蒸汽发生器泄漏率监测

基于主蒸汽管道中＾１６Ｎ核素监测的蒸汽发生器泄漏率监测方法,目前已成为蒸汽发生器泄漏监测的主导方法。本文主要介绍了蒸汽发生器泄漏率的监测原理、蒸汽发生器泄漏率与＾１６Ｎγ计数率的换算关系以及实际应用中存在的问题,并简要介绍了秦山和大亚湾核电站年用该类监测系统的主要特性。     

Improvement of steam generator tube failure propagation analysis code LEAP for evaluation of overheating rupture

As a part of safety assessment or design of steam generators of sodium-cooled fast reactors, it is necessary to evaluate the water leak rate under sodium–water reaction accident. The computer code called LEAP-II calculating a design basis water leak rate during long-time event progress including self-wastage, target-wastage, wastage-type failure propagation, water leak detection, and water/steam blowdown was developed for the prototype fast reactor in the past studies. In this study, a numerical analysis method to predict occurrence of overheating tube rupture was constructed and integrated into this code to expand its application range. The newly constructed method consists of the elemental analysis models for temperature distribution formed by a reacting jet, water-side thermal hydraulics, heat transfer at the tube wall, temperature and stress of the tube, and failure of the tube. Applicability of the method was investigated through the numerical analysis of the experiment on water vapor discharging into liquid sodium pool under the actual condition of the steam generator. The numerical analysis demonstrated that the method could provide the appropriately conservative result on the overheating-rupture-type failure propagation.     

传热管泄漏对蒸汽发生器动态特性的影响

以大亚湾核电站蒸汽发生器为研究对象,建立了基于漂移流理论的蒸汽发生器一维动态数学模型及传热管泄漏模型,并进行了蒸汽发生器不同工况下的稳态仿真。在验证所建立漂移流模型和传热管泄漏模型的基础上,研究了不同工况下传热管泄漏位置及泄漏流量对蒸汽发生器关键参数的影响。研究结果表明,所建立的漂移流模型和传热管泄漏模型能准确反映不同泄漏情况下蒸汽发生器质量含汽率及蒸汽压力等关键参数的变化规律,泄漏发生在热端沸腾段入口处时各参数变化最显著,泄漏量为冷却剂流量的5%时出口质量含汽率由0.261降到0.163。基于漂移流理论传热管泄漏对蒸汽发生器动态特性影响的成功预测,为蒸汽发生器传热管泄漏事故的监测与防范措施的制定提供一定参考。     

Modeling of reinforced concrete by the non-local microplane model

This paper reviews some recent results of non-local finite element fracture analysis of concrete structures using a non-local microplane material model. The microplane model and recently introduced non-local microcrack interaction approach are described briefly. It is demonstrated that the model used in smeared fracture finite element analysis does not exhibit mesh sensitivity. Results of a three-dimensional numerical study of fastening elements pulled out from cracked and uncracked reinforced concrete plates are shown. The capability of the model for correct prediction of the structural size effect is supported by one numerical example. In all examples numerical results are compared with experimental evidence and reasonably good agreement is observed.     

Leak rate calculation for LBB analysis in high temperature gas-cooled reactors

Leak rate calculation is very important for Leak Before Break (LBB) analysis. Helium is used as coolant in high temperature gas-cooled reactor (HTGR). Therefore the flows in the cracks of HTGR vessels and pipes are single phase, which are different from the two phase critical flows in the cracks of water reactors. In the present paper, simple leak rate calculation formulae for compressible laminar and turbulent flows in HTGR cracks are introduced. The velocity and pressure distributions in cracks as well as the leak rates are calculated using the formulae. Numerical simulations are also conducted for compressible laminar, turbulent and critical flows with different crack widths and depths. The results of the numerical simulation and theoretical formulae are compared with experimental data. The comparison shows that both the simple theoretical formulae and the numerical simulation can achieve good results.     

Detecting small leaks in pipes in nuclear power plants by measuring aerosol parameters

It is found experimentally that when through defects appear in water pipes the concentration and dispersion composition of the aerosols in the room air change. A water or steam leak in a pipe at an early stage of the development of a defect, when the leak still has no effect on the standard operation of the equipment, can be detected by monitoring the aerosol component of the air. The stage where a leak can be detected by aerosol monitoring depends on the background aerosol concentration in the room air, the temperature and pressure in the pipe, and the arrangement of samplers in the monitoring system. The leak-detection method proposed is much more sensitive than moisture content measurements.Translated from Atomnaya Énergiya, Vol. 97, No. 3, pp. 189–195, September, 2004.     

