Moisture flow in concrete under steady state non-uniform temperature states: experimental observations and theoretical modelling

The migration of water in concrete at temperatures up to 400 °C is controlled dominantly by pore vapour pressures. The free pore volume is as-cast concrete serves as a reservoir during the migration process and plays an important role in the mitigation of high pore vapour pressures in the hottest regions.Experimental results are presented for two concretes containing limestone and basalt aggregates. They illustrate two sets of overheat conditions in reactor containment walls: (i) long-term service conditions at steady state liner temperatures in the range 105–200 °C, with and without pressure venting close to the liner; (ii) short-term transient behaviour for an accident with temperatures to 400 °C. The results show the distributions of free and bound water in walls of two thicknesses (1.55 and 3.1 m) after approximately 1.5 years from the imposition of a temperature crossfall. The vented experiments confirm significantly higher rates of drying and the ability of water to migrate towards higher temperature locations when driven by pore pressure gradients which are in opposition to the local temperature gradients.A theoretical model, based on pore pressure gradients as the driving potentials, is introduced and used to predict water migration in a concrete wall of 5 m thickness, heated at the inner face to 200 °C. It is suggested that thick walls will take many years to dry significantly, eventhough they dry simultaneously near to the liner and at the exposed cold face. Finally it is demonstrated that the theoretical model is capable of predicting this special behaviour and therefore has an advantage over diffusion-based analyses which cannot model this feature.     

Moisture propagation and resulting stress in heated concrete walls

A mathematical model to simulate the coupled heat and mass transfer in heated porous media, as well as the resulting stress, is described. A finite element analysis, assuming homogeneous elastic material, is used to study the temperature and pore pressure distribution, and the rate of moisture propagation through a concrete containment wall under different time-dependent temperature boundary conditions. Results are also presented for the internal stresses caused by the presence of temperature gradients, pore pressure and the release of chemically bound water at high temperatures. Stress analysis calculations are superimposed over the calculations of the moisture propagation. The temperature, pore pressure and volume change resulting from the loss of bound water, as derived by the thermal mass transport calculation, are used as input for the stress analysis.     

Thermal shock in a nuclear containment wall due to loss-of-coolant accident

The safety analysis of reinforced concrete containments for nuclear reactors requires evaluation of thermal effects due to the loss-of-coolant accident. It is assumed that the inner surface of the containment is suddenly heated by the coolant getting out of the primary circuit. The wall is assumed to behave as a beam-column. Hoop forces and moments created by shell action are ignored. The plane wall section of the containment, normal to the reinforcing rods, is studied for evaluation of the stress increment due to the thermal shock. It is assumed that the section remains plane during the mechanical and thermal loading. The elastic-plastic model of material is chosen both for concrete and reinforcing steel. The section is considered cracked whereever the concrete is subjected to tensile stress. Thermal and mechanical material data are included in the program.The input-data for the computer program consist of the temperature of coolant inside the containment, the coefficient of heat transfer from the coolant to the wall, the axial force and the bending moment imposed by loading conditions before the thermal shock and the conditions of restraint for the considered wall section during the thermal shock. The computer program, based on the finite element method, consists of two sets of subroutines. The first set calculates the temperature increment after the prescribed time step. The second set calculates the elastic and plastic strain increments resulting from the increment of combined mechanical and thermal loading. The wall section of an actual containment is used, as an illustrative example, for the determination of thermal effect under various loading conditions. Results are presented in a diagram of axial force versus bending moment. A point on the diagram represents the load combination to which the wall section might be subjected. The thermal effect for various thermal loads, in form of the equivalent bending moment, it also plotted in the diagram.     

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.     

Study on behavior of tritium in concrete wall

It is required to understand the tritium behavior in concrete for establishment of tritium safety technology of a fusion reactor or a tritium handling facility because the concrete wall is used as the final containment to prevent tritium release to the environment. This paper discusses about the effect of adsorption and diffusion of water and isotope exchange reaction between physically adsorbed water and chemically adsorbed water or structural water. It is known in this study that a large amount of tritium can be trapped to the concrete wall because cement paste has the nature of porous hydrophilic material.     

Analytical correlation and interpretation of full scale containment specimen tests

The work presented in this paper is part of an EPRI-sponsored research program to develop experimentally verified methodology for predicting failure modes and leakage characteristics of concrete containments. This paper deals specifically with recent results of the analytical correlation and interpretation of full scale containment specimen tests. The tests under consideration are a wall/skirt-basemat specimen of a typical prestressed concrete containment, a specimen with a flawed liner to study liner crack growth, and a specimen with a typical steampipe penetration. Computational models of specimens are described, and pre-test and post-test analysis results are presented. The importance of local effects is discussed, and the role of specimen tests and analysis in failure prediction of containment structures is summarized.     

