Simulation of containment jet flows including condensation

The validation of a CFD code for light-water reactor containment applications requires among others the presence of steam in the different flow types like jets or buoyant plumes and leads to the need to simulate condensation phenomena.In this context the paper addresses the simulation of two “HYJET” experiments from the former Battelle Model Containment by the CFD code CFX. These experiments involve jet releases into the multi-compartment geometry of the test facility accompanied by condensation of steam at walls and in the bulk gas. In both experiments mixtures of helium and steam are injected. Helium is used to simulate hydrogen. One experiment represents a fast jet whereas in the second test a slow release of helium and steam is investigated. CFX was earlier extended by bulk and wall condensation models and is able to model all relevant phenomena observed during the experiments. The paper focuses on the simulation of the two experiments employing an identical model set-up. This provides together with other validation exercises the information on how well a wider range of flowing conditions in a full containment simulation can be covered with a single set of models (e.g. turbulence and condensation model). Some aspects related to numerical and modelling uncertainties of CFD calculations are included in the paper by investigating different turbulence models together with the modelling errors of the differencing schemes applied.     

Aerosol growth by steam condensation in a PWR containment

The fraction of steam which condenses as an aerosol in the cooling through surfaces of a well-mixed steam-air mixture confined in a cavity is calculated. A previous theory which predicts this fraction is extended so as to allow supersaturations, S, to occur in boundary layers, and the fraction is shown to be rapidly varying towards its maximum value which occurs in the saturated limit, S = 1. Only when the cooling surfaces are cold is the fractional aerosol condensation predicted to be large, and the aerosol is expected to evaporate slowly for small bulk-wall temperature differences. Results applicable to the cooling of PWR containments by water sprays are obtained. They indicate that it is unlikely that any net aerosol growth is produced.     

An overview of reactor containment structures

The purpose of this paper is to present an overview of reactor containment structures and to summarize the present state-of-the-art of containment design. The areas covered are types of containments used for nuclear power plants in operation and under construction, and their development. Also presented are codes which currently govern the design, materials, and construction of containments, as well as some thoughts on safety and methods of analysis.     

A diffusion layer model for steam condensation within the AP600 containment

Steam condensation plays a key role in removing heat from the atmosphere of the Westinghouse AP600 containment in case of a postulated accident. A model of steam condensation on containment surfaces under anticipated accident conditions is presented and validated against an extensive and sound database. Based on the diffusion layer theory and on the use of the heat/mass transfer analogy, one can deal with large temperature gradients across the gaseous boundary layer under high mass flux circumstances. The thermal resistance of the condensate film, as well as its wavy structure, have also been considered in this model. As compared to Anderson et al. (1998) (Experimental analysis of heat transfer within the AP600 containment under postulated accident conditions. Nucl. Eng. Des. (submitted)) experimental database, an average error lower than 15%, within the experimental confidence range, has demonstrated its remarkable accuracy. In particular, the model has shown a good response to the influence of primary variables in steam condensation (i.e. subcooling, noncondensable concentration and pressure), providing a mechanistic explanation for effects such as the presence of light noncondensable gas (i.e. helium as a simulant for hydrogen) in the gaseous mixture. In addition, the model has been contrasted against correlations used in safety analysis (i.e. Uchida, Tagami, Kataoka, etc.) and occasionally to Dehbi’s database. This cross-comparison has pointed out several shortcomings in the use of these correlations and has extended the model validation to other databases.     

Leak-tightness characteristics concerning the containment structures of the HTTR

The containment structures of the HTTR consist of the reactor containment vessel, the service area, and the emergency air purification system, which minimise the release of fission products in postulated accidents, which lead to fission product release from the reactor facilities. The reactor containment vessel is designed to withstand the temperature and pressure transients and to be leak-tight in the case of a rupture of the primary concentric hot-gas duct, etc. The pressure inside the service area is maintained at a negative pressure by the emergency air purification system. The emergency air purification system will also remove airborne radioactivity and will maintain a correct pressure in the service area.The leak-tightness characteristics of the containment structures are described in this paper. The measured leakage rates of the reactor containment vessel were enough less than the specified leakage limit of 0.1%/d confirmed during the commissioning tests and annual inspections. The service area was kept in a way that the design pressure becomes well below its allowable limitation by the emergency air purification system, which filters efficiency of particle removal and iodine removal well over the limited values.The obtained data demonstrate that the reactor containment structures were fabricated to minimise the release of fission products in the postulated accidents with fission product release from the reactor facilities.     

Theoretical study of aircraft impact on reactor containment structures

In containment design there is a requirement to protect the reactor system from the effects of external hazards and hence it is necessary to provide suitable wall thicknesses. Experimental work undertaken by the UKAEA is being carried out as a general study and this paper describes some theoretical studies for the particular case of an aircraft impact. The theoretical study utilizes a finite difference dynamic code based upon dynamic relaxation initially developed for static problems The code models concrete, reinforcement and prestressing throughout the short term non-linear range. Concrete is assumed to have a limited tensile stress capacity, coupled with a shear carrying capacity which is dependent upon the aggregate and crack size. In addition a yield condition can be specified to allow for triaxial stress states both initial and subsequent to failure. The paper briefly describes the theory and makes comparisons for different concrete thicknesses.     

