Condensation heat transfer with noncondensable gas for passive containment cooling of nuclear reactors

Noncondensable gases that come from the containment and the interaction of cladding and steam during a severe accident deteriorate a passive containment cooling system's performance by degrading the heat transfer capabilities of the condensers in passive containment cooling systems. This work contributes to the area of modeling condensation heat transfer with noncondensable gases in integral facilities. Previously existing correlations and models are for the through-flow of the mixture of steam and the noncondensable gases and this may not be applicable to passive containment cooling systems where there is no clear passage for the steam to escape. This work presents a condensation heat transfer model for the downward cocurrent flow of a steam/air mixture through a condenser tube, taking into account the atypical characteristics of the passive containment cooling system. An empirical model is developed that depends on the inlet conditions, including the mixture Reynolds number and noncondensable gas concentration.     

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.     

An integral model for a severe accident at a nuclear plant with a containment

Translated from Atomnaya Énergiya, Vol. 71, No. 1, pp. 8–13, July, 1991     

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.     

A numerical investigation of three-dimensional flows in large volumes in the context of passive containment cooling in BWRs

The paper describes Computational Fluid Dynamics (CFD) calculations undertaken in support of analyses of three-dimensional flows that take place in the drywell volumes of advanced boiling water reactors with passive decay–heat removal systems. Data for comparison are taken from the 1/40th-scale European Simplified Boiling Water Reactor (ESBWR) mock-up facility PANDA under conditions of symmetric steam injection and asymmetric outflow. Steady-state simulations for pure steam conditions illustrate how the separate flow streams mix to ensure balanced outflow conditions to the condenser units. A transient calculation has also been performed to examine how air, assumed to be released from solution in the PANDA boiler, would ultimately accumulate in the separate condenser units. Results provide a possible explanation for the rundown in performance of one of the condensers, behaviour which was repeatedly observed in some of the earlier PANDA tests. The work also provides more general insights on how trace amounts of non-condensable gases may accumulate in passive cooling equipment.     

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.     

Conceptual design and analysis of a semi-passive containment cooling system for a large concrete containment

An internal evaporator-only (IEO) concept has been developed as a semi-passive containment cooling system for a large dry concrete containment. The function of this system is to keep the containment integrity by maintaining the internal pressure not to exceed ultimate design pressure, i.e. 0.83 MPa (120 psia) in the absence of any other containment cooling following a severe accident, which postulates core damage and hydrogen combustion. The ability of the concept to protect the containment was evaluated for the design basis accident (DBA) large break loss of coolant accident (LB LOCA) and severe accident scenarios (LB LOCA without Emergency Core Cooling System (ECCS) and containment spray flow, 100% zirconium oxidation and complete hydrogen combustion). All were modeled using the GOTHIC computer code. It was concluded that a practical system requiring four IEO loops could be utilized to meet design criteria for severe accident scenarios.     

Variations of a passive safety containment for a BWR with active and passive safety systems

The paper presents variations of a certain passive safety containment for a near future BWR. It is tentatively named Mark S containment in the paper. It uses the operating dome as the upper secondary containment vessel (USCV) to where the pressure of the primary containment vessel (PCV) can be released through the upper vent pipes. One of the merits of the Mark S containment is very low peak pressure at severe accidents without venting the containment atmosphere to the environment. Another merit is the capability to submerge the PCV and the reactor pressure vessel (RPV) above the core level by flooding water from the gravity-driven cooling system (GDCS) pool and the upper pool. The third merit is robustness against external events such as a large commercial airplane crash owing to the reinforced concrete USCV. The Mark S containment is applicable to a large reactor that generates 1830 MW electric power. The paper presents several examples of BWRs that use the Mark S containment. In those examples active safety systems and passive safety systems function independently and constitute in-depth hybrid safety (IDHS). The concept of the IDHS is also presented in the paper.     

