Differential pressures on building walls during tornadoes

In the United States, containment structures and some auxiliary structures (control buildings, auxiliary buildings, spent fuel buildings, etc.) in nuclear power plants are required to be designed to withstand the effects of the design basis tornado. In addition to velocity pressures and missle impact a tornado also causes a rapid change in atmospheric pressure, which can, in cases of closed or partially vented structures, produce direct differential pressure loading. In this paper a digital computer program is described which applies a tornado-induced, time-dependent atmospheric pressure change to a building and calculates the differential pressure histories across the interior and exterior walls of the building. Laws for quasi-steady, one-dimensional motion of an ideal compressible gas are used to calculate the pressures due to the flow of air through ports, doors and windows in the building. Numerical examples show that for each assumed atmospheric pressure change history a vent area to compartment volume ratio may be specified as the criterion for a building to be considered fully vented.     

Leakage potential of LWR containment penetrations under severe accident conditions

This paper is an overview of a Sandia National Laboratories, Albuquerque (SNLA) study of the performance of mechanical penetrations in light-water reactor (LWR) containment buildings that are subjected to severe accident environments. The study is concerned with modes of failure as well as the magnitude of leakage. The following tests have been completed, are under way, or are planned: (a) seals and gaskets have been tested to register the effects of radiation aging, thermal aging, seal geometry, and squeeze on seal and gasket materials in severe accident environments; (b) the performance of a full-scale airlock will be evaluated at severe accident temperature and pressure levels; (c) personnel airlock and equipment hatch tests were made on a model of a steel containment building; and (d) tests of mechanical penetrations are planned as part of a test on a model of a reinforced concrete building. This program is part of an overall US Nuclear Regulatory Commission (USNRC) effort to evaluate the integrity of LWR containment buildings.     

Steel containment buckling

Two aspects of buckling of a free-standing nuclear steel containment building were investigated in a combined experimental and analytical program. In the first part of the study, the response of a scale model of a containment building to dynamic base excitation is investigated. A simple harmonic signal was used for preliminary studies followed by experiments with scaled earthquake signals as the excitation source. The experiments and accompanying analyses indicate that the scale model response to earthquake-type excitations is very complex and that current analytical methods may require that a dynamic capacity reduction factor be incorporated. The second part of the study quantified the effects of framing at large penetrations on the static buckling capacity of scale model containments. Results show little effect from the framing for the scale models constructed from the polycarbonate, Lexan. However, additional studies with a model constructed of the prototypic steel material are recommended.     

Calculations of hydrogen detonations in nuclear containments by the random choice method

Computer codes were developed for the prediction of pressure histories at different points of a nuclear containment wall due to postulated internal hydrogen detonations. These pressure histories are required to assess the structural response of a nuclear containment to hydrogen detonations. The compressible flow equations including detonation, which was treated as a sharp fluid discontinuity, were solved by the random choice method which reproduces maximum pressures and discontinuities sharply. The computer codes were validated by calculating pressure profiles and maximum wall pressures for plane and spherical geometries and comparing the results with exact analytic solutions. The two-dimensional axisymmetric program was used to calculate wall pressure histories in an actual nuclear containment. The numerical results for wall pressures are presented in a dimensionless form, which allows their use for different combinations of hydrogen concentration, and initial conditions.     

Axisymmetric modeling of prestressing tendons in nuclear containment dome

Simple axisymmetric modeling of a nuclear containment building has been often employed in practice to estimate structural behavior for the axisymmetric loadings such as internal pressure. In this case, the prestressing tendons placed in the containment dome should be axisymmetrically approximated, since most dome tendons are not arranged in an axisymmetric manner. Some procedures are proposed that can realistically implement the actual three-dimensional tendon stiffness and prestressing effect into the axisymmetric model. Prestressing tendons, which are arranged in two or three ways depending on a containment type, are converted into the equivalent layer to consider the stiffness contribution in meridional and hoop directions. In order to reflect the prestressing effect, the equivalent load method and the initial stress method are devised, respectively, and the corresponding loads or stresses are derived in terms of the axisymmetric model. The proposed schemes are verified through some numerical examples comparing the results of the axisymmetric models to those of the actual three-dimensional model. The examples show that the proper level of the prestressing in the hoop direction of the axisymmetric dome plays an important role in tracing the actual behavior induced by the prestressing. Finally, some correction factors are discussed that can further improve the analysis results.     