Reliability assessment of reinforced concrete containment structures

A reliability analysis method for seismic category I structures subjected to various load combinations is developed and numerical examples are worked out under various assumptions and idealizations. The method falls generally within the so-called level III category within the framework of reliability analysis and design.     

Penetration resistance of reinforced concrete containment structures

Containment structures not only provide a leak tight barrier, but also play a role in ensuring that the structures can withstand the impact load from projectile impacts or internal plant accidents. In assessing the containment structures of nuclear power plants, predicting the characteristics of impact resistance in relation to design and safety considerations is relevant. This investigation proposes a simple but effective method of performing numerical analysis on perforation resistance of reinforced concrete containment structures. In this work, normal and oblique impacting is considered to examine the residual velocity and impact phenomena of an ogive-nose steel projectile with various impact velocities against a reinforced concrete slab. Additionally, a phase diagram is devised to describe the ballistic terminal phenomena of projectile and target. This model could assess the resistance to penetration to results in the optimum design of the containment structures in nuclear power plants.     

Modeling of turbulent condensation heat transfer in the boiling water reactor primary containment

A condensation heat transfer model is developed for the purpose of predicting the atmosphere temperature response within the primary containment of a boiling water reactor during the initial forced convection heat transfer period following a postulated loss-of-coolant accident. The model utilizes simultaneous heat and mass transfer for the process of condensation in the presence of a non-condensible gas. The gas-vapor diffusion layer formed is in the mode of turbulent, forced convection. The predicted heat transfer is determined to be diffusion controlled with negligible resistance being contributed by the condensate film. The model is qualified through the analysis of the response of a containment test facility; the results compare favorably with experimental observations made by the General Electric Co. Predicted temperature responses for a typical containment are also shown and compared with those obtained through use of the Uchida heat transfer correlations.     

Some aspects of the reliability-based design of reactor containment structures

A rational procedure for the design of reactor containment structures is carried out within a probabilistic framework. Various risk concepts such as the return period, non-encounter probability and the reliability function are discussed. Internal load conditions caused by system failure such as LOCA pressure loads, and external load conditions caused, for instance, by impact due to aircraft crashes, external pressure waves and natural hazards such as earthquakes and severe storms, are described by extreme value distributions of the largest values of the Fisher-Tippett types. Statistical and physical arguments are given to support their application. The occurrence of these rare events with respect to time is modeled by a Poisson process. The ultimate strength of a PWR containment structure for the steel (liner) shell is also modeled by an extreme value distribution (of the smallest values). As a good approximation the load action of the shell structure is determined by linear elastic analysis. The failure criterion considered here is that of reaching the ultimate tensile strength at one point of the structure. A numerical example of the reliability analysis of a steel shell structure under internal overpressure is carried out.     

Creep prediction of concrete for reactor containment structures

Since the biggest time-dependent prestress loss of a prestressed concrete nuclear reactor containment structure is due to the creep of concrete, creep is one of the most important structural factors to be considered for the safety of a reactor containment structure during design, construction and maintenance. Creep in concrete has also recently been considered in evaluation of the crack resistance of concrete at an early-age in the durability examination of massive concrete structures like reactor containment structures. Existing empirical formulas on creep prediction show errors in their predictions due to simplified consideration of mixture proportions, and they also show large discrepancy among their predictions. In addition, they do not consider early-age behaviors of concrete and thus are mainly for the prediction of long-term creep at hardened concrete. In this paper, the creep characteristics of the reactor's both early-age and hardened reactor concrete made of type V cement are examined by carrying out both early-age and long-term creep tests. Then, the creep of the reactor concrete is predicted by using major creep-prediction equations of the AASHTO LRFD design specification, the Japanese standard specification for concrete structure, the ACI Committee 209 and the CEB/FIP model code and the Bazant and Panula's model, and the predicted results are compared with the test results. From the comparison, the applicability of the creep-prediction equations for the concrete of a reactor containment structure at both early-age and hardened stages is discussed.     

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.     

Numerical models for prestressing tendons in containment structures

Two modified stress–strain relations for bonded and unbonded internal tendons are proposed. The proposed relations can simulate the post-cracking behavior and tension stiffening effect in prestressed concrete containment structures. In the case of the bonded tendon, tensile forces between adjacent cracks are transmitted from a bonded tendon to concrete by bond forces. Therefore, the constitutive law of a bonded tendon stiffened by grout needs to be determined from the bond–slip relationship. On the other hand, a stress increase beyond the effective prestress in an unbonded tendon is not section-dependent but member-dependent. It means that the tendon stress unequivocally represents a uniform distribution along the length when the friction loss is excluded. Thus, using a strain reduction factor, the modified stress–strain curve of an unbonded tendon is derived by successive iterations. In advance, the prediction of cracking behavior and ultimate resisting capacity of prestressed concrete containment structures using the introduced numerical models are succeeded, and the need for the consideration of many influencing factors such as the tension stiffening effect, plastic hinge length and modification of stress–strain relation of tendon is emphasized. Finally, the developed numerical models are applied to prestressed concrete containment structures to verify the efficiency and applicability in simulating the structural behavior with bonded and/or unbonded tendons.     

