Energy Technology Data Exchange (ETDEWEB)

The water and steam circuits of steam generators in pressurized-water nuclear power plants are described together with the mechanism of denting, and the corrosion of spacer plates that leads to cracks in tubes by constriction. The different chemical specifications applicable to the water of the secondary circuit of the generators in normal operation and on first commissioning are listed. The results obtained and the measurements of chemical values taken in operation on the water in the secondary circuits of steam generators at Fessenheim and Bugey are presented.

International Nuclear Information System (INIS)

A survey was conducted of experience with steam generator tubes at nuclear power stations during 1976. Failures were reported at 25 out of 68 water-cooled reactors. The causes of these failures and the repair and inspection procedures designed to cope with them are summarized. Examination of the data indicates that corrosion was the major cause of steam generator tube failures. Improvements are needed in steam generator design, condenser integrity and secondary water chemistry control. (author).

International Nuclear Information System (INIS)

The manual presents the fundamentals of thermodynamics and fluid mechanics, the main components of steam power plants, and the power generation process. The following concepts and subjects are discussed at length: steam generator; steam turbines; turbogenerators; condensers; cooling technology; water/steam cycle and water treatment; design data of fossil-fuelled power plants; design and optimisation of nuclear power plant thermodynamics; pipelines and fittings; control systems in steam power plants; connection to the electricity grid and self-supply of thermal power plants; power plant transformer concepts and definitions. (HAG).

International Nuclear Information System (INIS)

Purpose: To provide a feedwater control device capable of minimizing the adverse response of steam drum level at low power. Consitution: In order to perform feedwater control at low power by the substantial control of three factors, that is, main steam flow rate, feedwater flow rate and steam drum level, the main steam flow rate is determined from the reactor output and feedwater rate is determined from the changes in the feedwater temperature due to the mixing of waters in the reactor clean up system and feedwater. If a difference is resulted between these flow rates, a starting feedwater regulator is controlled instantly to eliminate the difference. The water level in the steam drum is used for amending the difference from the final set value of the drum water level, by which the adverse response of the ...

International Nuclear Information System (INIS)

In this paper the thermal design of a horizontal steam generator for the Indian PHWR nuclear power plant is described. The main attraction is absence of tube sheet and use of stainless steel 'U' tubes. It is emphasised that with appropriate water chemistry it is possible to use stainless steel tubes, which is many times cheaper than the Incoloy tubes used elsewhere. The design approach, applicable equation for the design and the results of computation in the form of heat transfer area and some important dimensions of the steam generator are presented.

International Nuclear Information System (INIS)

The Steam Generator (SG) tubing degradation caused by corrosion and other age-related mechanisms continues to be a significant safety and cost concern for many Nuclear Power Plants (NPP). The understanding of the steam generator ageing mechanisms is the key to effective management of steam generator ageing and consists of the knowledge of steam generator materials and these one properties, stressors and operating conditions, like degradation sites and wear mechanisms. The principal types of corrosion are presented which can occur in CANDU steam generator. There are also presented the operation conditions, the specifications referring to the water chemistry and the construction materials of Steam Generator, the factors that have a great influence on the corrosion behaviour during the whole exploitation period of this ...

International Nuclear Information System (INIS)

Design considerations are given for the secondary side region of a vertical U-tube nuclear stream generator with an integral preheater. The thermal shield design, the novel recirculating water flow distribution scheme, the high porosity tube supports used in the parallel flow regions, and the U-bend supports are discussed for the Wolsung Plant steam generators. Experimental and analytical development programs undertaken to verify the design features are outlined.

Energy Technology Data Exchange (ETDEWEB)

A submerged nuclear power station has been proposed as an alternative power station with a relatively low environmental impact for use by both industrialized and developing countries. The station would be placed 10 m above the seabed at a depth of 30--100 m and a distance of 10--30 km from shore. The submerged nuclear power station would be manufactured and refueled in a central facility, thus gaining the economies of factoryfabrication and the flexibility of short-lead-time deployment. To minimize the size of the submerged hull, horizontal steam generators are proposed for the primary-to-secondary heat transfer, instead of the more traditional vertical steam generators. The horizontal steam generators for SNPS would be similar in design to the horizontal steam generators used in the N-Reactors except the tube orientation is horizontal (the tube`s inlet and ...

International Nuclear Information System (INIS)

Results of investigation of stress corrosion cracking of steam turbine materials in nuclear, fossil and geothermal power plants have been analysed. The role of factors that cause damage to rotor discs, mono block and welding rotors of steam turbines has been shown. These are yield stress and steel composition, stress intensity coefficient and crack growth rate, composition and temperature of the condensed steam and water, electrochemical conditions. The conclusion has been made about the state of stress corrosion cracking of the rotors materials, and main investigation trends which are necessary to solve this problem have been listed.

International Nuclear Information System (INIS)

The mechanism of the salt concentration increase in pits and crevices formed in a steam generator due to its imperfect manufacture or to its design features is described. The probability of corrosion can be reduced by choosing a suitable steel and by securing low concentrations of salts (chlorides in particular) and corrosion products in the feedwater. Attention is paid to the distribution of salts in the water-steam circuit and to the conditions of erosion corrosion as the principal source of corrosion products in feedwater. Experience with the suppression of erosion corrosion at nuclear power plants abroad is described. (E.J.).

Energy Technology Data Exchange (ETDEWEB)

Since the first operation of nuclear steam generator early 1960s, its performance requirements have been met but the steam generator problems have been met but the steam generator problems have been major cause of reducing the operational reliability, plant safety and availability. U-tube assembly of steam generator forms the primary system pressure boundary of the plant and have experienced several types of tube degradation problems. Tube failure and leakage resulting from the degradation will cause radioactive contamination of secondary system by the primary coolant, and this may lead to unplanned plant outages and costly repair operations such as tube plugging or steam generator replacement. For the case of steam generators for heavy water reactors, e.g. Wolsong 2, 3, and 4 NPP, a high cost of heavy ...

International Nuclear Information System (INIS)

In equipment industries, the equipments handling industrial water and pure water are numerous. In power generation including nuclear power generation, water serves as a working medium. Review is made on the experiences in the corrosion of iron and nickel base alloys in high temperature, high pressure water and the results of researches derived from them. Under high temperature and high pressure, carbon steel, low alloy steel, stainless steel and high nickel alloy cause corrosion even in pure water. But in the case of serious corrosion, chlorine, oxygen, alkali and others in water take part. The following matters are described: corrosion by steam; stress corrosion cracking in pure water; corrosion by impurities in high temperature, high pressure pure water, i.e. chlorine ions, ...

International Nuclear Information System (INIS)

Nuclear Engineering Laboratory of the Technical Research Centre of Finland has widely used RELAP5/MOD2 and -MOD3 codes to simulate horizontal steam generators. Several models have been developed and successfully used in the VVER-safety analysis. Nevertheless, the models developed have included only rather few nodes in the steam generator secondary side. The secondary side has normally been divided into about 10 to 15 nodes. Since the secondary side at the steam generators of VVER-440 type reactors consists of a rather large water pool, these models were only roughly capable to predict secondary side flows. The paper describes an attempt to use RELAP5/MOD3 code to predict secondary side flows in a steam generator of a VVER-440 reactor. A 2D/3D model has been developed using RELAP5/MOD3 codes cross-flow junctions. The model includes 90 volumes on the ...

International Nuclear Information System (INIS)

Special T-shaped feedwater distribution pipes were installed in steam generators at the Loviisa (Finland) and Rovno (Russia) nuclear power plants. The new shape was tested in an extensive testing programme. Since the tubes frequently suffer from corrosion damage, large-scale water hammer experiments were performed on a model facility in 1996. The main objectives of the water hammer experiments were to find out the prevailing parameters leading to water hammers, as well as the sensitivity of hammering to boundary conditions. A water hammer may occur when the mass flow rate into the steam generator exceeds 6 kg/s and the temperature difference between steam generator and feedwater exceeds 100 degC. Visual experiments and stress analyses of the pipe were also carried out. The weakest part, the T-joint, may hold against such ...

Energy Technology Data Exchange (ETDEWEB)

AREVA is the world leader in the design and construction of nuclear power plants, the manufacture of heavy components, and the supply of nuclear fuel and nuclear services such as maintenance and inspection. The Equipment Division provides the widest range of nuclear components and equipment, manufactured at its two facilities in Jeumont, northern France, and St. Marcel, in Burgundy. The St. Marcel plant, set on 35 ha (87.5 acres) near Chalon-sur-Saone, was established in 1973 in a region with a long history of specialized metalworking and mechanical activities to meet the demand for non-military nuclear requirements in France. The site offers two advantages: - excellent facilities for loading and transporting heavy components on the Saone river, - it's proximity to other group sites. Since its completion in 1975, the Chalon/St. Marcel facility has manufactured all the heavy ...

International Nuclear Information System (INIS)

Problem areas in BWRs, PWRs and PHWRs, from the viewpoint of chemistry, and the problem of fission product release in nuclear reactors are discussed. These problem areas are : fuel performance, off-normal water chemistry due to condenser leaks, the transport and deposition of the activated corrosion and fission products, denting in steam generators (in the case of PWRs), ingress of air in the cover gas helium and consequent radiolysis of D_2O in the moderator circuit (in the case of PHWRs). (M.G.B.).

