Welding overlay analysis of dissimilar metal weld cracking of feedwater nozzle

Inspection of the weld between the feedwater nozzle and the safe end at one Taiwan BWR showed axial indications in the Alloy 182 weld. The indication was sufficiently deep that continued operation could not be justified considering the crack growth for one cycle. A weld overlay was decided to implement for restoring the structural margin. This study reviews the cracking cases of feedwater nozzle welds in other nuclear plants, and reports the lesson learned in the engineering project of this weld overlay repair. The overlay design, the FCG calculation and the stress analysis by FEM are presented to confirm that the Code Case structural margins are met. The evaluations of the effect of weld shrinkage on the attached feedwater piping are also included. A number of challenges encountered in the engineering and analysis period are proposed for future study.     

Contribution to understanding of stress corrosion cracking of Alloy 600 in PWR primary water

Abstract

The stress corrosion cracking mechanism of Alloy 600 in pressurised water reactor (PWR) primary water has been investigated. U-bend specimens (extracted from tubular products) were tested in simulated PWR primary water containing H₂¹⁸O. After testing, crack tips was characterised by nanosecondary ion mass spectrometry (nanoSIMS) and unstressed areas were analysed by standard SIMS. The results allowed oxygen (¹⁶O, ¹⁸O) penetration into the material to be evaluated and the oxide present at crack tips to be characterised. Experimental data were compared with results obtained by calculations performed according to Fickian diffusion laws and Whipple Le Claire theory.     

Experimental investigation on the cracking of pre-combustion cracking gas with gliding arc discharge plasma

To improve the quality of pre-combustion cracking gas, the gliding arc discharge plasma is adopted in this paper. The influence of incoming flow rate, electrode gap distance and discharge frequency on cracking effect have been experimentally studied. The results show that through plasma cracking, the concentration of H₂ increases while the concentration of CH₄ and C₂H₄ decrease. With different fuel ratio conditions, the variation trend of C/H ratio as well as the mass flow of carbon and hydrogen atoms in the cracking products vary from each other. The cracking effect becomes weaker with the increase of incoming flow rate, while is better when the fuel ratio is larger. Given the flow rate of 18slpm and 24slpm, the best cracking effect can be acquired with the electrode gap of 1.5 mm. Besides, the effect of gliding arc discharge plasma cracking is weakened as the discharge frequency rises.     

Preliminary Investigation of Full Model of Two—Mode Scramjet

IntroductionThe latest Studies of the ovens characteristics ofhypersonic fifor vehicles with ~ engineshave shown that these vehicles are fairly prondsing.Some addihonal. PbolelnS arise, how~ that arerelated tO the dedrihon Of the geneal shape of theaircraft, the choice Of Propulsion type and operationmpbo, the enghe size and POsition on the body.The ~ version under consideration boltesthe use of a COmbined PtOPulsion act with a twO-modecombustor (subsonic or supersonic combustion). sucha…     

The impact of fouling on performance evaluation of multi-zone feedwater heaters

A numerical method for analyzing closed system feedwater heaters is presented. A general approach to determine area allocations among the desuperheating, condensing and subcooling zones under a known set of operating conditions is presented for a feedwater heater in a steam power plant. A significant amount of heat duty is handled by the condensing zone, whereas the subcooling zone handles a least amount of heat duty which essentially vanishes at low steam pressures. Fluids mass flow rates and accordingly the overall heat transfer coefficients have significant effects on the areas needed for desuperheating, condensing and subcooling in a feedwater heater. Two fouling models are considered to examine their effect on the heat exchanger performance. Insignificant changes were noticed when comparing the heat transfer rate and outlet temperature results of both the models. It is found that heat duty of the heat exchanger decreases by 2.7% in 3 years when we use the recommended fouling resistance, while the outlet shell-side fluid temperature increased by 6.3%.     