Iodine removal by condensation from containment atmosphere in post external event conditions

An integral solution is derived for the problem of passive removal rate of a typical fission product (iodine), from a gas-vapor mixture, by condensate liquid film adjacent to a containment wall. The analytical model consists of a coupled set of five conservation equations: momentum, energy and three matter conservation equations for each individual component of the gas mixture: air, steam and elemental iodine. The set is solved in conjunction with two balance equations for the mass and energy transport at the interface with the condensate layer. The model accounts for free convection due to temperature and concentration gradients, for mass and thermal diffusion and for variable properties in both the liquid and the gas-vapor regions. An economic solution procedure of this model is presented and employed for a wide range of parameters. The computational results of this study are used to derive an efficient correlation which provides a quick and simplified means of the calculation of the iodine mass removal coefficient, as a function of the bulk conditions. Some results are compared to other theoretical and experimental works showing good agreement within about 10%. The significance of the removal process in the “external event” scenario is analyzed and found to be much higher than in scenarios that start with a mechanical failure in the primary system.     

LOCA steam condensation loads in BWR Mark II pressure suppression containment system

Hydrodynamic loads induced in the BWR Mark II pressure suppression containment system during a loss-of-coolant accident (LOCA) were investigated using a large scale test facility. The maximum-bounding loading conditions on the pressure suppression pool-boundary structures were defined by conducting experiments for a wide range of parameters. The maximum-bounding loads occurred when steam, with air concentration less than 1% in weight, was injected at moderate rates ( 30 kg/m²·s) into a low-temperature (below 310 K) pool. Such conditions are most likely to be encountered during LOCAs with intermediate break sizes.     

Aging management of containment structures in nuclear power plants

Research is being conducted by Oak Ridge National Laboratory under US Nuclear Regulatory commission (USNRC) sponsorship to address aging management of nuclear power plant containment and other safety-related structures. Documentation is being prepared to provide the USNRC with potential structural safety issues and acceptance criteria for use in continued service evaluations of nuclear power plants. Accomplishments include development of a Structural Materials Information Center containing data and information on the time variation of 144 material properties under the influence of pertinent environmental stressors or aging factors, evaluation of models for potential concrete containment degradation factors, development of a procedure to identify critical structures and degradation factors important to aging management, evaluations of non-destructive evaluation techniques, assessments of European and North American repair practices for concrete, review of parameters affecting corrosion of metals embedded in concrete, and development of methodologies for making current condition assessments and service life predictions of new or existing reinforced concrete structures in nuclear power plants.     

Evaluation of calculational and material models for concrete containment structures

A computer code utilizing an appropriate finite element, material and constitutive model has been under development as a part of a comprehensive effort by the Electric Power Research Institute (EPRI) to develop and validate a realistic methodology for the ultimate load analysis of concrete containment structures. A preliminary evaluation of the reinforced and prestressed concrete modeling capabilities recently implemented in the ABAQUS-EPGEN code has been completed. This effort focuses on using a state-of-the-art calculational model to predict the behavior of large-scale reinforced concrete slabs tested under uniaxial and biaxial tension to simulate the wall of a typical concrete containment structure under internal pressure. This paper gives comparisons between calculations and experimental measurements for a uniaxially-loaded specimen. The calculated strains compare well with the measued strains in the reinforcing steel; however, the calculations gave diffused cracking patterns that do not agree with the discrete cracking observed in the experiments. Recommendations for improvement of the calculational models are given.     

Reliability-based assessment of integrity of bonded prestressed concrete containment structures

The main function of a nuclear containment structure is to prevent the leakage of radioactive materials from the reactor in the event of a serious failure in the process system. To maintain a high level of leak integrity, prestressed concrete is widely utilized in containment construction. In bonded prestressing systems, excessive prestressing losses caused by unexpected material deformations and degradation of tendons could result in the loss of leak integrity under an accident. To safeguard against this, the Canadian Standard, CSA N287.7 (1995), recommends periodic inspection and evaluation of prestressing systems of CANDU containments. As bonded tendons are not amenable to direct inspection, the evaluation is based on the testing of a set of beams with features identical to the containment. The paper presents a quantitative reliability-based approach to evaluate the containment integrity in terms of the condition of bonded prestressing systems. The proposed approach utilizes the results of lift-off, destructive, and flexural tests to update the probability distribution of prestressing force, and to revise the calculated reliability against through-wall cracking of containment elements. An acceptable criterion for the results of beam tests is established on the basis of maintaining adequate reliability throughout the service life of the containment.