British Library Electronic Table of Contents (United Kingdom)

The effect of the water-chemical regime (WCR) on damage sustained by heating surfaces of steam generators at NPP is analyzed. It is indicated that phosphate treatment with minimal excesses of phosphates in the steamgenerator water is the most optimal method of managing the WCR regime of horizontal steam generators.

International Nuclear Information System (INIS)

The present process allows to eliminate the corrosion products formed on the tube plates and in the interstices of plate-tube crosspieces of a PWR steam generator in order to avoid a corrosion phenomenon which may cause denting by presence of oxides. The process consists in applying on these oxides at about 50-100 degrees, an aqueous solution containing 6-8% of gluconic acid, 3-5% of citric acid, about 0.5% of a corrosion inhibitor and ammonia until a pH of 3-9.5 is obtained.

International Nuclear Information System (INIS)

Imatran Voima Oy , (IVO) operates two Russian designed nuclear power plants of type VVER440/213. Unit 1 has been operating since 1977 and unit 2 since 1981. First damage of feed water distribution (FWD) pipes was observed in 1989. In closer examinations FWD-pipe T-connection and distribution nozzles suffered from severe erosion corrosion damage. Similar damages have been found also in other VVER-440 type NPPs. In 1994 the first FWD-pipe was replaced by a new design mounted over the tube bundle instead of the old FWD-pipe, which was located inside the tube bundle. The purpose of this paper is to describe the new FWD-pipe and discuss its effects on the steam generator chemistry. (author)

Energy Technology Data Exchange (ETDEWEB)

The chemical operational concept originally established for the water-steam circuit of Angra II nuclear power plant has undergone several modifications throughout the development of the project. This work discusses the two main modifications giving special attention to the costs involved and analyses the main points and the consequences of such modifications 1 ref., 4 figs., 2 tabs.

International Nuclear Information System (INIS)

Three-dimensional computer simulation and analyses of the horizontal steam generator thermal-hydraulics of the WWER 1000 nuclear power plant have been performed for 50% and 75% partial loads, 100% nominal load and 110% over-load. Presented results show water and steam mass flow rate vectors, steam void fraction spatial distribution, recirculation zones, swell level position, water mass inventory on the shell side, and other important thermal-hydraulic parameters. The simulations have been performed with the computer code 3D ANA, based on the 'two-fluid' model approach. Steam-water interface transport processes, as well as tube bundle flow resistance, energy transfer, and steam generation within tube bundles are modelled with "closure laws". Applied approach implies non-equilibrium thermal and flow conditions. The model ...

International Nuclear Information System (INIS)

Hydrogen is a potential alternative energy source and produced commercially by methane (natural gas) or LPG steam reforming, a process that requires high temperatures, which are produced by burning fossil fuels. However, since this process emits large amounts of CO_2, replacement of the combustion heat source with a nuclear heat source for 773-1173 K processes has been proposed in order to eliminate these CO_2 emissions. This paper proposes a novel method of low-temperature nuclear hydrogen production by reforming dimethyl ether (DME) with steam produced by a low-temperature nuclear reactor at about 573 K. The authors identified conditions that provide high hydrogen production fraction at low pressure and temperatures of about 523-573 K. By setting this low-temperature hydrogen production process at about 573K upstream from a turbine, it was found theoretically that the total energy ...

Science.gov (United States)

Measurement of local steam condensation rates of cocurrent stratified flow of steam and subcooled water was carried out at atmospheric pressure in a horizontal rectangular channel. The channel was constructed of stainless steel with pyrex glass windows, a...

International Nuclear Information System (INIS)

Solid solution-strengthened nickel base alloys have been used for nuclear stream generator tubing in pressurized water reactors since the beginnings of commercial nuclear power. The purpose of this paper is to recap and update the authors understanding of the relationship between processing, resulting structure, and properties for Alloy 600 and to discuss the requirements for optimized performance in both primary and secondary environments. Potential replacement materials and their performance will be discussed. Also discussed is the role and importance of system chemistry, bulk and local, and control and its relationship to performance. A discussion of potential mechanisms of environmentally assisted failure is also discussed.

Science.gov (United States)

Results are given of numerical simulation of three-dimensional pattern of flow of a two-phase steam-water mixture in the house of a PGV-1000 horizontal steam generator obtained using the BAGIRA best-estimate thermohydrodynamic computer codes. The space distributions of velocities and local void fractions in the steam generator housing for different modes of operation of power-generating unit are calculated and compared with available experimental data.

International Nuclear Information System (INIS)

Steam generators of four VVER-440 units at nuclear power plants V-1 and V-2 in Jaslovske Bohunice (Slovakia) were gradually changed by new original 'Bohunice' design in period 1994-1998. Corrosion processes before and after these design and material changes in Bohunice secondary circuit were studied using Moessbauer spectroscopy during last 25 years. Innovations in the feed water pipeline design as well as material composition improvements were evaluated positively. Moessbauer spectroscopy studies of phase composition of corrosion products were performed on real specimens scrapped from water pipelines or in form of filter deposits. The corrosion of new feed water pipelines system (from austenitic steel) in combination to innovated operation regimes goes dominantly to magnetite. The hematite presence is mostly on the internal surface of steam generator body and ...

Energy Technology Data Exchange (ETDEWEB)

A model was developed to compute concentration of impurities in the WWER 440 steam generator (SG) secondary water along the tube bundle. Calculated values were verified by concentration values obtained from secondary water sample chemical analysis. (orig.). 2 refs.

International Nuclear Information System (INIS)

The detection, in European steam generators, of intergranular stress corrosion cracking initiating from the primary side (PWSCC) and the somewhat similar detection of intergranular stress corrosion (IGSCC) and intergranular corrosion (IGA) initiated from the secondary side in Japanese steam generators has led to a growing awareness of the potential for such corrosion forms in United States PWR steam generators. What had been a minor occurrence at several units is now a cause of concern and repair measures at more than 30 sites in the United States, and in some cases, is the major cause of steam generator unavailability. The United States nuclear utilities and EPRI formed the first Steam Generator Owners' Group (SGOG) in 1977 in response to the prospect of continued denting corrosion. The occasional appearance of PWSCC, IGSCC and IGA in domestic ...

International Nuclear Information System (INIS)

Nuclear grade production tubing of Alloy 600 was evaluated for stress corrosion cracking (SCC) susceptibility in high purity water at 365, 345, 325, and 290 C. Reverse tube U-bend specimens provided crack initiation data and constant extension rate tests were employed to determine the crack velocities experienced in th crack propagation stage. Initial results indicate that a linear extrapolation of data received from high temperature tests can be used to predict the service life of steam generator tubing that has been plastically deformed or is continually deforming by ''denting.''.

Energy Technology Data Exchange (ETDEWEB)

Large scale steam storage power generation, one of the new energy storage systems for the future of inflexible electric power sources consisting of nuclear and coal power plants has been studied on the subjects of the systems to be attached to coal and nuclear power units, of the definition of storage efficiency and of the vertical steam storage vessel technology. Steam storage power generation may be hopeful for its higher efficiency similarly defined as of pumped storage plants while high temperature heat storage and the internal structure of large vertical steam storage vessel (accumulator) need to be developed.

Energy Technology Data Exchange (ETDEWEB)

The steam generator low water level trip setpoint of KORI NPP units 3 and 4(KNU 3 and 4), three-loop Westinghouse pressurized water reactor, is higher than that of OPR1000. In addition, steam generator downcomer water level in KNU 3 and 4 could fluctuate easily during a transient because of smaller downcomer water inventory, compared to the total water inventory in the steam generator. Due to these reasons, there is a higher possibility of unnecessary reactor trips caused by the steam generator low-low water level in KNU 3 and 4. Its operating history shows that most of reactor trips were caused by steam generator low-low level reactor trip signal. Such reactor trips, especially unnecessary ones, result in time and economic losses. In this paper, a ...

Science.gov (United States)

The main steam piping of nuclear power plant which runs between steam generator and high pressure turbine has been experienced to have a severe effect on the safe operation of the plant due to the vibration induced by the steam flowing inside the piping. ...

International Nuclear Information System (INIS)

This is a manuscript for a lecture contents: 1) Steam power and fundamentals of the steam power process, 3) conventional, nuclear and other steam generation processes, 4) cooling systems for steam power plants, 5) gas turbine power plants and combined-cycle power plants, 6) cogeneration, 7) development of thermal power plants and environmental effects. (GL).

International Nuclear Information System (INIS)

PWR plants have made efforts to maintain the long-term integrity of the steam generators (SG) by reducing the amount of corrosion products entering the secondary side of the SG. Iron entered the SG can cause several problems: degraded heat conductivity of the SG tubes in locations where iron is deposited, water level oscillations in the SG due to tube support plate hole blockage, and initiation and propagation of inter-granular attacks (IGA) and stress corrosion cracking (SCC). One of the most effective measures, high all-volatile treatment (AVT) chemistry has been applied to actual plants to reduce the flow-accelerated corrosion (FAC) coming from the carbon steel piping. The secondary water chemistry at Genkai NPS 1 and 2 changed, from the Low AVT chemistry to the High AVT chemistry, in November 2006. In this paper, we will describe the results of experiments in applying the use of High pH water in the ...