功能表面降膜蒸发传热特性的实验研究

研究了处理表面镀铬铝管、PTFE铜管和纯铝氧化管水平管降膜蒸发传热，研究了喷淋密度、热流密度、管内蒸汽速度和管表面处理对降膜蒸发传热特性的影响。实验结果表明：在表面蒸发区，水平管降膜蒸发传热系数随热流密度的增加而提高，随喷淋密度增大先降低后升高，冷凝例传热系数基本保持不变。总传热系数对操作条件变化很不明显，表面阳极氧化膜使传热系数略有下降，但由于其优良的抗垢时蚀性能，非常有必要再进行深入地研究。     

Preliminary investigation of factors determining self-organised structures preparation in polymer layers

Factors determining creation of self-organized structures, Bénar–Marangoni cells, during the process of solvent evaporation from the polymer solution and formation of polymer film were studied. Examined parameters were temperature, temperature gradient, rate of drying, height of a liquid layer, area for film preparation, viscosity, molecular weight distribution, etc. A special apparatus, micro condensation drying system, was engineered for this study. As a model system, hydroxyethyl cellulose aqueous solution was used for its excellent film-forming ability and the tendency to self-aggregation. Experimental results, presented in a wide spectrum of self-organized patterns, show the complexity of the problem and the crucial role of molecular weight distribution of the polymer in the fixation of organized structures under highly non-equilibrium conditions.     

Cracking in feedwater pipework of light water reactors: Causes and remedies

The present paper deals with cracking in ferritic pipework of light water reactor (BWR) feedwater systems whose causes are unsuitable design and manufacture and, at least occasionally, unsuitable water quality. Thus in two BWR plants in the German Federal Republic large circumferential cracks have been found in the circumferential welds in the main feedwater pipelines immediately adjacent to the reactor pressure vessel, and in a further BWR large longitudinal cracks have been found in pipe bends of the reactor water purification pipework connected to the main feedwater pipelines.The piping regions near the reactor pressure vessel feedwater nozzles represent a boundary region of varying thermodynamic states of the pressurised water. The reactor pressure vessel is heated to saturation temperature by the radioactive decay heat, and then cooler water, and sometimes (during start-up) cold water is injected into the reactor pressure vessel in order to maintain a constant water level about 2 m above the upper edge of the feedwater nozzles. In order to improve the state of knowledge regarding the stressing conditions under prevailing operating conditions, extensive strain and temperature measurements have been carried out. The results of these measurements carried out in several BWRs confirm the occurrence of rapid temperature changes in the feedwater pipework in the regions where it connects to the reactor pressure vessel, leading to varying stresses which at times reach the plastic region. These processes are triggered by special operating states such as start-up and shut-down or hot standby operation with feedwater flows smaller than 6% relative to normal operation under full load.     

Life duration of PWR nuclear power plants

The importance of the Pressurized Water Reactor (PWR) nuclear plants in the whole of the electrical production facilities of EDF is a good reason why a special effort is devoted to assessing their life duration. It seems that the longevity of the PWRs is going to show more favourable technical and economical conditions than did the previous steam generating plants.

The EDF ‘life duration project’ includes a four-year task program. First, a full inventory of the available data will be drawn up in order to forecast the ageing of the sensitive materials of a power plant and to follow up its development so as to anticipate or to remedy problems. These sensitive materials are those for which the difficulties and the cost of restoration would not allow the inclusion of those operations within the normal framework of maintenance programs.

The first analysis of how much these materials may be damaged leads one to believe that the main results of such ageing, which occur mainly in reactor vessels as embrittlement due to exposure and also in the fatigue of the main components of the boilers, will not limit the life of these sensitive materials, except in the very long term.

Overcoming the effects of certain ageing or materials corrosion damage may well give difficult problems to maintenance so that it is vital to forecast and assess them. If this necessary effort is undertaken, the lifetime of the PWR units should be very long.     