Analytical study on dynamic buckling of reactor containment vessel

The dynamic buckling of a reactor containment vessel under earthquake conditions is evaluated using a nonlinear finite element method. It is based on the four-node MITC (mixed interpolated tensorial components) shell element originally proposed by K.J. Bathe, which has been modified by the authors to include the effect of large rotational increments. At first, the buckling modes for a thin cylindrical shell under a simplified base excitation were classified, then the dynamic buckling analysis of a typical PWR steel containment vessel was carried out, considering both geometrical and material nonlinearities, to compare the results with those of a conventional static analysis. It was found that the global shear buckling of a steel containment vessel occurred under a load level several times greater than the design earthquake, and the buckling load estimated by the conventional analysis was smaller than the buckling load estimated by the dynamic analysis.     

Analytical study on dynamic buckling of reactor containment vessel

The dynamic buckling of a reactor containment vessel under earthquake conditions is evaluated using a nonlinear finite element method. It is based on the four-node MITC (mixed interpolated tensorial components) shell element originally proposed by K.J. Bathe, which has been modified by the authors to include the effect of large rotational increments. At first, the buckling modes for a thin cylindrical shell under a simplified base excitation were classified, then the dynamic buckling analysis of a typical PWR steel containment vessel was carried out, considering both geometrical and material nonlinearities, to compare the results with those of a conventional static analysis. It was found that the global shear buckling of a steel containment vessel occurred under a load level several times greater than the design earthquake, and the buckling load estimated by the conventional analysis was smaller than the buckling load estimated by the dynamic analysis.     

Investigation of steel containment buckling under dynamic loads

Buckling of freestanding nuclear steel containment buildings from dynamic base excitation was investigated in a combined experimental/numerical program. A polycarbonate scale model of a containment building was excited with scaled earthquake transients and single-frequency harmonic transients to determine the peak base acceleration levels required to induce buckling. Buckling was identified using recorded signals from strain gages and accelerometers, with high-speed video records, and by audibility. Experimental results are compared with numerical results obtained by using a freezing-in-time technique. The results are preliminary, since several more tests are to be performed. However, the limited data obtained indicate that the freezing-in-time technique approximates the required acceleration levels reasonably well, although not conservatively. Additional experiments are described that will take containment asymmetries into account, as well as use instrumentation that will provide more accurate measures of the occurrence of buckling.     

Methods for assessing NPP containment pressure boundary integrity

Research is being conducted to address aging of the containment pressure boundary in light-water reactor plants. Objectives of this research are to (1) understand the significant factors relating to corrosion occurrence, efficacy of inspection, and structural capacity reduction of steel containments and of liners of concrete containments; (2) provide the U.S. Nuclear Regulatory Commission (USNRC) reviewers a means of establishing current structural capacity margins or estimating future residual structural capacity margins for steel containments and concrete containments as limited by liner integrity; and (3) provide recommendations, as appropriate, on information to be requested of licensees for guidance that could be utilized by USNRC reviewers in assessing the seriousness of reported incidences of containment degradation. Activities include development of a degradation assessment methodology; reviews of techniques and methods for inspection and repair of containment metallic pressure boundaries; evaluation of candidate techniques for inspection of inaccessible regions of containment metallic pressure boundaries; establishment of a methodology for reliability-based condition assessments of steel containments and liners; and fragility assessments of steel containments with localized corrosion.     

Soil-structure interaction for transient loads due to safety relief valve discharges

Dynamic responses of BWR Mark II containment structures subjected to axisymmetric transient pressure loadings due to simultaneous safety relief valve discharges were investigated using finite element analysis, including the soil-structure interaction effect. To properly consider the soil-structure interaction effect, a simplified lumped parameter foundation model and an axisymmetric finite element foundation model with viscous boundary impedance are used. Analytical results are presented to demonstrate the effectiveness of the simplified foundation model and to exhibit the dynamic response behavior of the structure as the transient loading frequency and the foundation rigidity vary. The impact of the dynamic structural response due to this type of loading on the equipment design is also discussed.     