Energy Technology Data Exchange (ETDEWEB)

A reflux condensing start-up system includes a steam generator, a start-up vessel connected parallel to the steam generator, a main steam line connecting steam outlets of the steam generator and start-up vessel to a steam turbine, a condenser connected to an outlet of the turbine and a feedwater return line connected between the condenser and inlets of the steam generator and start-up vessel. The start-up vessel has one or more heaters at the bottom thereof for heating feedwater which is supplied over a start-up line to the start-up vessel. Steam is thus generated to pressurize the steam generator before the steam generator is supplied with a heat transfer medium, for example liquid sodium, in the case of a liquid metal fast breeder reactor. The start-up vessel includes upper and ...

International Nuclear Information System (INIS)

... hydrology norway water pollution pollution scandinavia surface waters

International Nuclear Information System (INIS)

By August 2007, there were 438 nuclear power plants (NPPs) in operation worldwide, with a total capacity of 371.7 GW(e). Further, 31 units, totaling 24.1 GW(e), were under construction. During 2006 nuclear power produced 2659.7 billion kWh of electricity, which was 15.2% of the world's total. The vast majority of these plants use water-cooled reactors. Based on information provided by its Member States, the IAEA projects that nuclear power will grow significantly, producing between 2760 and 2810 billion kWh annually by 2010, between 3120 and 3840 billion kWh annually by 2020, and between 3325 and 5040 billion kWh annually by 2030. There are several reasons for these rising expectations for nuclear power: - Nuclear power's lengthening experience and good performance: The industry now has more than 12 000 reactor years of experience, and the global average ...

International Nuclear Information System (INIS)

By August 2007, there were 438 nuclear power plants (NPPs) in operation worldwide, with a total capacity of 371.7 GW(e). Further, 31 units, totaling 24.1 GW(e), were under construction. During 2006 nuclear power produced 2659.7 billion kWh of electricity, which was 15.2% of the world's total. The vast majority of these plants use water-cooled reactors. Based on information provided by its Member States, the IAEA projects that nuclear power will grow significantly, producing between 2760 and 2810 billion kWh annually by 2010, between 3120 and 3840 billion kWh annually by 2020, and between 3325 and 5040 billion kWh annually by 2030. There are several reasons for these rising expectations for nuclear power: - Nuclear power's lengthening experience and good performance: The industry now has more than 12 000 reactor years of experience, and the global average ...

Science.gov (United States)

This 1978 annual report contains the results of ''Seismic testing of Wolsung-1 steam generator model'' which was initiated in 1977 as a part of a study on nuclear components testing. Model 78, improved version of Model 77 which did not take into account f...

Science.gov (United States)

In the study three loss-of-feedwater type experiments which were preformed with the PACTEL facility has been calculated with two computer codes. The purpose of the experiments was to gain information about the behaviour of horizontal steam generator in a ...

International Nuclear Information System (INIS)

General corrosion behavior of ALLOY 800 in high temperature water was studied in relation to its surface film structure. The surface film formed in water was found to decrease the corrosion rate of ALLOY 800. This film is composed of Ni ferrite, and can be obtained by oxidation in air or steam. Based on these results, air or steam oxidation treatment to inhibit Ni and Co release of ALLOY 800 into high temperature water is proposed. (author).

Science.gov (United States)

Eight years ago, corrosion and tube denting seriously threatened the reliability and design life of steam generators, especially for closed loop arrangements in pressurized water reactors (PWRs). Concentrated research by the Steam Generator Owners Group (SGOG) diagnosed the causes and produced effective solutions, notably guidelines for water chemistry control in the secondary loop. The guidelines recommend specific levels of water impurities and remedial actions to prevent cooling-water leaks in the condenser, prevent air leaks, limit corrosion product buildup, and remove some impurities while neutralizing others. Continued research in SGOB-II is investigating intergranular corrosion and stress corrosion cracking. 3 figures.

Energy Technology Data Exchange (ETDEWEB)

The restricted temperature stability of the synthetic thermal oil which is used as heat carrier fluid in parabolic trough collectors so far limits the live steam parameters in the steam cycle to approximately 375 Celsius. In order to break through this limit, already for quite some time it is researched to replace the thermal oil by boiler feeding waters and to accomplish the evaporation in the collectors. The contribution under consideration gives an overview on the direct evaporation concept and summarizes the past operational experiences. Moreover, the challenges with the integration of this technology in a steam turbine cycle are elaborated.

International Nuclear Information System (INIS)

This paper reports on steam injection cycle which is similar to supplementary fired combined cycle, but for the utilized steam medium produced by HRSG, its temperature is higher and pressure is lower than in the combined cycle. In comparison with the thermodynamic advantage of the two cycles, a clear understanding of physical concept can be gotten simply by energy balance. The difference of total power output between them is subtraction of enthalpy difference of exhaust steam and feed water of HRSG in steam injection cycle from the rejected heat by water coolant of condenser in combined cycle, when using the identical gas turbine and the same amount of total fuel consumption. In general case, formulas and data are given to indicate this comparison by the ratio of steam mass flow supplied by HRSG of the two cycles. The analysis of Cheng Cycle ...

International Nuclear Information System (INIS)

No information on steam/water two-phase flow behavior in large diameter pipes (10 inch or larger) at elevated pressures is available in the open literature. However, there are many applications, in the nuclear, chemical and petroleum industries among others where two-phase flows in large diameter pipes at elevated pressures and temperatures are encountered routinely or under accident scenarios. Experimental data on steam-water two-phase flow in a large diameter (20 inch, 50.08 cm I.D.) vertical pipe at elevated pressures and temperatures (2.8 MPa/230 C--6.4 MPa/280 C) have been obtained. Void fraction, two-phase mass flux, phase and velocity distributions as well as pressure drop along the test pipe have been measured using the Ontario Hydro Technologies (OHT) Pump Test Loop. The void fraction distributions were found to be axially symmetric and nearly flat over a wide range of two-phase flow conditions. The two-phase flow ...

International Nuclear Information System (INIS)

TVA is currently engaged in an extensive program to mitigate steam generator tube denting at the Sequoyah Nuclear Plant, which is in commercial operation, and to prevent or minimize the onset of denting at the Watts Bar Nuclear Plant, which is under construction. This paper describes TVA's denting mitigation program, which is primarily feedwater chemistry and system operation improvement, and the effect the changes that have been implemented have had on the incidence as well as the progression rate of steam generator tube denting during the past 220 effective full power days of Sequoyah unit 1 operation.

International Nuclear Information System (INIS)

Site studies for CANDU type NPP began in a careful manner since 1982 as a first part of the Nuclear Power Plant Romanian Program adopted by political and governmental authorities at the time. A team was charged to develop all packages of the necessary main studies. The first Romanian NPP CANDU 6 type reactor gone to erection on Cernavoda site, planned to have 5 units and, like Wolsong NPP, applied the same design for the nuclear island. For the BOP parts the ANSALDO-GE project was applied with a thorough concern about requirements raised by connection to NSP. The first mission of design and research multi-branch team was to adapt the NPP Cernavoda project having an open water cooling circuit 'once-through' to the new parameters of a close recirculation water cooling circuit. Also, the structural design was re-evaluated for the case of soft foundation strata instead of hard rock ones. The close ...

Energy Technology Data Exchange (ETDEWEB)

We defined the Korean unique safety-critical software development methodology by modifying Dr. Harel`s statechart-based on formal methods in order to digitalized the reactor protection system. It is suggested software requirement specification guideline to specify design specification which is basis for requirement specification and automatic programming by the caused by shutdown parameter logic of the steam generator water level for Wolsung 2/3/4 unit SDS no.1 and simulated it by binding the Graphic User Interface (GUI). We generated the K and R C code automatically by utilizing the Statemate MAGNUM Sharpshooter/C code generator. Auto-generated K and R C code is machine independent code and has high productivity, quality and provability. The following are the summaries of major research and development. - Set up the Korean unique safety-critical software development methodology - Developed software requirement specification guidelines - ...

Energy Technology Data Exchange (ETDEWEB)

This document presents the analysis of the simulation of the loss of coolant accident at uprate power conditions, that is 2027 MWt (105% of the current rated power of 1931MWt). This power was reached allowing an increase in the turbine steam flow rate without changing the steam dome pressure value at its rated conditions (1020 psiaJ. There are also presented the results of the simulation of the main steam isolation va/ve transient at overpressure conditions 1065 psia and 1067 MWt), for Laguna Verde Nuclear Power Station. Both simulations were performed with the best estimate computer code TRA C BF1. The results obtained in the loss of coolant accident show that the emergency core coolant systems can recover the water level in the core before fuel temperature increases excessively, and that the peak pressure reached in the drywell is always below its design pressure. Therefore it is ...