Ensuring the long-term functionality of passive auto-catalytic recombiners under operational containment atmosphere conditions—An interdisciplinary investigation

In order to sustain the structural integrity of the containment and other safety relevant components i.e. to avoid a detonation of the hydrogen-air mixture generated during a severe accident in light water reactors, passive auto-catalytic recombiners (PAR) are used for hydrogen removal in many European nuclear power plants (NPP). In 1999, the German NPP Emsland (KKE) was equipped with 58 PAR of AREVA design as an internal accident management measure for a beyond-design accident. Since that time the recombiners are in a stand-by state. As the catalyst elements are exposed to various airborne substances during normal plant operation their function is controlled periodically by testing selected catalyst sheets in a specially designed device. Under the conservative test conditions during this procedure some catalyst sheets showed a delayed responding behavior. First internal analysis gave indication of a beginning fouling on the catalytic surface.The aim of a precautionary investigation performed in cooperation between KKE, Forschungszentrum Juelich and RWTH Aachen University was to characterize the composition of the fouling and to correlate it with potential sources within the containment.In the framework of the investigation the reports of the periodic inspections were analyzed and appropriate sample sheets were selected from the installation. These samples were subjected to a comprehensive chemical surface analysis in order to identify effects like thermal sintering, poisoning or a blocking of the catalytic surface (Baerns, M., 2004. Basic Principles in Applied Catalysis, Springer Verlag). Along with the chemical analysis the catalytic activity of the samples was assessed in several test series in order to correlate the chemically quantified deposition on the catalyst samples with the characteristics of the start-up and the steady-state performance of the recombination reaction. In a final step, possible sources of the fouling were analyzed with regard to their possible contribution to the phenomena. According to the results achieved, measures have been implemented at KKE in order to optimize procedures and to enhance the performance of the PARs.     

Semi-probabilistic fracture mechanics approach for a quality assurance system for welds of a containment for a 1000 MW nuclear power unit

The paper deals with a quality control system based on: (1) limit stage design for a given level of failure probability (P_f≤10⁻⁶), and (2) fracture mechanics requirements for welds, the heat-affected zone (HAZ), and parent material (St 355 E). This quality control system was carried out under on-site-conditions when building a steel-sheet/reinforced concrete composite construction of a containment for a nuclear power station. The basic element of the quality assurance system are the control welds produced simultaneously with the welds on site (manual electrode welding). The materials testing program of such control welds and statistical evaluation of test results are described. The results show a fairly good reproducibility of measured J-integral values of welds gained from on-site-specimens (control welds) and those obtained from pre-tested welding technologies. The objective of this paper is to encourage the application of the proposed semi-probabilistic fracture mechanics approach for cases, when the area of proven experience for designing and fabricating welded metallic constructions must be left. Some conclusions of practical interest are discussed, for instance: (1) restrictions for high-strength steels, resulting from the limitation of yield to ultimate tensile strength ratio (R_p0,2/R_m≤0.75) in the standards and regulations; (2) importance of demand for a portion of plastic component of the J-integral to exclude brittle fracture of welds; and (3) derivation of a fracture toughness criterion for application to high strength steels.     

Hydrogen combustion in nuclear plant accidents and associated containment loads

The objective of this paper is to describe the reactor safety problems in severe accidents due to hydrogen combustion. The generation of hydrogen is discussed. The combustion of hydrogen is considered in terms of diffusion flames, deflagrations, flammability limits, incomplete combustion of very lean deflagrations, detonations and accelerated flames. An example of analyses of hydrogen combustion in a Mark III BWR is given. The example shows the strong dependence of the predicted peak pressure and number of burns on the assumptions of the analyses, the compartmentalization of the model and the inclusion of buoyancy-driven flows.     

Non-iterative model for condensation heat transfer in presence of non-condensable gases inside passive containment cooling vertical tubes

Some contributions have been stated in order to improve the modeling of concurrent downflow condensation in presence of non-condensables inside vertical tubes. In particular, the influence of non-condensables over the liquid side heat transfer has been considered. The new proposed mechanistic models solve explicitly the real interface temperature by means of a cubic or a fourth order equation. As these models have a non-iterative nature, they can avoid the weakest point of the traditional mechanistic models, which is the slowdown computation if the model had to be implemented in a code. Moreover, as the main non-condensables effects can be accounted for in the heat and mass transfer processes, the new models will be more realistic. The models have been validated with the Vierow experimental data, obtaining a total average relative error, for the fourth order equation method model, of 21% for 268 points.     