Dynamic effects on structures and equipment due to Safety Relief Valve discharge loads

This paper presents analytical models, procedure, and results of a sensitivity study performed to investigate how the Safety Relief Valves (SRV) response of the steel containment of a Boiling Water Reactor (BWR) Mark III plant is influenced by the high frequency energy or noise associated with the idealized bubble forcing function. Various possible structural modifications to those plants already designed to have a steel containment are also presented and discussed with regard to minimizing the dynamic effects of SRV discharge loads.Special attention is given to the concept of filling concrete in the annulus between the steel containment and the shield building. The dynamic analyses of the containment structures with and without concrete in the annulus were performed and the results compared. Throughout the paper, several aspects of the SRV dynamic problem are emphasized and the relevant areas are identified for further investigations and studies.     

Leakage behavior of inflatable seals subject to severe accident conditions

The paper describes tests to determine the leakage behavior of inflatable seals when subjected to containment pressures that exceed the design basis.² Inflatable seals are used to prevent leakage around personnel and escape lock doors in about 10% of the commercial nuclear power plant containment structures in the United States. All of the installations are in either Pressurized Water Reactor (PWR) or Boiling Water Reactor (BWR) Mark-Ill type containments. This work is a part of an overall effort at Sandia National Laboratories to develop proven techniques for evaluating the performance of Light Water Reactor (LWR) containment buildings for beyond design basis loadings.Inflatable seals were tested at both room temperature and at elevated temperatures representative of postulated severe accident conditions. Parameters that were monitored and recorded during each test were the internal seal pressure and temperature, chamber (containment) pressure, leakage past the seals, and temperature of the test chamber and fixture to which the seals were attached. An empirically based, analytical method is presented to predict the containment pressure at which significant leakage past inflatable seals can be expected.     

Optimization Analysis on Pressure Suppression System of Small Reactor Containment

The containment pressure rises rapidly after LOCA, especially for the small reactors containment with very small free capacity, in order to avoid the rapid rise of containment pressure in the short term after LOCA, a pressure suppression system should be arranged in the containment. In this paper, the GOTHIC program was used to model the containment with pressure suppression system, and sensitivity analysis was carried out on the thermal response of containment after LOCA under different configuration schemes of pressure suppression system, the demonstration method of containment capacity with pressure suppression pool system and the optimal scheme were obtained. The analysis results show that the pressure suppression pool can significantly reduce the pressure in the containment, the pressure in the containment varies greatly under different configurations of pressure suppression pool modules, and the optimal configuration should be carried out for the containment design scheme in the design process.     

小型堆安全壳抑压系统优化分析

LOCA后安全壳内压力迅速升高,特别是自由体积较小的小型堆安全壳,为避免安全壳压力在LOCA后短期内快速升高,需在安全壳内配置抑压系统。本文通过采用GOTHIC程序对有抑压系统的安全壳进行建模并对不同抑压系统配置方案下LOCA后的安全壳热工响应进行敏感性分析,得到了有抑压水池系统的安全壳容量论证方法及抑压系统最优配置方案。分析表明：抑压水池能显著降低安全壳内的压力,不同抑压水池模块配置下安全壳内的压力差异较大,在设计过程中需针对安全壳设计方案进行优化配置。     

Accuracy of dynamic calculations using shell models under local impulse loading

Depending on soil conditions and load cases in dynamic calculations of nuclear power plants today more exact mathematical models may be used. For axisymmetric structures like reactor buildings, steel containments, circular tanks or coolant towers mathematical idealisations are used which especially deal with axisymmetric shell models. The calculations for these structures mentioned above, in the last 10 years, were generally carried out by applying specialised and qualified FE-programs.In order to qualify the results obtained using axisymmetric shell models as well the approved computer program MESY (Schrader 1976, 1978) several comparisons between computation and measurements were performed. As an example for these comparisons, impulse loadings, such as aircraft impact, applied by means of a pendulum on the HDR reactor will be shown.The analytical results were obtained prior to the general tests based on a loading function measured in a preliminary test step. In these calculations 11 harmonics were considered in the frequency range up to 80 Hz.Typical results will be shown and discussed, particularly the distribution of the maximum acceleration in the meridional and circumferential direction of the building.The analytical results for the structural response obtained using axisymmetric shell models conform satisfactorily to test results, especially in the area of load introduction in both (meridian and circumferential) directions.     