International Nuclear Information System (INIS)

A century of technical development put steam turbines on a high level regarding efficiency and reliability. This procedure is still ongoing. The technological-commercial point of view - influenced intensively by liberalisation of the energy-market - makes great demands on field services. Well suited concepts in service and modernization are the solutions, as shown in NPP Emsland upgrade.

International Nuclear Information System (INIS)

The investigations carried out on Alloy 800 in aqueous high-temperature environments in France as well as in other countries are reviewed. These studies are mainly concerned with nuclear industry where Alloy 800 can be used as structural material for steam generators of PWR, breeders or HTR. As results referred to in the literature on cracking in caustic environmens do not always agree, a discussion is presented on the matter. The behaviour of Alloy 800 in superheated steam is examined. (Auth.).

Energy Technology Data Exchange (ETDEWEB)

Fundamental phenomena of condensation heat transfer at a steam-water interface have been studied related to the thermo-hydrodynamics of the emergency core cooling system for light water reactors. In this study temperature fluctuations near the interface and in the liquid phase were measured using fine thermocouples for a saturated steam-subcooled water co-current stratified two-phase flow in a nearly horizontal rectangular channel, and heat transfer coefficients were determined experimentally. The values of the condensation heat transfer coefficients in this experiment are from 6 to 40 kW/m{sup 2}K. In the regions of high Reynolds numbers, as the steam Reynolds numbers become larger, the average interfacial heat transfer coefficients tend to increase. The corelations of Nusselt numbers were obtained from the heat transfer data. (author)

International Nuclear Information System (INIS)

Based on the occurrence of a number of plant incidents during low power and shutdown operating conditions, the Nuclear Regulatory Commission (NRC) has initiated several programs to better quantify risk during these periods. One specific issue of interest is the loss of residual heat removal (RHR) under reduced coolant inventory conditions. This issue is also of interest in the Federal Republic of Germany and an experiment was performed in the integral PKL-3 experimental facility at Siemens-KWU to supply applicable data. Recently, an effort has been undertaken at the Idaho National Engineering Laboratory (INEL) to identify and analyze the important thermal-hydraulic phenomena in pressurized water reactors following loss of vital AC power and consequent loss of the RHR system during reduced inventory operation. The thermal-hydraulic response of a nuclear steam supply system (NSSS) with a closed reactor ...

International Nuclear Information System (INIS)

A mathematical and physical model was constructed describing the hydrodynamics of the two-phase mixture in the horizontal steam generator. The HP 9830 A desk-top calculator was used for the computations. The output variable of the solution was the level shape. A quantitative and qualitative comparison was made of the results of computations and experimental data. (author).

Science.gov (United States)

We present the results obtained from tests and studies carried out on the model of tube bundles for a PGV-1000 horizontal steam generator that were conducted for experimentally substantiating the design service life of a steam generator tube bundle intended for use at new nuclear power stations equipped with a PGV-1000MKP steam generator. Measures taken to minimize the incipience and development of local corrosion damage to the heat-transfer tubes and ensure their design service life are substantiated and confirmed.

International Nuclear Information System (INIS)

The use of pre-dehumidified air having widely established itself as a safe and economical conservation method during the stoppage of steam turbines, it was obvious that this method would be applied also to steam boilers and ancillary aggregates. Experience with the conservation of condensers, prheaters and the water/steam-side of steam boilers in large-scale power stations is reported on. By the example of projects with topical interest, the problems and possibilities of boiler conservation on the firing side are illustrated. The contribution is to stimulate further discussions. (orig.).

International Nuclear Information System (INIS)

Steam generators (SG) are among the most critical components of pressurized water Nuclear Power Plants (NPP). SG tubes must provide a reliable pressure boundary between the primary and secondary cooling water. It is because that any leakage from tube defects could result in the release of radioactivity to the environment. Thus degradations of steam generators tubes should be monitored and inspected periodically under nuclear regulatory. In-service inspections of SG tubes are carried out using eddy current test (ECT) and the defected tubes are usually plugged. Because the radioactivity in the internal of SG chambers limits free access of human worker, remote manipulators are required. In South Korea, Manipulators such as the Zetec SM series and the Westinghouse ROSA series have been used. Such manipulators are rigidly mounted to manways or tube sheets of SG. ...

International Nuclear Information System (INIS)

The objective of the C-E/EPRI project, ''Alternative Steam Generator Materials and Designs,'' was to evaluate the corrosion behavior of contemporary or alternative steam generator materials under prototypic design and secondary fault (high contaminant) water conditions. Two model steam generators built with various support materials and designs were tested under representative thermal and hydraulic conditions. One model operated under seawater faulted all-volatile treatment (AVT) secondary water chemistry conditions. The other model operated under acidified fresh water faulted AVT conditions. This presentation focuses on the tube support and tubesheet corrosion results obtained by destructive examination of both models.

International Nuclear Information System (INIS)

The nuclear power industry has many applications where electrical equipment is required to operate in environments which contain high relative humidity, steam and water or chemical sprays. Equipment is susceptible to these environments by moisture entering through inherent design provisions such as O-rings and through the electrical interface. The objective of this research is to review electrical interface sealing techniques and to recommend methods for providing effective electrical interface sealing. The most common methods of sealing are the use of electrical fittings containing sealants, add-on assemblies and hardware changeout. The experience gained in several years of testing electrical interface seals on such devices as limit switches, solenoid valves and transmitters is presented to provide information on seal effectiveness and preferred application. The effectiveness of various sealing methods is discussed, each ...

Energy Technology Data Exchange (ETDEWEB)

On February 1, 1984, the US Nuclear Regulatory Commission issued Generic Letter 84-04 on the subject of postulated pipe breaks in pressurized water reactor (PWR) primary coolant loops, opening the way for pipe-whip restraint exemptions. The letter substitutes the leak-before-break (LBB) criteria for the double-ended guillotine break regarding PWR primary reactor coolant system (RCS) piping and asymmetric blowdown loads. The LBB criterion refers to the fact that a piping flaw will leak before it breaks. The current requirement to provide pipe-whip restraints is applied within the plant to all high-energy piping with a potential for damaging structures, systems, and components essential to safe reactor shutdown. This includes primary RCS piping 30 in. and larger as well as smaller piping systems. A study was performed to evaluate the applicability of the LBB criteria proposed in NUREG-1061 to the latter set. The costs and benefits of this kind of ...

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In the nuclear power stations of GDR the rough radioactive waste includes borat-containing evaporator bottoms and spent ion exchanger resins. For its final disposal in deep geological formations (rock salt mines) this waste has to be solidified. The experience of one year lasting operation of a steam heated thin-film evaporator (heating surface 2 m"3) for evaporator bottoms to be solidified with a solid content of 200-250 g/l are reported on. In short time such amount of water is abstracted from the rough waste that due to the borate content a hot high-viscous product passes from evaporator to waste drum and there solidifies like glass to monolith. The product quality depends on the adjustment of the flow-equilibrium in the evaporator. Boric acid is used as matrix for the radioactive residues. The residual water content of the solidified waste product was about 15-20%, the volume reduction was ...

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... v. 1267(1) RADIATION CHEMISTRY, RADIOCHEMISTRY AND NUCLEAR

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Large scale experiments were performed to determine the effectiveness of thermal glow plug igniters to burn hydrogen in a condensing steam environment due to the presence of water sprays. The experiments were designed to determine if a detonation or accelerated flame could occur in a hydrogen-air-steam mixture which was initially nonflammable due to steam dilution but was rendered flammable by rapid steam condensation due to water sprays. Eleven Hydrogen Igniter Tests were conducted in the test vessel. The vessel was instrumented with pressure transducers, thermocouple rakes, gas grab sample bottles, hydrogen microsensors, and cameras. The vessel contained two prototypic engineered systems: (1) a deliberate hydrogen ignition system and (2) a water spray system. Experiments were conducted under conditions scaled to be nearly prototypic of ...

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A nonlinear mathematical model is presented of a WWER-440 nuclear power plant horizontal steam generator. On the proposed model is based a computer program for investigating transients in steam generators during loss of coolant accidents. Processes taking place at the primary side of the steam generator are described by a set of partial differential equations while those at the secondary side of the steam generator are described by plain differential equations with the variables being complex time functions. The model takes account of the coolant as both a single- and two-phase medium, of changes in the direction of the primary coolant flow and of changes in the direction of heat transfer. Heat transfer through the wall is based on a simple model of heat transfer through a thin-walled tube and includes a correction for the heat resistance of the wall. (author).

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For the conversion of thermal energy into electric energy in modern condenser power plants, according to the way of steam generation, two different types of power stations are built: power stations for fossile fuels and nuclear power stations. Also two classes of steam turbines were developed, corresponding to the two power station types, whose steam conditions, by experience and extensive calculations of economy, were determined so that a minimum of power generating cost will result. The two classes, the HMN and the SN series, are composed according to the modular system and designed in such a manner that with a small number of standard components, steam turbines for the power range between 100 and 2,500 MW can be built. (orig.).