Failure pressure evaluation of the containment building of a large dry nuclear power plant

The evaluation of the failure pressure of the containment building of a large dry PWR-W three loops nuclear power plant, based on computer numerical simulation, is described in this paper. The proposed method considers fully three-dimensional finite element models in order to take into account the effect of the most significant structural characteristics (presence of three buttresses, penetrations, additional reinforcement around the penetrations, etc.), the lack of symmetry of the forces generated by the prestressing system, as well as the nonlinear behaviour of the materials and the sensitivity of the results to uncertainties associated with several parameters. The computational model is completely described, including the constitutive equations for the concrete, the reinforcing steel and prestressing tendons, the spatial discretization—isoparametric elements including the reinforcement are used. The structural models and the analyses performed for their calibration are also described. The influence on the failure pressure of incorporating the foundation slab in the structural model, and the influence of the thermal effects, are discussed. One of the conclusions of the numerical study is that the failure process can be appropriately simulated by means of a structural model which does not include either the foundation slab or the thermal effects. Finally, results of a probabilistic simulation of the failure pressure are given.     

Different variations of a passive safety containment for a BWR with active and passive safety systems

The paper presents probable variations of passive safety boiling water reactor (BWR). In order to improve safety and economy of passive safety BWR, the authors thought of use of a kind of improved Mark III type containment. The paper presents the basic configuration of the passive safety BWR that has an improved Mark III type containment. We tentatively call this passive safety BWR advanced safer BWR⁺ (ASBWR⁺) and the containment Mark X containment in the paper. One of the merits of the Mark X containment is double containment function against fission products (FP) release. Another merit is very low peak pressure at severe accidents without active cooling systems. The third merit is coolability by natural circulation of outside air. Therefore, the Mark X containment is very suitable for passive safety BWRs. It does not need a reactor building (R/B) as the secondary containment, because it is a double containment by itself. The Mark X containment is a general concept and also useful for half-passive safety BWRs that have both active and passive safety systems. In those examples, active safety systems and passive safety systems function independently and constitute in-depth hybrid safety (IDHS). The concept of the IDHS is also presented in the paper.     

核电站机组故障诊断系统的知识获取

核电站机组故障诊断系统知识的获取和知识库的建立是影响诊断系统能否应用于实际的关键步骤。针对实现核电站机组故障诊断系统给出了知识获取的一种方法和步骤,使其有章可循,加强了在实际核电站中可操作性。按照文中提出的工作框架组织人员完成各项任务,可以最终完成核电站机组故障诊断系统知识库的建立。     

Demonstrative testing of honeycomb passive autocatalytic recombiner for nuclear power plant

Passive autocatalytic recombiners (PAR) are widely being used as hydrogen control device in the current and advanced light water reactors (ALWRs). The PARs lend themselves to very effective means of circumventing buildup of combustible or detonable hydrogen gas mixtures in the reactor containment. Korea Nuclear Technology Inc. has recently developed a new PAR system with high porous catalyst material in the shape of honeycomb. The honeycomb PAR catalyst has a design characteristic of improved hydrogen removal performance by increasing the surface area and enhancing the flow rate through the catalyst at the same time, without increasing PAR size compared to the conventional PARs. The experimental study was focused on the development of the hydrogen depletion rate correlation of the honeycomb PAR. Two different sizes of PARs, KPAR-40 and KPAR-T2, have been employed in the tailor-made Integral Test Facility and Performance Test Facility. Multiple tests were conducted in various conditions of pressure, temperature, and hydrogen concentration. The hydrogen depletion rate correlation and the PAR performance constant were determined from the experimental results, which can be applied to the honeycomb PAR system. Also determined was the scale effect due to the PAR size, i.e., the number of catalysts in a PAR.     

Thermohydrodynamic calculation of the subdome part of the containment envelope of a nuclear power plant

Main Scientific Center of the Russian Federation — Physics and Power Engineering Institute. Translated from Atomnaya énergiya, Vol. 78, No. 2, pp. 172–176, March, 1995.     

非能动先进压水堆核电厂SGTR事故概率安全评价

蒸汽发生器传热管破裂(Steam Generator Tube Rupture,SGTR)事故是核电厂的重要事故之一,并具有其自身的特点。该事故的研究和评价对核电站安全具有较大意义。选取典型非能动先进压水堆核电厂AP1000的SGTR事故进行一级概率安全评价(Probabilistic Safety Assessment,PSA),采用事件树分析方法得到电厂事件发生后系统、设备和人员不同响应所产生的事故序列,然后建立相关系统的故障树模型进行可靠性分析。借助Risk Spectrum软件,计算SGTR事故导致AP1000核电厂的堆芯损伤频率(Core Damage Probability,CDF),并进行堆芯损伤的最小割集分析及重要度和敏感性分析。通过一系列分析得到导致堆芯损伤的重要基本事件,从而找到系统存在的薄弱环节。     

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.