Assessment of boiler tubes overheating mechanisms during a postulated loss of feedwater accident

Due to their simplicity and passive nature, natural circulation loops have many industrial applications and being increasingly used in many innovative designs as the new generation of nuclear reactor cooling systems. Consequently, special attention is increasingly considered towards their safety issues. In the current framework, a safety aspect related to the loss of feedwater in an industrial D-type steam boiler is assessed. Indeed, loss of feedwater event occurs frequently during steam generator facilities lifetime. Under such conditions, the system integrity is in jeopardy and serious hazards and economic losses could occur. Up to now, specific numerical models have been used to simulate the thermal-hydraulic phenomena occurring under such circumstances. Such models are generally based upon simplified assumptions, and if complex model are performed, their applications are generally restricted to the facility they have been developed for. In the current framework, an attempt is made to apply a multipurpose best estimate (BE) thermal-hydraulic system code, namely, RELAP5/Mod3.2 to simulate steady-state and transient dynamic behaviors of a two-phase natural circulation steam boiler. For this purpose, two loss of feedwater scenarios, with and without the actuation of the boiler control system, have been performed. The analysis results show on one hand the achievement in applying RELAP5 for a conventional boiler system, and on the other hand the efficiency of the control systems to mitigate the accident consequences.     

高压给水加热器人孔密封结构两种形式的探讨

高压加热器作为火力发电机组中回热系统的关键设备,其水室承受水侧高压力,对水室检修人孔的密封性能要求很高。本文主要对高压加热器水室检修人孔椭圆自密封和伍德密封两种形式进行探讨。     

Experimental investigation of two-phase closed thermosyphon used in heavy duty extrusion pelleting line

Efficient cooling system is an essential part of heavy duty extrusion pelleting line which plays an important role in production engineering of megaton polyolefin. In this paper, a new cooling system based on two-phase closed thermosyphon (TPCT) used in heavy duty extrusion pelleting line was presented. Comparative experimental results show that thermal performance of TPCT is more efficient, and the temperature uniformity is much better than traditional extruder (similar to coiled heat exchanger) in preheating process and extrusion reaction process. The effects of different operation conditions: filling ratio (0.2, 0.35, 0.5, 0.65 and 0.8), flow rate of cooling water (120 L/h, 180 L/h, 240 L/h, 300 L/h and 360 L/h) and heating power (7 kW, 9 kW and 11 kW) on thermal characteristics were experimental investigated, respectively. The results show that temperature of barrel inner wall increased significantly as filling ratio increased and evaporator appeared dry out when filling ratio is 0.2. The flow rate of cooling water affected the condenser section obviously, but had little influence on evaporation section. With increase of heating power, the start-up time decreases and the heat transfer coefficient increased. An ideal cooling scheme was concluded: the liquid filling ratio was 0.35, the cooling water flow rate was 180 L/h and the heating power was 11 kW when working medium was water.     

On thermal annealing of irradiated PWR pressure vessels

Radiation embrittlement in some pressurised water reactors has been so fast that, in spite of other applied mitigation methods, thermal annealing has been practically the only solution permitting further operation. The annealings have been reported to be successful and resulted in no damage. In cases where the whole fuel core zone area of the reactor pressure vessel has to be annealed, a fully successful annealing has yet to be convincingly proven. High thermal stresses may make the thermal treatment troublesome to carry out. The recovery mechanisms have been difficult to study because of the exceptionally small size of irradiation defects. The degree of recovery cannot be yet fully calculated precisely from material and annealing information. Especially the recovery processes in restoring elastic and elastic-plastic fracture toughness properties needs much more work to be carried out in order to be clarified. Recovery annealing at a proper temperature is, however, a very effective method and, in many cases, practically the only alternative for extending the service life of a pressure vessel embrittled by radiation. There seems to be no restrictions to repeating the thermal treatment, but every pressure vessel should be independently studied and assessed for the achievement of safe results.     