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Steam generators for nuclear plants require access to the primary and secondary side for various inspection, cleaning and repair procedures and upon achieving access various inspection and cleaning tasks must be carried out. Steam generators currently at manufacture are therefore provided with; primary side access via the primary manways, steam drum access via the steam drum manway, U-bend region access via the secondary manway and drum-internals access passages, tubesheet secondary side access via tubesheet level inspection ports and tube support plate access via tube support level access ports. Many CANDU steam generators were built with only primary and drum manways and either no tubesheet ports or ports which were located ineffectively. The steam generators at Wolsong 1 were built with secondary side inspection ports that provided only ...

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BWRs are so designed that the exhaust steam from main steam safety relief valves is led to pressure suppression pools, and the steam is condensed in pool water, but at this time, dynamic load seems to arise in the pool water. In Tokai No. 2 Power Station, a Mark-2 containment vessel was adopted to improve the reliability as much as possible and to obtain the design with margin. In this report, the result of actual machine test in Tokai No. 2 Power Station and the method of reducing the load are described. When a relief valve works, the discharge of water in exhaust pipes into a suppression pool, the exhaust of air in exhaust pipes and repeated expansion and contraction of bubbles in pool water, and the exhaust of steam and condensation occur. As for the construction of the suppression pool in Tokai No. 2 Power Station, ...

International Nuclear Information System (INIS)

Alloy 690 and Alloy600 are used as a material for the steam generator tubing in the pressurized water reactor(PWR) of nuclear power plants due to its high corrosion resistance. Although those are a highly corrosion resistance material, their stress corrosion cracking(SCC) have been found on occasion, which are deeply related to a surface oxide film on a base material which have occurred on the primary side as well as the secondary side of a tubing. And The SCC is accelerated in the existing Pb which is the impurity of secondary steam generator components. The Oxide on a steel surface in an aqueous solution above 100 .deg. C is composed of a duplex film structure. The inner layer of the oxide is dense and less porous, which is formed by a growth of the oxide layer on the metal surface. The outer layer of the oxide is less adhesive, which is formed by a dissolution and precipitation mechanism. Growth ...

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The effect of heat transfer is described from heat exchange tubes of a horizontal steam generator on the distribution of primary water to the individual tubes of the tube bundle. It is shown that in a broad interval of mass flow rates and lengths of heat exchange tubes, the simplified method of calcualtion, i.e., calculation of the distribution of primary water into heat exchange tubes neqlecting the changes of physical properties of water along the heat exchange tubes, will yield sufficiently accurate results. (author).

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Initial major applications of neutron radiography (NR) to nuclear engineering were nondestructive inspections of nuclear fuel, control rods, reactor materials and some other components. Increase in the available neutron flux over 10{sup 8} n/cm{sup 2}s at the JRR-3M thermal neutron radiography facility (TNRF) in 1991 has expanded the application field to the dynamic but clear imaging of moving objects and fluid phenomena. The JRR-3M TNRF is facilitated with three major imaging systems, being characterized by spatial and/or temporal resolutions: 1. Static neutron radiography (SNR), 2. real-time neutron radiography (RNR) with an imaging rate of 30 frames/s and 3. High-frame-rate neutron radiography (HFRNR). SNR has been used for three-dimensional visualization of air-water two-phase flows in a simulated rod bundle. Three-dimensional computed tomography clearly illustrated average void fraction distributions around tie ...

International Nuclear Information System (INIS)

Vertical U-tube steam generators in Pressurized Water Reactors (PWRs) operating an All Volatile Treatment (AVT) secondary chemistry have experienced corrosion problems, particularly denting and sludges. The studies reported evaluate the feasibility of using a low-concentration (0.5 wt%) chemical cleaning process to remove corrosion product deposits from steam generator surfaces and magnetite from tube-to-support plate crevices of PWR steam generators. The process potentially may be applied at schedule intervals, such as during normal refueling outages, to maintain a steam generator in clean operating condition. This report describes the results of testing to evaluate the effectiveness of several chelant acids for dissolving steam generator sludges and crevice magnetite. Corrosion of carbon steel by the chelant acids and the effects of various inhibitors are ...

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As in the years before, the situation of steam boiler engineering was characterized by the following influencing parameters: Slow increase in electric power consumption; enhanced controversy over the CO{sub 2} emissions of fossil-fuel combustion; controversy over nuclear power; too much enthusiasm about the applications of additive and renewable energy. In all, six turbines with capacities over 100 MW were started but only one turbine of 180 MW was newly ordered. (orig./GL).

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It is desirable to avoid vibration problems by following appropriate guidelines and specifications at the design stage. Accordingly, design specifications were developed to prevent tube failures due to vibration in nuclear steam generators and liquid heat exchangers. These specifications are outlined in this report. (author). 14 refs., 2 figs.

International Nuclear Information System (INIS)

This paper documents some of the recent uses of robotics for inspection and maintenance activities in Ontario Hydro's nuclear power plants in areas other than fuel channels and steam generators. 7 figs.

International Nuclear Information System (INIS)

In nuclear power plant Paks, Hungary, alkaline oxidative (NaOH, KMnO_4, H_2O) and acidic reductive (citric- and oxalic acid, water) liquids are using for the decontamination of primary circuit equipment (main liquid circulating pumps, steam generators, pipelines etc). The above mentioned decontamination liquids are containing "1"1"0"mAg, "9"5Nb, "5"4Mn, "5"8 Co, "6"0Co, "5"1 Cr, "1"2"4 Sb radioisotopes, summarized radioactivity is between 10"3-8x10"4 kBq/dm"3 liquid. The decontamination liquid can be cleaned with reactive adsorption (active carbon) and ion-exchange process at elevated temperature (333-368 K) in multilayered columns. After purification the summarized radioactivity for "5"4Mn, "6"0Co, and "1"1"0"mAg are in the outlet liquid below 1 kBq/dm"3. Decontamination factor DF#approx =#10"3-10"4, volumetric reduction factor VRF#approx =#50-500.

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Bobbin-coil-type eddy current probes, which are conventionally used for nondestructive inspection of steam generator tubes in pressurized-water-type nuclear power plants, have poor detectability for circumferential flaws. Hence a new type of eddy current probe was proposed to detect effectively the magnetic flux component disturbed by a flaw and thus to eliminate the flaw direction dependency on the flaw detectability. In the course of development of the proposed method, structures of the probe were investigated based on the measurement of magnetic fields induced by exciting flat coils with several shapes. The new type of probe proposed here consists of differential pick-up coils detecting magnetic flux and exciting coils having a parallelogrammic shape, and its structure was fabricated experimentally in order to detect flaws independently of their directions. Nondestructive flaw detection tests was then conducted by using ...

International Nuclear Information System (INIS)

Some of the origins of corrosion encountered in the secondary side of pressurized water reactor steam generators are:-sludge accumulation (a mixture of metal oxides, primarily magnetite and copper) on tube sheet and attack of tube support plates by aggressive impurities leading to denting. Although Electricite de France has not suffered from these problems, it has developed a chemical cleaning process to dissolve corrosion products at both locations. (author).

International Nuclear Information System (INIS)

This thesis was carried out at Forsmark Nuclear Power Plant. The power plant in Forsmark consists of three boiling water reactors (BWR) which produce about 17% of Swedish electricity. In a BWR the nuclear reactions are used to boil water inside the reactor vessel. The water works both as a coolant and as a moderator and the resulting steam is used directly to run the turbines. A problem when running a BWR at low flow conditions is the density wave oscillations that might occur to the water flow inside the fuel assemblies. These oscillations arise due to the connection between power and flow rate in a heated channel with two-phase flow. In order to improve the stability performance of the channel an orifice plate is placed at the inlet of each fuel assembly. Today these orifice plates have sharp edges and a constant resistance coefficient. ...

Energy Technology Data Exchange (ETDEWEB)

The thesis gives an overview of the validation process for thermal-hydraulic system codes and it presents in more detail the assessment and validation of the French code CATHARE for VVER calculations. Three assessment cases are presented: loop seal clearing, core reflooding and flow in a horizontal steam generator. The experience gained during these assessment and validation calculations has been used to analyze the behavior of the horizontal steam generator and the natural circulation in the geometry of the Loviisa nuclear power plant. Large part of the work has been performed in cooperation with the CATHARE-team in Grenoble, France. (41 refs., 11 figs., 8 tabs.).

International Nuclear Information System (INIS)

A reasonable understanding of PWR steam generator corrosion mechanisms such as denting and wastage has been developed, and adequate chemistry control programs defined to obviate the magnitude and effects of these modes of attack. However, relatively unique corrosion attack modes have been encountered at several plants notwithstanding the presence of a reasonable to very good chemistry control program when considered in light of the Steam Generator Owners Group chemistry guidelines. The uniqueness of attack also suggests that parameters not routinely measured or monitored may be playing a significant role. In the authors opinions, the only reasonable method of routinely identifying corrosion accelerating species present in crevices, sludge piles, and deposits in PWR steam generators is by performing detailed chemical return studies during power transients, shutdowns, and long term layups. Although it would be preferable to ...

