Current forging technology for integrated type steel forgings for nuclear steam supply system components

This brief review concerns the trend in using large size, integrated type steel forgings for nuclear steam supply system components as intended for easier in-service inspection (ISI).To achieve this trend, unique forging techniques have been developed.The forging techniques corresponding to each integrated type steel forging product together with a brief explanation of the development of forging facilities as well as a future aspects are introduced.     

Leak behavior of SCC degraded steam generator tubings of nuclear power plant

A forced outage due to a steam generator tube leak in a Korean nuclear power plant has been reported [Kim, J.S., Hwang, S.S., et al., 1999. KAERI Internal Report (Korean). Destructive analysis on pulled tubes from Ulchin unit 1. Korea Atomic Energy Research Institute]. Primary water stress corrosion cracking has occurred in many tubes in the plant, and they were repaired using sleeves or plugs. In order to develop proper repair criteria, it is necessary to understand the leak behavior of the tubes containing stress corrosion cracks. Cracked specimens were prepared using a room temperature cracking technique, and the leak rates and burst pressures of the degraded tubes were determined both at room temperature and at a high temperature. Some tubes with 100% through wall cracks did not show a leakage at 10.8 MPa, which is the typical pressure difference of the pressurized water reactors (PWRs) during a normal operation. In some tests, the leak rates of the tubes increased with time at a constant pressure. In a high temperature pressure test at 282 °C one specimen showed a very small leakage at 18.6 MPa, which stopped after a small increase in the test pressure. Because stress corrosion cracks can develop at relatively low stresses, even 100% through wall cracks can be so tight that they will not leak at a normal operating pressure.     

Mathematical model of impurities hide out and return in nuclear power plant steam generator

This paper discusses a simple mathematical–physical model for describing the dissolved impurities hide out and return phenomenon in the boiling channel, within the viscous sub layer. By solving the equation of convection-diffusion for the liquid phase near the heating surface and the bulk stream, the effect of heat flux in distribution of impurities is considered. With deriving the impurities mass balance based on the presented model and using the impurities mass balance in the viscous sub layer as the boundary conditions, the derived differential equations are solved by the finite volume (upwind) method.     

Droplets deposition in steam piping connecting steam generator and steam turbine in nuclear plant

A numerical investigation is carried out for turbulent droplet-laden flow of saturated steam produced in a steam generator (SG) that feeds steam turbine (ST) through a long and multi-bend steam piping. The main purpose of the study is to analyze deposition of droplets that form a wall film in the piping system. Commercial CFD code StarCD is used for the solution of turbulent flow field of droplet-laden steam. Turbulence is treated using k–ω model of turbulence. Wall film formation is solved by additional conservation equations. Two tasks were performed: parametric study of the deposition in a 90° elbow positioned with different orientation and the deposition in a more complex piping system. This system starts with outlets from steam generator with five mouthpieces leading to a collector pipe and connecting the steam piping leading to a steam turbine. The steam piping consists of three straight segments of pipes and two 90° elbows in the total length 17 m. The diameter of the steam piping is 0.425 m. Results of the simulations show where droplets deposit and where a liquid separator should be placed to drain away the water film and to avoid droplets from entering the steam turbine.     

Prediction of crack coalescence of steam generator tubes in nuclear power plants

Prediction of failure pressures of cracked steam generator tubes of nuclear power plants is an important ingredient in scheduling inspection and repair of tubes. Prediction is usually based on nondestructive evaluation (NDE) of cracks. NDE often reveals two neighboring cracks. If the cracks interact, the tube pressure under which the ligament between the two cracks fails could be much lower than the critical burst pressure of an individual equivalent crack. The ability to accurately predict the ligament failure pressure, called “coalescence pressure,” is important. The failure criterion was established by nonlinear finite element model (FEM) analyses of coalescence of two 100% through-wall collinear cracks. The ligament failure is precipitated by local instability of the ligament under plane strain conditions. As a result of this local instability, the ligament thickness in the radial direction decreases abruptly with pressure. Good correlation of FEM analysis results with experimental data obtained at Argonne National Laboratory’s Energy Technology Division demonstrated that nonlinear FEM analyses are capable of predicting the coalescence pressure accurately for 100% through-wall cracks. This failure criterion and FEA work have been extended to axial cracks of varying ligament width, crack length, and cases where cracks are offset by axial or circumferential ligaments.     

