     

An experimental investigation of a passive cooling unit for nuclear plant containment

A set of condensation experiments in the presence of noncondensables (e.g. air, helium) was conducted to evaluate the heat removal capacity of a passive cooling unit in a post-accident containment. Condensation heat transfer coefficients on a vertically mounted smooth tube have been obtained for total pressure ranging from 2.48×10⁵ Pa(abs) to 4.55×10⁵ Pa(abs) and air mass fraction ranging from 0.30 to 0.65. An empirical correlation for heat transfer coefficient (h), has been developed in terms of a parameter group made up of steam mole fraction (Xs), total pressure (P_t), temperature difference between bulk gas and wall surface (dT). This correlation covers all data points within 20%. All data points are also in good agreement with the prediction of the diffusion layer model (DLM) with suction and are approximately 2.2 times the Uchida heat transfer correlation. Experiments with an axial shroud around the test tube to model the restriction on radial flow experienced within a tube bundle demonstrated a reduction of the heat transfer coefficient by a factor of about 0.6. The effect of helium (simulating hydrogen) on the heat transfer coefficient was investigated for helium mole fraction in noncondensable gases (X_He/X_nc) at 15, 30 and 60%. It was found that the condensation heat transfer coefficients are generally lower when introducing helium into noncondensable gas. The difference is within 20% of air-only cases when X_He/X_nc is less than 30% and total pressure is less than 4.55×10⁵ Pa(abs). A gas stratification phenomenon was clearly observed for helium mole fraction in excess of 60%.     

Conceptual design and analysis of a passive residual heat removal system for a 10 MW molten salt reactor experiment

A conceptual design of a passive residual heat removal system was developed for a 10 MW molten salt reactor experiment (MSRE) designed by Oak Ridge National Laboratory (ORNL). The principle, main components and design parameters of the system were presented, and thermal-hydraulic behaviors, such as natural circulation and heat removal ability, were numerically analyzed in the code of C++, especially for the bayonet cooling thimbles. The results show that the system can effectively remove decay heat in the molten salt in an MSRE and has a heat removal rate that approximates to the decay heat generation rate, thus causing the temperature of the molten salt to decrease steadily. The width of the gas gap in the bayonet cooling thimbles has little effect on either the heat exchange or the natural circulation inside the thimbles, while the width of the steam riser, in spite of its slight effect on the heat transfer of the system, greatly influences the natural circulation. With the width of the steam riser increase from 3.6 to 5.1 mm, the mass flow rate increases from 1.9 kg/s to 4.79 kg/s. Finally, three operational schemes were proposed for the passive residual heat removal system, among which that of reducing the bayonet cooling thimbles by three-quarters had the best comprehensive performance.