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The CPCU steam distribution network is supplemented by a return network for the condensation water. The data system installed in 1988 provides, for the real time, management of the function of the two networks and a reduction in production costs. For the steam, data required in the network, the boiler houses and from external sources are processed by local network of five microprocessors and permit: - with time delay: technical and economic production optimizing calculations, or forecasts, for the following day, of the total required output and the procedure necessary for supplying this at the lowest cost; - in real time: on the basis of the forecasts for the previous day, creating the production instructions for the boiler houses and the instructions for the network remote control elements; - in case of an unexpected occurrence: immediate creation of new operating forecasts for the boiler houses for the establishing ...

International Nuclear Information System (INIS)

A three-dimensional (3D) thermal-hydraulic model and a numerical procedure for the simulation and analysis of a steady-state, as well as transient operation of nuclear power plant components are presented. A two-fluid approach is applied to modelling of two-phase flow. Thermal-hydraulics of a horizontal steam generator in the WWER 1000 nuclear power plant has been simulated at the full load, steady-state operation. A comparison of the numerical results with data measured at the NPP Novovoronjezh shows good agreement. 3D numerical results can be used in plant design or retrofitting, in nuclear power plant operation and safety analysis and as improvement of existing one-dimensional thermal-hydraulics models of the horizontal steam generator which are assessed by system codes used for the nuclear power plant safety analyses. (author)

International Nuclear Information System (INIS)

There are a total of nine operational nuclear plants in Spain totalling 7.350 MWe. These units produced 54.265 x 106 KWh in 1990, 36% of the total generation in Spain. Seven of these plants are of the PWR type. The first plant in operation was Jose Cabrera (ZORITA) in 1968, one loop Westinghouse plant with a model 24 Steam Generator. Due to the design margin and careful operation of the Steam Generator of this plant its performance have been very good, with only 5% tubes plugged after 23 years of operation. This is one of the few units in the world that remains in phosphate chemistry. During the period 1981-1985 a total of four units, two in Almaraz and two in Asco entered in operation. These three loop s Westinghouse units use model D-3 preheater Steam Generators. The poor design and manufacture of the Steam Generators of these units have caused a large number of problems: ...

International Nuclear Information System (INIS)

An ion chromatograph has been modified to automatically sample ten liquid water streams from the secondary side of three pressurized water reactors, Calvert Cliffs, Unit One, Rancho Seco and McGuire, Unit 1. Sampling and measurement is semicontinuous with a cycle time of approximately five hours for 10 locations with sensitivities in the range of 0.1 to 0.5 ppb. The efficiency of the condensate polishing system and subsequent transport of sodium, chloride, and sulfate around the system can be readily followed. Sulfate has been shown to have unusual volatility into the steam phase from the steam generator as well as a tendency to pass through the condensate polisher.

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The research committee of the Atomic Energy Society of Japan on water chemistry standard aims at establishing the private standard of water chemistry of nuclear power plants. The committee gathers up 'BWR water chemistry management manual', 'PWR primary system water chemistry management manual' and 'PWR water chemical analysis standard method', and furthermore aims at the standardization of those in future. Looking back on the committee's activities for the past four years, latest results of research of water chemistry mainly contributing to safe and reliable nuclear power plants were described with the future perspective of water chemistry and a demanded break-through. (T.T.)

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... 2432 v. 43(1) SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS

International Nuclear Information System (INIS)

Steam generators (SGs) in PWRs and CANDUs are designed for at least a 30-year operating life. However, in the 15-25 years that SGs tubed with Alloy 600 have operated commercially, they have experienced reduced reliability, mainly due to SG tubing degradation. One of the degradation mechanisms that Alloy 600 SG tubing has suffered from is lead-induced stress corrosion cracking (PbSCC) in AVT and near-neutral SG environments. In contrast to Alloy 600 tubing, test data obtained in high temperature water indicate that Alloy 800 is resistant to cracking by lead and lead compounds. For Alloy 800 (cold worked, shot peened), the most aggressive environment is reported to be alkaline with minor concentration of PbO (80 ppm). Work by Max Helie et al. also concluded that Alloy 800 is not sensitive to lead assisted SCC for pH values close to neutrality, whereas it could be affected in high alkaline conditions. The work reported here investigated the ...

Energy Technology Data Exchange (ETDEWEB)

The authors report how the Midland cogeneration venture (MCV) is repowering the incomplete Midland nuclear plant to operate as a gas-fired combined cycle cogeneration facility. They discuss how their company is responsible for performing engineering, procurement, licensing, construction, start-up, training, and operational assistance for this facility. As shown, twelve gas turbine generator sets supply heat to twelve heat recovery steam generators (HRSGs), which are headered together on the steam side to provide energy to either of two existing steam turbine generators. This combination of equipment enables approximately 1,380 MWe of electrical generation capability, while supplying an average steam flow of 629,000 lbs/hr to an adjacent Dow Chemical Company pant. The MCV facility will also provide 60 MW of electric power to Dow. The authors report how the Midland ccogeneration ...

Energy Technology Data Exchange (ETDEWEB)

This report considers the effect of pipe-whip restraints and snubbers on the normal operation of piping systems in nuclear power plants. Also considered are the effect of these postulated event devices on reliability, economics, and the exposure of plant personnel to radiation. Field data were gathered from three nuclear power plants that had applied for Operating Licenses. Criteria, design philosophies, and data were obtained from the respective nuclear steam system suppliers, architects-engineers and utilities.

Energy Technology Data Exchange (ETDEWEB)

Discusses the resolution of the 2. Congress of the Polish Nuclear Society on the energy situation in Poland and recommendations for energy policy. Recommendations for use of nuclear power plants in Poland are made considering environmental pollution from coal combustion (air pollution by sulfur dioxide, nitrogen oxides and carbon dioxide as well as water pollution by salt from mine water discharged to rivers), development of the Polish economy, forecast increase in energy consumption and the role of nuclear energy in other European countries. Research on nuclear power plants, safety and environmental aspects as well as comparative efficiency of coal-fired power plants and nuclear power plants is evaluated.

International Nuclear Information System (INIS)

Project RPS116, ''Implementation of Boric Acid in the Field,'' was designed to demonstrate that a selected steam generator boric acid treatment is effective in arresting the progressing of denting in an operating steam generator. The PWR nuclear steam generators chosen for the testing were those at the Indian Point Unit 3 nuclear site. Hydrogen monitoring measurements and eddy current examinations had indicated that the Indian Point Unit 3, Series 44 steam generators had already reached an advanced stage of denting and tube support plate ligament discontinuities were observed. The objective of the boric acid treatment was to reduce the rate of denting by attempting to reproduce in the field the positive results achieved with boric acid in laboratory-simulated denting tests. Laboratory testing has indicated that implementation of a four day low power (25% power) ...

International Nuclear Information System (INIS)

The main reasons of tube failures in steam generators (SG) are considered. 1.Stress corrosion craining which has 28% of SG (most of them have stainless steel tubes). 2. Corrosion loss of metal, which accounts for 24% of tubes (phosphate corrosion due to addition of PO into water). 3.Denting-peripheral pressing of tubes in the openings of the foundation plates by the corrosive products, which are formed on internal surface of drillings in the foundation plates made of carbon steel. 4.Separation of a plating layer on tube panels. 5.Dratting-corrosion. 6.Metal fatigue. A series of experiments were conducted to study the influence of material selection on tube reliability (stainless steel 304, inconel-600, mone-400, incalloy-800). The problem of increase of SG elements reliability is a complex one and can be solved by direct selection of material, proper control of water-chemical conditions and other measures of corrosion ...

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The special geodetic problems encountered during the construction of the Paks nuclear power plants are treated. The main building with its hermetically connected components including the reactor, the steam generators, the circulation pumps etc. impose special requirements on the control net of datum points. The geodesy tasks solved during the construction of the main building are presented in details.

International Nuclear Information System (INIS)

Economic and safe operation of nuclear power plants requires reliable steam turbines with high efficiencies. The progress in flow mechanics achieved over the past few years has allowed the use of powerful methods of flow calculation in developments of new blading with greatly enhanced efficiencies. Thanks to the latest manufacturing techniques, the newly developed blading systems can be produced at low cost. Next to progress in flow mechanics, also the broadbased use of advanced finite-element calculations resulted in a more thorough grasp of the many problems associated with structural mechanics assessment of steam turbine components. The theoretical methods have been supplemented by comprehensive efforts in fracture mechanics and experimental materials studies, thus helping to create a reliable knowledge base which will help to avoid the dreaded stress corrosion cracking phenomenon, in components of ...

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Short communication.

Energy Technology Data Exchange (ETDEWEB)

Short communication

Energy Technology Data Exchange (ETDEWEB)

Many power plants use treated surface water as boiler feedwater and have difficulties in keeping within the limits for organic substances in feedwater (TOC < 200 ppb) or acid conductivity in steam condensate (< 0.2 {mu}S/cm). The characteristics of surface water and the role of organic substances in boiler feedwater are explained and discussed. [German] Viele Kraftwerke bereiten fuer die Erzeugung von Kesselspeisewasser Oberflaechenwasser auf und haben Probleme, die Richtlinien fuer organische Stoffe im Speisewasser (TOC < 200 ppb) oder Saeure-Leitfaehigkeit in Dampfkondensaten (< 0,2 {mu}S/cm) einzuhalten. Die besonderen Eigenschaften von Oberflaechenwaessern und die Bedeutung organischer Stoffe in Kesselspeisewasser werden erlaeutert und diskutiert. (orig.)