Experimental investigations of pressure and temperature loads on a containment after a loss-of-coolant accident

For the design of an LWR containment one of the important conditions to be considered is the rapid rise of internal pressure and temperature caused by a loss-of-coolant accident (LOCA) of the primary cooling system. The phenomena occurring within a containment during a LOCA are currently investigated through experiments with a model containment. The experimental results are compared with the results of model calculations to improve the calculational methods.An experimental facility was built, consisting of a primary coolant circuit and a special model containment. The model containment, built in conventional reinforced concrete, has a diameter of 12 m, a height of 12.5 m, a capacity of 580 m³ and is designed for an internal pressure of 6 bar. The interior is divided by concrete walls and removable partitions into several compartments, which are interconnected through openings with adjustable cross sections. By exchanging the removable partitions it is possible to modify the interior of the containment and to simulate different containment shapes. For the first experiments a PWR configuration with nine compartments has been installed. The model scales of the compartment volumes and the overflow areas are about 1:64 compared to the 1200 MW PWR plant Biblis A.Up to now the test facility has been used for four trial runs and nine PWR LOCA experiments with single- and double-ended pipe ruptures of 100 mm dia. in a steam generator compartment and in the nozzle compartment. The initial conditions of the pressurized water in the coolant circuit before rupture were 140 bar and 290°C. About 0.1 sec after the rupture the flow rate at the site of rupture reaches its maximum of about 400 kg/sec (single-ended rupture) and 800 kg/sec (double-ended rupture). From the compartment where the rupture takes place a water-steam-air mixture streams through openings into the other compartments of the containment. Differential pressures between the compartments were measured with maximums of up to a few bar 0.15–0.5 sec after rupture, depending on the positions of rooms and transducers.Approximately 30–40 sec after rupture the blowdown has finished and the pressure in the containment has reached about 4–5 bar. The maximum pressure in a model containment is lower and the decrease of the pressure by condensation is faster than in a full-scale containment, due to the greater ratio of inner surface area to volume of a model containment. During blowdown the temperature of the containment atmosphere rises to about 150°C. Several minutes later the temperature of the concrete walls has increased non-uniformly causing considerable stress in the walls. Approximately 30 min after rupture measurements on the outside of the outer containment wall show a temperature-caused strain of about 30–60% of the maximum pressure-caused strain. A comparison between experiments and calculations shows discrepancies indicating the need for further development of calculational methods.     

Model tests of turbine missile impact on reinforced concrete

The results of 25 impact tests on 1/11-scale models of reinforced concrete nuclear plant walls are presented. These tests determined experimentally the maximum velocity at which postulated turbine missiles are contained by typical reinforced concrete walls. The parameters varied were missile weight, velocity, orientation, and impact angle, as well as target design and thickness. The results showed that the NDRC perforation formula used extensively in current practice is overly conservative, whereas a newer empirical formula (CEA-EDF) gave reasonably conservative predictions of the test results. All but the most energetic postulated missiles are stopped by containment wall models, and steel liners on these walls are effective in suppressing backface concrete scabbing.     

Ultimate analysis of BWR Mark III reinforced concrete containment subjected to internal pressure

Numerical analyses are carried out by using the ABAQUS finite element program to predict the ultimate pressure capacity and the failure mode of the BWR Mark III reinforced concrete containment at Kuosheng Nuclear Power Plant, Taiwan, R.O.C. Material nonlinearity such as concrete cracking, tension stiffening, shear retention, concrete plasticity, yielding of reinforcing steel, yielding of liner plate and degradation of material properties as a result of high temperature effects are all simulated with proper constitutive models. Geometric nonlinearity as a result of finite deformation has also been considered. The results of the analysis show that when the reinforced concrete containment fails, extensive cracks take place at the apex of the dome, the intersection of the dome and the cylinder and the lower part of cylinder where there is a discontinuity in the thickness of the containment. In addition, the ultimate pressure capacity of the containment is 23.9 psi and is about 59% higher than the design pressure 15 psi.     