History of the PWR and its worldwide development

Pressurized water reactors and the US company Westinghouse are sometimes seen as synonymous with automobiles and Ford, or computers and IBM. It is certainly true that PWRs of Westinghouse design have a dominant position in the world market for nuclear power plants, but the situation is not directly comparable for a number of reasons. First, a nuclear power plant is not a clearly defined product and, even with a high degree of standardization of the reactor, it calls for a considerable amount of individualized engineering design to incorporate it into the overall complex of a power station and to meet the licensing requirements of different countries. Moreover, while Westinghouse has played the leading role, many others organizations in different countries have contributed significantly to the development of the PWR from its earliest days and are playing an increasing role in its further development and exploitation. In addition, the hardware content of a nuclear power plant, including the reactor, incorporates components from many different manufacturers. What can be said, however, is that the basic configuration of the Westinghouse design of PWR is now so widely adopted that there is a unique worldwide fund of design, manufacture, construction and operating experience.     

锅炉给水自动控制系统的故障处理一例

针对长广电厂#3机组负荷扰动下，锅炉给水自动系统稳定时间过长，受“虚假水位”干扰严重的故障，分析原因，采取措施，解决了问题。     

Controlled cracking of tubes

A method is described of inducing an internal, longitudinal fatigue crack in a tube. It is based on a local compression of the tube, using a special loading device, to initiate and propagate a fatigue crack. It is shown that the geometry of this crack can be varied according to loading conditions. Some examples of small as well as of through cracks are given.     

Preliminary investigation on hydrogen-rich gas production by co-steam-reforming of biomass and crude glycerin

Hydrogen production from renewable biomass is of great interests. Co-steam-reforming of biomass and crude glycerin with the ratio of 1:1(w/w) was investigated in a fixed-bed gasifier aiming at improving biomass to hydrogen conversion, focusing on the effects of temperature, pressure, H₂O/C ratio and Ca/C ratio on producing gas composition. The results show that high temperature and low pressure favors hydrogen production. With temperature increasing from 650 to 825 °C, H₂ yield shows a linear increase from 0.053 mol/kg to 0.059 mol/kg. Both H₂ yield and its mole fraction increase obviously with the increasing H₂O/C ratio. No influence on gas production is found for Ca/C ratio > 1. The study reveals the optimum condition for producing hydrogen is: temperature of 700-750 °C, pressure of 0.1 MPa, H₂O/C ratio of 1.7-2.25 and Ca/C ratio of 1. Our experimental analysis shows co-steam-reforming of biomass and crude glycerin for hydrogen production is feasible and promising.     

Increasing capacity of a nuclear power plant unit using the hydrogen-fueled feedwater heating system

The article considers a novel approach to a challenging issue of modern power systems related with the future of nuclear power energy as having clear environmental advantages. To solve the problem, a number of alternative ways of using hydrogen fuel as a source of clean energy were proposed, which are aimed to increase the capacity and maneuverability of operating and designed two-circuit nuclear power plants (NPPs) with installed water coolants. In particular, it was suggested to use hydrogen energy to improve the performance of steam generators and capacity of NPP units by heating the feedwater. The revealed economic benefits of the given approach include the effect from replacing environmentally harmful stations based on gas turbine units. A research was made into the impact of pressure in the hydrogen-oxygen combustion chamber on the effectiveness of the proposed solutions. The necessary design parameters to the hydrogen heating system, the gross and net power ratings of the power unit, and the net present value were determined. It is shown that lowering the pressure in the hydrogen-oxygen combustion chamber leads to the increase in the net present value by 24 to 71 million dollars and by 43 to 90 million dollars for the operating and designed NPPs, respectively. Meanwhile, using the proposed approach to the designed NPPs can ensure a higher capacity for the power units, which equals 179 and 163 MW at atmospheric and elevated pressures, respectively. The efficiency of off-peak energy may equal 26.25 to 27.21 and 26.53% to 27.55% for the operating and designed NPPs, respectively. It should be noted that the proposed schematics prove to be economically competitive across the marginal cost of off-peak energy. The accrued net present value equals 106 to 404 million dollars and 170 to 468 million dollars at the elevated pressure, and 130 to 475 million dollars and 213 to 558 million dollars at the atmospheric pressure for the operating and designed NPPs, respectively.