1000MW核电管板纯净钢锻件制造工艺及其性能

根据1000MW核电管板用纯净钢锻件性能和组织的要求,利用合金化原理确定了冶炼时钢中各合金元素的成分控制方向;采用电炉加钢包炉加真空浇注进行冶炼浇注,真空浇注过程中加保护防止二次氧化;采用特殊的镦粗工艺避免了管板镦粗过程心部产生超标缺陷;采用合理的热处理工艺,保证管板锻件的组织和性能。经检验表明,锻件用钢的质量达到了纯净钢的要求,管板锻件的综合性能达到世界领先水平。     

Application of 20 MnMoNi 5 5 steel with lowered Si content to heavy thick steam generator tube sheet forgings

The 20 MnMoNi 5 5 steel with a lowered Si content, in which the addition of Al for grain refining is made before vacuum degassing in the steel making process, was applied to the heavy thick tube sheet forging for the steam generator of a light water reactor. The properties of the forging were compared with a forging of the conventional Si/Al deoxidized steel. The steel with lowered Si content showed less A- and V-segregations than the Si/Al deoxidized steel. And also the steel with lowered Si content has improved impact properties and ductility as well as a greater isotropy of properties. The excelent weldability was confirmed by the tangential sectioning test. The less segregated 20 MnMoNi 5 5 steel with lowered Si content, which has improved material properties such as impact properties and weldability, assures a better integrity of the component and easier fabrication.     

Numerical model for estimation of corrosion location in nuclear power plant steam generators

Deposition of dissolved impurities and corrosion in steam generators is a significant problem in the operation of nuclear power plants. Impurities and corrosion products usually accumulate in the secondary sides of steam generators (SG) and form deposits on the SG surfaces. A high level of impurity concentration close to the SG heating surface causes the corrosion process to occur with more intensity. The aim of this study is to estimate the most probable locations of impurity concentration and deposition in a SG. Equations representing the convection and diffusion in the liquid phase close to the heated surface (the viscous sub layer) are derived. Based on the mass balance of impurities in the viscous sub layer as the boundary condition, the derived differential equations are solved by the finite volume (upwind) methods. The distribution of impurities, sediment formation rate and the location of the depositions in the viscous sub layer at different heat flux values are studied in steady and unsteady states.     

基于CFNN的核蒸汽发生器水位控制

鉴于常规的PID控制存在控制对象参数变化时控制参数无法改变的不足,从而根据一个核蒸汽发生器(NSG)的简化数学模型,将一种补偿模糊神经网络(CFNN)用于NSG水位的控制。该网络由于引入了补偿神经元,使网络的容错性更好,系统更稳定。同时在神经网络学习算法中动态优化补偿模糊运算,使网络更适应,训练速度更快。仿真表明,该方法在装置负荷变化时比常规的PID控制方法超调量小,收敛速度快。该网络能在线调整参数,动态优化模糊规则,适于在线学习控制。该控制方法对NSG水位智能控制研究具有一定意义。     

Manufacturing of ultra-large diameter 20 MnMoNi 5 5 steel forgings for reactor pressure vessels and their properties

As the structural material for RPV typical of increased dimensions, a set of ultra-large diameter steel forgings for a PHWR RPV is presented as outlined below.

1. (1) Material designation: 20 MnMoNi 5 5 (similar to SA508, Cl.3)

2. (2) Size of the forgings: flanges, 8.440 mm OD, a weight of 238 tons for shell flange; shells and torus, 7,920 mm OD, with large height; cover dome, 6,800 mm OD in chord and 460 mm thick; blank before formed to dome is ca. 8,000 mm OD.

3. (3) Chemical composition: particular effort was made for minimizing the tramp elements as P, S, As, Sn, Sb, Cu.

4. (4) Manufacturing, key points: steel making - combined refining and degassing in ladle; ingot making - largest size ingots, including 570 ton and 500 ton ingots; forging - special “outside-the-press” forging and forming techniques; heat treatment - prevention of H₂ flaking in normalizing and tempering and handling of the extra-large forgings at water quenching.

5. (5) Metallurgical properties: sufficiently uniform carbon distributions in the forgings; a lowest possible content of hydrogen, non-metallic inclusions and oxygen.

Mechanical properties: uniformity in tensile and toughness properties; flaws - only limited number of spots of UT indications under 2 mm EFG (EFS).     

II. Steady state thermal measurements on the secondary side of a PWR steam generator (BUGEY-4 nuclear power plant)

Steady state thermal measurements, water and shell temperature, are made in the downcomer of the steam generator of the PWR BUGEY-4. The measurements show that the flow is of a swirling type with the degree of swirl being a function of the load (power). The angular temperature distribution of water and shell exhibits a minimum that rotates from the hot leg to the cold leg with increasing load. The results also admit the conclusion that the feedwater and recirculation water are well mixed in the cylindrical section of the downcomer. Measurements and computations of fluid and shell temperatures are in good agreement for the steam generator in a state of natural circulation. For the state of forced circulation, the agreement is poor when it is assumed that the flow in the downcomer is turbulent and the steam generator outer shell is well insulated. The agreement is excellent for turbulent flow but with air leakage, or infiltration, between the steam generator outer shell and its insulation. If the hypothesis of air infiltration is ruled out, agreement between measurements and computations is much improved when the flow in the downcomer is taken as a turbulent core flow with an attendant laminar boundary layer. The existence of a laminar boundary layer would require the flat plate transition Reynolds number be enhanced by a factor of 100. This enhancement could result from the combined effects of swirl and density gradient in the downcomer.     