Preliminary evaluation of aircraft impact on a near term nuclear power plant

For the assessment of the safety and durability of a nuclear power plant (NPP), the containment building behaviour shall be evaluated, under various service and extreme conditions, both natural or produced by natural accident or vicious man activities, like September 2001 jet aircraft crashes.The aim of this paper is to preliminary evaluate the effects and consequences of the energy transmitted to the outer containment walls (according to the international safety and design code guidelines, as NRC or IAEA ones) due to a military or civil aircraft impact into a nuclear plant, considered as a ‘beyond design basis’ event.To perform reliable analysis of such a large-scale structure and determine the structural effects of the propagation of this types of impulsive loads (response of containment structure), a realistic but still feasible numerical model with suitable materials characteristics were used by means of which relevant physical phenomena are reflected. Moreover a sensitivity analysis has also been carried out considering the effects of different containment wall thickness and reinforced/prestressed concrete features. The obtained results were analysed to check the NPP containment strength margins.     

Prestressed concrete containment of nuclear power station with PWR

The initial steps of the development of prestressed concrete containment (PCC) for nuclear power plant (NPP) with pressurized water reactors (PWR) in the former USSR are analyzed. The constructive and technological decisions, accepted for primary PCC of Novovoronez NPP, such as the positioning of reinforcement elements and seaths in cylindrical wall and dome of the containment, the anchorage of reinforcement element ends, the technological aspects of concrete works, system and technology of a high level of biaxial pressing on a thin-wall structure at large wrapping angles of power reinforcing strands and etc. are observed. Experience won through the construction and operation of the primary PCC served as a basis for development of a new generation of improved unified PCC (IUPCC) for serial NPP, equipped with PWR of capacity of 1000 MW. The IUPCC is actually a cylinder 45 m in diameter and 54-m high covered with a gently sloping spherical dome. Thickness of cylinder wall is 1200 mm and that of dome wall is 1100 mm. The principle novelty of this PCC is the type and positioning of reinforcement strands. The paper describes strand arrangement and their anchorage in IUPCC. In the vertical part of PCC, strands are arranged on a helical-loop scheme and both strand ends are firmly anchored at the ring girder. Each strand is bended at the bottom of the containment. In the dome, strands are grouped on the orthogonal-loop scheme with the anchorage on one side and with bend of loop on the opposite side of the ring girder. To prevent the leakage of gases and to ensure tightness of the IUPCC an inner metal 8-mm liner with special anti-corrosion coating is designed. Monitoring and checking the stress and strain state of IUPCC is possible during the building, testing as well as operating periods. If any defects or decreased prestress of concrete are detected it is possible to tighten or replace the strands. It is noted that the more than 20 IUPCCs are in-service in Russia, Ukraine, and Bulgaria where NPP with PWR of capacity of 1000 MW were constructed.     

An experimental study on evaporative heat transfer coefficient and applications for passive cooling of AP600 steel containment

After TMI and Chernobyl accidents, many efforts have been made to enhance the nuclear safety with passive features. Among such passive features, the passive containment cooling system (PCCS) has been suggested by Westinghouse in the AP600 plant. The containment with PCCS is a dual containment, and consists of a stainless steel vessel and a concrete wall. In the gap between these structures, air and water can counter-currently pass and cool the steel surface. This paper experimentally investigates evaporative heat and mass transfer at the surface of a falling water film with counter-current air flow in a vertical duct with one-side heated plate. Experiments included various conditions of mass flow rate of film and air. Experimental results show the strong effects of water temperature and air mass flow rate, but little effect of the water flow rate. Also, simple analyses based on heat and mass transfer analogy were performed to evaluate the experimental results. With experimental data, a new correlation on evaporative mass transfer coefficient was developed, and with the correlation, the containment pressure and temperature was calculated for the design basis accident of AP600 by the use of CONTEMPT4/MOD5 code implementation.     

严重事故下预应力混凝土安全壳非线性分析及性能评估

预应力混凝土安全壳作为核电厂重要防泄漏屏障,对保证核电厂正常运行、确保人员安全至关重要。本文基于顺序热力耦合方法对严重事故工况下预应力混凝土安全壳进行非线性有限元分析,考虑了温度和内压荷载共同的影响,分析了安全壳结构典型不连续区域和连续区域的位移响应。研究表明:安全壳混凝土不连续区域位移响应沿厚度方向上差异较为显著,而连续区域处的差异相对较小;安全壳泄漏失效模式由设备闸门位置控制,50%和95%分位水平的内压分别为1.266 MPa和1.072 MPa;破口失效模式由筒体某一位置控制,50%和95%分位水平的内压分别为2.224 MPa和1.883 MPa;本文所分析的预应力混凝土安全壳的内压承载力满足最小安全裕度不小于2.5的要求。      