International Nuclear Information System (INIS)

A transient flow reduction burnout experiment was conducted with water in a uniformly heated, vertically oriented tube. Test pressures ranged from 0.5 to 3.9 MPa. An analytical method was developed to obtain transient burnout conditions at the exit. A simple correlation to predict the deviation of the transient burnout mass velocity at the tube exit from the steady state mass velocity obtained as a function of steam-water density ratio and flow reduction rate. The correlation was also compared with the other data. (author).

International Nuclear Information System (INIS)

This report focuses on a better understanding of the physical phenomenon related to the enhancement of boiling and non-boiling heat transfer by applying ultrasonic waves. Experimental results obtained both in a pool of water and in a vertically upward water flow proved clearly that macroscopic acoustic steam induced by ultrasonics is a major contribution to heat transfer augmentation. (author).

International Nuclear Information System (INIS)

Although there is service experience of Alloy 800 as tubing for superheaters in conventional and nuclear (HTR) power stations and in PWR heat exchangers, there is no corresponding service experience in sodium-cooled fast reactor steam generators. However, some limited experimental studies have been made of corrosion behaviour, and of possible structure modifications and effects on mechanical properties which occur during exposure of this material to a high temperature sodium environment, and these are summarised in the paper. It is concluded that further work needs to be done before Alloy 800 can be confidently endorsed for use as tubing in fast reactor steam generators. (author).

Energy Technology Data Exchange (ETDEWEB)

This project evaluated the feasibility of using a low-concentration (approx. 0.5 wt %) chemical cleaning process to remove corrosion product deposits from steam generator surfaces and magnetite from tube-to-support plate crevices of PWR steam generators. The primary objective was to develop a dilute process that could be safely applied at scheduled intervals, such as during normal refueling outages, to maintain a clean operating condition in the steam generator. The dilute chemical cleaning process developed in this project was demonstrated successfully on two model generators which were operated on faulted chemistry by DOE/CRC at Commonwealth's State Line Facility. Unit 5 was cleaned after 48 days of operation with 1% seawater fouling, and Unit 6 was cleaned after 112 days of operations with Lake Michigan water. This report describes work leading to the model generator cleaning demonstrations ...

International Nuclear Information System (INIS)

Corrosion of steam generator heat transfer tubes (SGHTT) is one of the important problems which affect safety operation of nuclear power plants (NPP), and the hazard of pitting corrosion of heat transfer tubes is the most serious. With an acoustic emission device, the signals during a corrosion test on SGHTT were collected and analyzed, and the corrosion points in the tubes were located accurately. The results show that pitting corrosion of heat transfer tubes has passed through three periods in its development: expansion phase, stationary phase and rapid developing phase. The corrosion damage of HTT can be found earlier with acoustic emission than any other non-destructive testing methods. Acoustic emission can be used for on-line and real-time monitoring of the safety and operation of the steam generator and has therefore a great significance. (orig.)

Energy Technology Data Exchange (ETDEWEB)

This report discusses the work done on EPRI program S149-1. In this program the feasibility of cleaning steam generators off line with organic chelants as a means of arresting denting corrosion was investigated. The rationale behind this program is to make use of those periods during which nuclear steam generators are in cold shutdown or wet layup to carry out a low-temperature soak with a combined chelant-inhibitor solution in order to dissolve some of the magnetite which has built up in crevices and to concomitantly remove entrained corrodents such as chloride ion. It was hoped that these soaks would be effective in reducing carbon steel support plate corrosion which produces tube denting.

International Nuclear Information System (INIS)

Corrosion is increasingly becoming an important factor reducing the reliability of many nuclear power plant components. The significance is evaluated of corrosion phenomena with respect to the reliability of primary circuit components of LWR's, viz., the reactor pressure vessel, primary piping, steam generator, and fuel elements. The mechanism of corrosion phenomena is explained and methods of minimizing their effects are presented. An analysis is made of the needs to solve the corrosion problems of nuclear power plants from the point of view of Czechoslovak producers and research and development activities. International cooperation is reviewed and main problems are formulated on which the solution of corrosion problems of structural materials used in WWER type nuclear power plants should be focussed. (author).

Energy Technology Data Exchange (ETDEWEB)

This paper presents the objective, scope, and status of the US Department of Energy's Materials and Structures Program to develop a data base on mechanical properties of structural materials for out-of-core structures and components for LMFBRs. Information on the development of a reference data base on materials for the reactor system, reactor enclosure system, primary heat transport system, intermediate heat transport system, and steam generator system is included. In addition, the development of the data and analyses to account for the effects of temperature and stress, as well as water/steam, sodium, and radiation environments, is described. Plans for the development of alternative materials for future out-of-core applications are presented.

International Nuclear Information System (INIS)

The importance that must be attached to the phenomenon of stress corrosion cracking of austenitic alloys is emphasized. The relation between chemical composition of various alloys and their sensitivity to cracking is shown with particular reference to the behaviour of Alloy 800. The different effects of alkaline anc chloride environments are discussed. Studies are reported of the general corrosion of Alloy 800 and other alloys in an environment representative of the primary coolant of PWR reactors; and of the behaviour of various alloys (including Alloy 800) in the conditions envisaged for their use for steam generators with superheat up to about 550 deg.C. (U.K.).

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Case studies are presented for treatment of refinery wastes in a pilot-scale thin-film evaporator, the removal of volatiles from industrial wastewater for two steam strippers, and the removal of semivolatiles from water by steam stripping followed by liquid-phase carbon adsorption. The report provides data on removal efficiency, air emissions, process residuals, treatment costs, and process limitations. Details on sampling and analytical procedures, quality assurance, and process data are contained in the Appendixes (Volume II).

Energy Technology Data Exchange (ETDEWEB)

The Superphenix 750-MW (thermal), once-throughtype steam generators are the result of over 10 years of intensive research and development. Detailed design and manufacture of the components lasted from 1977 to 1982. The main difficulties encountered during the construction of these prototype units concerned: demonstration of satisfactory high-temperature properties of Alloy 800 tubes; proven resistance of tube welds to thermal fatigue and of structures and shells to sodium/ water reaction effects; prevention of the tube bundle to flow-induced vibration; necessity of manual welding of shells to prevent hot cracking hazards related to the boron content of the steel; and welding of heat exchange tubes. None of these difficulties, however, invalidated initial major design choices, but provided a wealth of technical and technological experience and knowledge for the steam generators of the future. Plans for the future include ...

British Library Electronic Table of Contents (United Kingdom)

To analyze the effect of containment spray on gas mixing and depressurization, two experiments (ST3_1 and ST3_2) were performed with two interconnected vessels. These experiments were conducted in the frame of the OECD/SETH-2 project using the PANDA facility. The vessels were preconditioned such that a helium-rich layer is formed in the upper section of the first vessel, henceforth referred to as Vessel-1. In the case of the first experiment (ST3_1), the remaining volume of Vessel-1 and the entirety of the second vessel, Vessel-2, were filled with pure steam. For ST3_2, the second experiment presented here, pure steam was replaced with a steam-air mixture instead. Water was injected from the top of Vessel-1 with a spray nozzle projecting downwards. Transient behavior of system pressure, as...

International Nuclear Information System (INIS)

MOX fuel has been one of the mature nuclear fuels which can be used in light water nuclear power plant now. The development status in this domain in foreign countries, the major influence of MOX fuel on reactor performance and the countermeasures are introduced in this paper. The application of MOX fuel in China's PWR is discussed in the end. (authors)

Energy Technology Data Exchange (ETDEWEB)

In order to predict the effect of a fuel-coolant interaction after a hypothetical core-melt-down accident, a phenomenological model has been developed to describe the thermal and hydrodynamic behavior of a high-temperature molten jet when it interacts with saturated or subcooled water in a film boiling regime. The mechanisms of jet-material erosion were analyzed by Kelvin-Helmholtz instabilities on the coherent column and by boundary layer stripping on the leading edge. The heat transfer coefficient, vapor-film thickness, and net steam generation, all of which strongly affect the jet-breakup behavior, were solved analytically. It was found that the jet breakup (or erosion) depends strongly on the steam generation from the jet/water interaction. The jet-breakup length (i.e., penetration distance) was found to be sensitive to the initial jet temperature, water subcooling, and the ...

International Nuclear Information System (INIS)

In this paper, we calculated the radioactive concentration distribution of radioactive waste water, the temperature distribution of drained cooling water and the effect of implement from the Daya Bay Nuclear Power Plant on nearby waters range, discussed and analysed some problems of computational results and computation with Alternating Direction Implicit Method (ADI). The contents of the article included: the establishment of two-dimension tidal current equation, radioactive waste water pollutant dispersion equation and cooling water heat convection diffusion equation, the numerical difference calculation model of tidal current field, concentration field as well as temperature field, effect impingement with ADI method, numerical calculation results. The result of research showed that: when the Daya Bay Nuclear Power Plant is on normal ...