Application of high strength MnMoNi steel to pressure vessels for nuclear power plant

Recent increase in output of nuclear power plant has been attained by enlargement of major components such as pressure vessels. Such large components have almost reached a size limit from the points of manufacturing capacity and cost in both forgemasters and fabricaters. In order to solve this problem, it must be beneficial to apply design by use of material of higher strength, which brings reduction of pressure vessel thickness and weight. The Japan Steel Works Ltd. (JSW) has many manufacturing experiences of large integrated forgings made from high strength MnMoNi steel with tensile strength level of 620 MPa for steam generator (SG) pressure vessel, and has performed confirmation tests of its material properties. This paper describes the confirmation test results such as tensile and impact properties, nil-ductility transition temperature (NDT-T), static and dynamic fracture toughness, weldability including under-clad cracking (UCC) sensitivity, as well as metallurgical factors which influence on such material properties.     

III. PWR steam generator response to a scram at 50% load and an open grid transient (bugey-4 nuclear power plant)

Two transients, an open grid and a scram at 50% load, were conducted on unit 4 of the PWR power plant Bugey. The thermal hydraulic response of the steam generator was recorded. For the open grid test, the following observations are noted:No alarming phenomena are observed in the steam generator during the transient. Primary pressure oscillations were very mild, and did not exceed about 4.8 bar/min with a maximum amplitude of ±8 bar. This condition should not result in significant stress levels. Steam generator outer shell metal temperature gradients remained within very acceptable limits; a maximum amplitude of about +13°C and a rate not exceeding 0.8°C/min are obtained. This slow rate is explained by a fall in primary water temperature that allows for a temperature decrease inside the U-tube bundle. Similarly, the temperature rise on the tube sheet does not exceed an amplitude of 20°C with a rate of about 2°C/min. Again these conditions do not lead to any significant thermal shock on the tube sheet. The steam generator feed controls maintain the level within the normal operation range and the small addition of colder feedwater does not lead to great temperature changes because of the large mass of the recirculation water in the steam generator.For the scram at 50% load, the following observations are noted: no severe thermal or pressure transients are observed in this test. Fluid temperature fluctuations occur with rates not exceeding 1°C/s and a maximum amplitude of about 20°C in the downcomer and 10°C on the tube sheet. Steam generator outer shell temperature varies at a rate of about ±0.8°C/min with a maximum amplitude of about 16°C. These thermal transients should lead to thermally induced stresses of acceptable levels.     

Flow-induced vibration of nuclear steam generator U-tubes in two-phase flow

Flow-induced vibration characteristics of a U-tube bundle were experimentally investigated in air-water two-phase flow. The test section was equipped with 39 U-tubes, simulating the innermost region of an actual steam generator. The U-tubes were made of Inconel 690 with a diameter of 19.05 mm. The horizontal region of the U-tubes had a rotated square array with a pitch of 31.11 mm and a p/d of 1.633. The U-tubes and supporting structures have almost the same prototypical geometries. Vibration responses of six U-tubes were measured with ten 3-axis accelerometers. Two sets of experiments were performed to investigate an onset of fluid-elastic instability, damping ratio, and hydrodynamic mass of the U-tubes. The experiments were performed for a void fraction of 70-95%. The instability constant (K) of the Connors’ equation for the present U-tube bundle was evaluated to be in the range of 6.5-10.5.     

Development of thick wall 2.25Cr-1MoNiNb steel forgings for steam generators of fast breeder reactors

The large and thick forgings made of 2.25Cr-1MoNiNb steel are required for vessel material of steam generator (SG) of the fast breeder reactor SNR 300.In order to study the feasibility of 2.25Cr-1MoNiNb steel heavy section forging, chemical composition, melting practice and ingot making, hot working and heat treatment conditions were investigated. The following recommendations were derived: (1) 0.04% C, 0.10% ΔNb, (2) application of electro-slag remelting process, (3) grain refining by hot working, (4) two-step austenitizing at 1020°C.Based on these recommendations, the actual products such as hollow cylinders with maximum 290 mm thickness, solid bars with 420 mm diameter and forged plates with maximum 185 mm thickness could be supplied for application in the helical coiled SG of the SNR 300. Statistical analysis of the products showed the sufficient and isotropic material properties, which fulfill the requirements of the basic safety rules.