Study of leaktightness integrity of containment wall without liner in high performance concrete under accidental conditions—I. Experimentation

When a partially saturated concrete wall is subjected to accidental conditions (high temperature and steam water pressure, as a LOCA or more severe conditions), water vapour penetrates the containment wall until saturation level of the containment atmosphere is achieved. The rate of penetration of water vapour through concrete is progressively reduced, leading to improvement of the leaktightness integrity of the concrete wall. In this paper, experimental studies involving the measurement of temperature, moisture propagation and pore pressures in a concrete containment wall are presented. The tests have been carried out on cylindrical specimens, made of high performance concrete (HPC) and having 1.3 m thickness (same thickness as a containment wall of a nuclear power plant). A finite element analysis is used to study the heat and mass transfer through the concrete wall. The results of this numerical modelling technique are presented in the second part of this study.     

Response of gas-cooled fast breeder reactors to depressurization accidents

The gas-cooled fast breeder reactor (GCFR) under design by Gulf General Atomic is cooled with helium pressurized to 85 atm and has the reactor core, the steam generators and their associated steam turbine-driven helium circulators, and auxiliary core cooling loops all contained within a massive prestressed concrete reactor vessel (PCRV).The response of the GCFR to coolant depressurization accidents has been investigated and it has been shown that this class of accidents can be safely handled with considerable safety margin. Rapid depressurization is assumed to be caused by a seal failure in a large concrete plug closing one of the large PCRV cavities and the depressurization rate is controlled by a flow restrictor incorporated within the closure plug. Continued core cooling is provided by the main core cooling loops. The plant transient reponse following a depressurization accident has been calculated with a computer code developed at GGA. The results obtained indicate rather mild increases in peak clad temperature for a depressurization accident with the leak area defined by the flow restrictor.Additional cases investigating larger leak areas to explore safety margins indicate that the peak cladding temperature does not increase rapidly with increasing leak area. Secondary containment conditions in a depressurization accident have also been evaluated.     

Special features of steam–gas flow through a crack in the protective envelope of a nuclear power plant

Experimental data on the escape of steam and liquid drops with the outflow of a steam–air mixture through a model of a crack in a concrete protective envelope are presented. The experiments are performed using cold and heated concrete assemblies. Complete condensation of the steam on the surface of a crack occurs at relatively low temperatures of the concrete assembly. As the concrete assembly is heated, drops appear in the gas flow at the exit.Translated from Atomnaya Énergiya, Vol. 97, No. 5, pp. 333–338, November, 2004.     

Essential results of analyses accompanying the leak rate experiments E22 at HDR

During phase III of the HDR Safety Programme (HDR: decommissioned overheated steam reactor in Karlstein, Germany), experiments were performed in test group E22 on small-bore austenitic straight piping and on pipe elbows and branches containing through-wall cracks. The main aim was the determination of crack opening and leak rate behaviour for the cracked components under almost operational pressure and temperature loading conditions, especially including transient bending moments. In addition to machined slits, naturally grown fatigue cracks were also considered to cover the leakage behaviour. The experiments were accompanied by calculations, mainly performed by GRS. The paper describes the most important aspects and the essential results of the calculations and analysis. The main outcome was that the crack opening and initiation of crack growth can be described with the finite element techniques applied with sufficient accuracy. However, the qualification of the leak rate models could not be completed succssfully, and therefore more sophisticated experiments of this kind are needed.     

Small Leak Detection by Measuring Surface Oscillation during Sodium-Water Reaction in Steam Generator

Small leak sodium-water reaction tests were conducted to develop various kinds of leak detectors for the sodium-heated steam generator in FBR. The super-heated steam was injected into sodium in a reaction vessel having a sodium free surface, simulating the steam generator. The level gauge in the reaction vessel generated the most reliable signal among detectors, as long as the leak rates were relatively high. The level gauge signal was estimated to be the sodium surface oscillation caused by hydrogen bubbles produced in sodium-water reaction.

Experimental correlation was derived, predicting the amplitude as a function of leak rate, hydrogen dissolution ratio, bubble rise velocity and other parameters concerned, assuming that the surface oscillation is in proportion to the gas hold-up. The noise amplitude under normal operation without water leak was increased with sodium flow rate and found to be well correlated with Froud number. These two correlations predict that a water leak in a “MONJU” class (300 MWe) steam generator could possibly be detected by level gauges at a leak rate above 2g/sec.     

Investigation on the interaction effect of two parallel axial through-wall cracks existing in steam generator tube

It is commonly required that steam generator tubes wall-thinned in excess of 40% should be plugged. However, the plugging criterion is known to be too conservative for some locations and types of defects and its application is confined to a single crack. In the previous study, the conservatism of the present plugging criterion of steam generator tubes was reviewed and a crack coalescence model applicable to steam generator tubes with two collinear axial through-wall cracks was proposed. Since parallel axial cracks are more frequently detected during in-service inspections than collinear axial cracks, the studies on parallel axial cracks spaced in circumferential direction are necessary. The objective of this paper is to investigate the interaction effect between two parallel axial through-wall cracks existing in a steam generator tube. Finite element analyses were performed and a new failure model of the steam generator tube with these types of cracks is suggested. Interaction effects between two adjacent cracks were investigated to explain the deformation behavior of cracked tubes.     