Thermal stress analysis of containment building in case of a Main Steam Line Break (MSLB)

In Indian PHWR, containment building is one of the primary barriers for mitigating the consequences of a Loss of Coolant Accident (LOCA), and Main Steam Line Break (MSLB) accident. It is desired to know the temperature transients as well as the resulting thermal stresses in the containment structures of 220MW(e) PHWR, Kaiga Nuclear Power plant under postulated MSLB event. The high enthalpy steam discharged into the containment space comes in contact with the Structural Wall (SW) of containment, Inner Containment Wall (ICW) and Raft. The containment wall temperature rises due to heat transfer from steam-air mixture. To calculate the transient temperature distribution across the containment walls, it is necessary to determine containment ambient temperature and heat transfer coefficient for the condensing steam on the internal structures. Hence, at the outset, a thermal hydraulic code was developed to predict the pressure-temperature transients and condensation heat transfer coefficient transients (using various condensation models) based on the mass and energy of high enthalpy steam released into containment. The effect of various condensation models on containment pressure-temperature was evaluated. The thermal boundary conditions such as containment temperature and heat transfer coefficient, evaluated from the thermal hydraulic code using Uchida condensation model, were subsequently applied as boundary conditions to a two-dimensional axi-symmetric containment model developed using a FEM code for estimating two-dimensional temperature profiles and the resulting thermal stresses.     

AP1000钢制安全壳厚度对传热性能的影响

AP1000是目前世界上安全性最高的第三代大型压水堆之一,相比于二代压水堆,其重要特征是将预应力混凝土的安全壳改为钢制安全壳,在整个冷却过程中钢制安全壳起着重要的作用。本文利用WGOTHIC程序建立AP1000整体长期空气冷却模型,对安全壳厚度进行研究,得到了传热性能与安全壳厚度的关系。结果表明,在一定范围内随安全壳厚度的增加,总体安全性得到较大提升,这为采用钢制安全壳的核电站设计提供了理论参考。     

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.     

Experiments to evaluate behavior of containment piping bellows under severe accident conditions

Bellows are an integral part of the containment pressure boundary in nuclear power plants. They are used at piping penetrations to allow relative movement between piping and the containment wall. In a severe accident they may be subjected to high pressure and temperature and a combination of axial and lateral deflections. A test program to determine the leak-tight capacity of containment penetration bellows is being conducted at Sandia National Laboratories, Albuquerque, NM. Several different bellows geometries representative of actual containment bellows are being subjected to extreme deflections along with pressure and temperature loads. The bellows geometries and loading conditions are described along with the testing apparatus and procedures. A total of 13 tests have been conducted. The tests showed that bellows are capable of withstanding relatively large deformations up to or near the point of full compression before developing leakage. The test data are presented and discussed.     

大型先进压水堆IVR策略评估和裂变产物分布计算

大型先进压水堆通过堆内熔融物滞留（IVR）策略来缓解严重事故后果以降低安全壳失效风险。其中堆腔注水系统（CIS）被引入来实现IVR。本文使用严重事故分析软件计算大型先进压水堆在冷管段双端断裂事故下的事故进程、热工水力行为、堆芯退化过程和下封头熔融池传热行为,评估能动CIS的事故缓解能力。计算结果表明,事故后72 h,下封头外表面热流密度始终低于临界热流密度（CHF）,表明IVR策略有效。此外,计算分析了惰性气体、非挥发性和挥发性裂变产物的释放和迁移行为。计算发现,IVR下更多的放射性裂变产物分布在主系统内,壁面核素再悬浮形成气溶胶的行为被消除,安全壳壁面上沉积的核素被大量冷凝水冲刷进入底部水池。总体来说,IVR策略能更好地管理放射性核素分布,减小放射性泄漏威胁。     

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.     

Testing of large-scale concrete containment structural elements

The tests described in this paper are part of an Electric Power Research Institute (EPRI) program (Research Project 2172-2) to provide a test-verified analytical method of estimating capacities of concrete reactor containment buildings under internal overpressurization from postulated degraded core accidents.Experimental study in Phase 2 of the investigation, on which this paper is based, includes tests of five large-scale specimens with steel liner plates representing structural elements of prestressed concrete containment buildings. Four square wall element specimens and one specimen representing the wall/basemat junction region were tested.This experimental work indicates that under internal overpressurization or other accident conditions, highly localized strains in the steel liner plate can result in liner tearing and subsequent containment leakage. These results support the theory of leak before break where liner tearing occurs in a controlled manner and leakage and depressurization occur rather than global failure.