An investigation of containment pressurization by sodium spray fires

An investigation of the sodium spray burning phase of LMFBR hypothetical low probability core disruptive accidents (HCDAs) has been undertaken in order to test the response of various containment designs. The HCDAs are produced by arbitrarily inserting unrealistically large amounts of reactivity in a short period of time. The spray fires result from a HCDA which causes head failure due to high-velocity impact by the sodium pool followed by rotating plug jump, loss of rotating plug seals, control rod ejection, failure of instrument tubes, or breach of in-vessel transfer machine ports. Head failure can in principle lead to the injection of significant amounts of sodium into the reactor containment building by residual pressure of the HCDA gas bubble, which forces the upper plenum sodium through the interstitial spaces (e.g., rotating plug gaps, control rod housings, etc.) in the breached head structure.Calculations were made of the hydraulic behavior of the sodium under various injection scenarios for both pool and loop reactor systems. In the case of plug jump, although massive amounts of sodium can be injected into the containment building, the injection will be primarily in a radial direction, and the major consequences could be a low-intensity pool fire rather than a high-intensity sodium spray fire. However, if the bearing housing on the rotating plug fails, a 20 atm initial HCDA residual bubble pressure has the potential for injecting a sodium stream through the rotating plug gaps which could potentially impact the containment building ceiling. Sodium discharges through broken control rod housings could also impact the ceiling and become widely dispersed.The SOMIX-1 sodium spray fire code was used to calculate the energy releases corresponding to a variety of head failure scenarios corresponding to the cases where a high velocity jet impinges on the ceiling of the containment building. The calculated maximum pressure rise was about 2.1 atm. The analysis showed that containment building pressures do not always increase with increasing sodium injection rates since the oxygen concentration can be reduced to a level where the spray begins to cool rather than heat the gas.     

Three dimensional aspects of fast reactor containment studies

In assessing the strength of the primary containment of a pool type fast reactor with respect to the dynamic loading from core disruptive accidents (HCDA's), proper account must be taken of the 3-dimensional geometry of the components within the primary containment vessel. This paper reports on a series of experiments and the associated analysis carried out at AEE Winfrith to investigate this aspect of the containment loading. The experiments suggest that asymmetries in the containment loading, induced by symmetric or asymmetric rings of intermediate heat exchangers and sodium pumps, will be small. Calculations performed with the 2-dimensional axisymmetric code SEURBNUK provide a satisfactory estimate of the loads on the IHX's/pumps and their effect on the containment loading.     

Evaluation of Dynamic Loads Induced by Chugging

A dynamic load evaluation method has been proposed for chugging phenomena which are assumed to occur and produce relatively large amplitude pressure spikes in the pressure suppression pool of a BWR containment, in case of a postulated loss of coolant accident. The proposed method is based on the analysis code developed by the authors and on theseven vent full scale tests performed at Japan Atomic Energy Research Institute (JAERI CRT), considering random nature of chugging phenomena. The dynamic loads are obtained by applying the design source functions of impulsive nature to the vent pipe exists in each BWR containment analysis model. The design source functions are defined to produce dynamic pressures which reasonably envelope the design spectrum based on JAERI CRT data in frequency domain.

As an application example, the dynamic loads induced by chugging have been assessed based on the proposed method and on the reported JAERI CRT data from the view point of conservative load evaluation.

The applicability of the analysis code has also been confirmed, since the simulated dynamic pressures have shown features and magnitudes similar to those observed in JAERI CRT.     

Analysis of an ice-condenser containment building anchorage system

A 3-D non-linear finite-element analysis of an ice-condenser steel containment anchorage system, which considers the parameters that affect this complex system, was performed. The model included a portion of the containment shell, knuckle plate, base plate, reinforced concrete mat, anchor bolt, anchorage system, soil foundation material, and a portion of the containment shield building. The results showed the early formation of conical failure surfaces within the concrete that are associated with the brittle failure mode. However, these surfaces were not completely developed to the top of the containment basemat. No high strains were recorded in the anchorage system or the containment shell. Hence, failure of the containment anchorage system was not hypothesized.     

非能动安全壳冷却系统外侧辐射换热与自然对流研究

本文建立了1∶10的二维钢制安全壳外侧辐射换热和自然对流模型,并用先进流体计算软件Fluent对流场进行计算,得到了完整流道下的速度流场、钢制安全壳上封头顶部的空气速度矢量图,并得出钢制安全壳上封头顶部存在空气滞留区的结论。分析了通道宽度、空气进流速度及壁面黑度对通道换热的影响,结果表明：适当的通道宽度和空气进流速度均能提高通道的换热和换热效率;壁面黑度的提高能明显增强钢制安全壳上封头处的辐射换热。