Analysis of reinforced concrete containment vessels considering concrete cracking

Cracking of concrete influences the stress analysis of concrete containment vessels. If cracking is ignored, the resulting shell analysis can be unconservative in some cases and extremely conservative in others. A cracked concrete shell is a structurally orthotropic one. That is, it does not have the same properties in membrane action and bending action. Closed form equations are presented for cracked concrete shells using the split rigidity concept. The equations cover symmetrically loaded cylindrical shells, effects of concentrated forces and moments on spherical shells, and effects of openings and concentrated forces and moments on cylindrical shells. In addition, methods are discussed that can be applied to cracked concrete shells by using finite element techniques.     

快堆蒸汽发生器小泄漏下三维流场数值模拟

为了计算快堆蒸汽发生器小泄漏后钠水反应产物的传输扩散,对快堆壳管式蒸汽发生器内稳态钠流场进行了三维数值模拟,建立了蒸汽发生器内钠流场三维数值模型。管束对钠流的影响用分布阻力、体积多孔率、面渗透率来模拟,并根据横掠管束与纵掠管束的压力阻力关系式来关联管束的分布阻力。采用κ－ε湍流模型,壳和支撑板壁面采和壁面函数法来处理。模型计算压降与实验数据比较,二者吻合较好。     

Screening Test for Adsorbent for Kr at Liquid Argon Temperature under Nitrogen Atmosphere

A new acoustic leak detection system for sodium-cooled reactor steam generators using a delay-andsum beamformer is proposed. The major advantage of the delay-and-sum beamformer is that it could provide information on the acoustic source direction. An acoustic source of a sodium-water reaction is supposed to be localized, while the background noise of the steam generator operation is uniformly distributed in the steam generator tube region. Therefore, the delay-and-sum beamformer could distinguish the acoustic source of the sodium-water reaction from the steam generator background noise. In this paper, results of numerical analyses are provided to show the fundamental feasibility of the new method.     

Air–steam leakage through cracks in concrete walls

In the context of a severe accident in a PWR nuclear plant, the evaluation of the leakage through the containment wall remains a key point of the safety analysis. Here we calculate the leakage of an air steam mixture through a traversing crack taking into account condensation. A 40 h test has been performed on a representative concrete slab with measurements of crack openings and flow rates. The CAST3M code enables us to simulate this test by making thermo-mechanical calculations and calculation of the leakage flow rate. Thermo-mechanical calculations provide data needed by the leakage calculations which are not measurable in the experiment. These are the internal crack profiles (variation of the opening with the curvilinear coordinate of the crack inside the concrete slab). Thermo-mechanical calculations are difficult to perform because boundary conditions of the test are complicated. Leakage calculations are performed with various hypotheses for the internal cracks profiles. A coefficient is applied on the friction factor to take into account additional complexity of the crack geometry.     

Nonlinear fluid-structure interaction calculation of the leakage behaviour of cracked concrete walls

In the case of severe accidents in nuclear power plants, containments are the last barrier to prevent the release of environmentally hazardous substances. Therefore, the leaktightness of the containment is of decisive importance for the safety and protection of the environment in case of an accident. A numerical model based on the Finite Element Method has been developed to calculate the leakage behaviour of reinforced concrete walls. Leakage flow and structural response are solved iteratively. For the calculation of the leakage flow a fluid model has been used which takes into account the condensation of the steam part within the air-steam mixture. Both, the release of the latent heat in the case of condensation and the following two-phase flow of air and water have been considered, too. Tests with the SIMIBE Experimental facility [Caroli, C., Coulon, N., Renson, C., 1995. Steam leakage through concrete cracks: parametric study with SIMIBE experiment and interpretation of the results. Tech. Rep. Commissariat A L’Energie Atomique (CEA)] are used for verification of the condensation and two-phase flow models.     

Leak area and leakage rate prediction for probabilistic risk assessment of concrete containments under severe core conditions

The fundamental gap in knowledge for estimating release for probabilistic risk assessment of concrete containments subject to beyond design basis loads is in estimating leak areas and leakage rates. By evaluating the available literature and carefully studying the test results, several generic rules are postulated for leak areas and leakage rates of concrete containments. These rules, coupled with theory-based leakage flow equations and empirically-based crack roughness constants, provide a realistic estimate of leak rates through liner tears in concrete containments.     

旋叶汽水分离器试验和数值模拟研究

针对旋叶汽水分离器的缩比模型展开空气-水冷态试验性能分析和数值模拟研究。试验表明：分离效率主要受水流量的影响,随水流量的增大呈逐渐上升的趋势,当水流量增大到0.3 m³/h后增加趋势逐渐减缓进而出现下降的趋势。压降对空气流量变化敏感,随空气流量的增大显著上升。进一步建立旋叶分离器CFD数值分析模型,采用欧拉两流体模型,气相为连续相,液相为离散相,并使用雷诺应力RSM方法求解湍流应力。计算分析了液滴粒径对分离特性的影响,结果表明,液滴粒径分布对分离效率有显著影响,微小粒径液滴的存在显著降低了气液分离效率。