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The project is focused on the thermal hydraulic analyses of nuclear power plants. Specific areas of research have been the modelling of heat transfer in horizontal steam generator in presence of non-condensable gas, and the development of tools for multidimensional two-phase flow simulations. The effect of non-condensable gas on the heat transfer in the horizontal steam generator (SG) has been studied by calculating with APROS the PACTEL experiments NCG-1 (air injection) and NCG-3 (helium injection). The work done for the two-phase flow model development consists of two parts; improving the solution algorithm of porous media code PORFLO, and adding a homogeneous two-phase model to the commercial CFD code Fluent. (orig.)

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Measurements were completed on the solubility of cupric and cuprous oxides in liquid water and steam at controlled pH conditions from 25 to 400 C (77 to 752 F). The results of this study have been combined with those reported from this laboratory in two previous EPRI reports to provide a complete description of the solubility of these oxides and the speciation of copper dissolved in liquid water and steam as a function of oxidation state, temperature, pH, and in the case of steam, pressure. These constitute the first set of reliable data for cuprous oxide solubility over this range of conditions. For the more intensively studied CuO case, agreement was found between our results and those of previous studies of its solubility in steam, whereas only partial agreement was evident for its solubility in liquid water. For both oxides this ...

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Studies technological safety of installing a water spray pressure vessel between electrostatic dedusters and coal sludge ducts. These sprays are in use elsewhere for steam generator ash removal. Dust ignition and explosion tests were carried out to examine flame and pressure wave propagation through the vessel into ducts. Water jet diameter, amount of water sprayed and coal dust removed were varied. Pressure waves exceeded 250 Pa. Test results show the vessel to be suitable for installation in briquetting plants due to its flame and explosion barrier effect and extermination of smoldering dust fires. The only disadvantage of the vessel is seen as its water and electric power consumption; about 8/sup 3//h of water and 1.5 kW/h of power per vessel serving dedusters of a 2,200 m/sup 2/ rotary brown coal dryer.

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The present paper deals with the experimental determination of pressure drop across a four-cusped vertical channel. This geometry represents, ideally, the blockage condition in a typical pressurized water reactor with core degraded by accident. Experiments were performed for both single and two-phase flow. Water was utilized for the single-phase measurements whilst simultaneous flow of air and water simulated the steam-water flow. Observation of the prevailing two-phase flow regime was carried out, so that its mechanism could be fully understood. The averaged void fraction was also measured, by the gamma-ray attenuation technique. A wide range of water and air mass flow rates was covered, so that all flow conditions, possible to exist in a reactor with LOCA, could be investigated. New correlations for pressure drop are proposed. (Author).

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Real-time neutron radiography has been used to study the dynamic behavior of two-phase flow and measure vapor fractions in a steam-water duct at atmospheric pressure. This unique experimental technique offers one the opportunity to observe and record on videotape now Patterns and transient behavior of two-phase flow inside opaque containers without perturbing the environment. The neutron radiographic technique is non-intrusive and requires no special transparent window region. Data are recorded simultaneously over a large area of interest. Image processing of the video data can be employed to measure bubble velocities and time-averaged and Instantaneous vapor fractions.

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Hydrothermal systems typically consist of hot permeable rock which contains either liquid or liquid and saturated steam within the voids. These systems vent fluids at the surface through hot springs, fumaroles, mud pools, steaming ground and geysers. They are simultaneously recharged as meteoric water percolates through the surrounding rock or through the active injection of water at various geothermal reservoirs. In a number of geothermal reservoirs from which significant amounts of hot fluid have been extracted and passed through turbines, superheated regions of vapor have developed. As liquid migrates through a superheated region of a hydrothermal system, some of the liquid vaporizes at a migrating liquid-vapor interface. Using simple physical arguments, and analogue laboratory experiments we show that, under the influence of gravity, the liquid-vapor interface may become unstable and break up into ...

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A test program was conducted under contract to Sandia National Laboratories to investigate water/steam effects on elevated temperature low cycle fatigue and creep-fatigue of Alloy 800. This report presents interpretation and analysis of the test results. Tubular specimens with water sealed inside were cycled to failure under strain control. Tests were conducted to 616K (650/sup 0/F) and 922K (1200/sup 0/F); some at 922K included tensile or compressive hold periods to simulate creep-fatigue conditions. The tubular specimens showed significantly lower lives than solid bar specimens cycled at equivalent strain ranges. Rough internal surfaces contributed to early crack initiation with these specimens. Inclusion of hold periods caused further large reductions in cycles to failure.

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A new technique has been developed for the measurement of steam mass flowrate, water mass flowrate and total enthalpy of two-phase fluids produced from geothermal wells. The method involves precisely metered injection of liquid and vapor phase tracers into the two-phase production pipeline and concurrent sampling of each phase downstream of the injection point. Subsequent chemical analysis of the steam and water samples for tracer content enables the calculation of mass flowrate for each phase given the known mass injection rates of tracer. This technique has now been used extensively at the Coso geothermal project, owned and operated by California Energy Company. Initial validation of the method was performed at the Roosevelt Hot Springs geothermal project on wells producing to individual production separators equipped with orificeplate flowmeters for each phase.

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Titanium stabilized austenitic steel is sensitive to SCC in the secondary water under the horizontal steam generator operating conditions. SCC was observed under crevice conditions both at the primary collector flanges and the heat exchange tubes. In the crevice environment sulfates and chlorides as aggressive species and silicates and alumino-silicates as ''non-aggressive'' species are present in significant amounts. Local water chemistry parameters were evaluated using the MULTEQ Code. SCC experiments were carried out by rising displacement tests ar 275 deg C in an environment simulating the crevice conditions. Crack growth rate and K_I_S8C_C were determined for the environment where contents of some species were from 10"2 to 10"4 times higher than in blowdowns. (authors)

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Jun 1973. 2 p. United States USAEC Directorate of Regulatory Standards,

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This paper describes the modeling of horizontal steam generator with the TRACE code and calculation results of a loss-of-feedwater (LOF-10) experiment at the PACTEL facility. Parallel Channel Test Loop (PACTEL) is an integral test facility for a VVER-440 type nuclear reactor. The main objectives were to prepare a simulation model for its horizontal steam generator with the TRACE thermal hydraulic code and assess different modeling options of the code. PACTEL experiment LOF-10 was chosen for this assessment. The calculation results showed that TRACE is capable in simulating horizontal steam generator behavior both in steady state and during loss-of-feedwater transient. The phenomenon of heat transfer from primary to secondary side, steam superheating and flow reversal in the lowest heat exchange tubes were studied in detail. Different nodalization options were introduced. In the ...

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The thermal performance monitoring and optimization system TEMPO is developed at the OECD Halden Reactor Project. The system supports staff of nuclear power plants in identification and correction of problems, which cause small decreases in plant efficiency but which may lead to significant economical losses. The system-wide physical model consists of mathematical description of individual components, such as the reactor, the pumps, the heat exchangers, or the turbines, etc. TEMPO code has recently been extended with new steam generator (SG) models. The present paper summarizes the thermal-hydraulic modelling aspects of the vertical and the horizontal SG. The heat balance equations and their solution are shown with the appropriate initial and boundary conditions. The method of the calculation of the pressure losses are also introduced. The vertical SG model is based on a U-tube structure and treated as a 1D flow channel. The horizontal approach ...

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Since 1984 the thermal-hydraulic code ATHLET has been also applied for the analyses of LOCA and transients in VVER plants. The specific design of these plants especially of the steam generator design requires a specific modelling of the phenomena which may occur under LOCA and transient conditions in these plants. Differences in design compared to the design of western reactors have been briefly listed. Specific phenomena occurring under small leak accidents are shortly described. The consideration of the simulation of the boiler-condenser mode illustrates the modelling requirements for a code which may be applied to the prediction of such a thermal-hydraulic behaviour. Facing the lack of experimental data, the reliability of the simulation has been discussed by means of plausibility studies based on the momentum balance for steam and water. In summary: The VVER reactors differ in design compared to reactors of western ...

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The solar central receiver at the Barstow Pilot Plant is a once-through steam boiler consisting of vertical arrays of Alloy 800 tubes. Water/steam leaks associated with tube bends near the receiver outlet were observed after 16 service months. The leaks resulted from through-wall cracks localized in the crown of tube bends operating in the temperature range from 550 to 650/sup 0/C. Initiation occurred on the ID (steam side) of the tube and propagated transgranular through the tube wall. Cracking was axial and circumferential; in general, the circumferential cracks were more severe than the axial cracks. Thick oxide layers showed on the ID of the receiver tubes; a 25-..mu..m thick oxide layer had formed on tubing which operated at 650/sup 0/C. In addition, an enhanced oxidation layer was observed along a narrow band in the crown of the tube. This band was up to five times thicker than the oxide elsewhere in the tube. All ...

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A software called KEPCO RCM workstation (KRCM) has been developed to optimize the maintenance strategies of Korean nuclear power plants. The program modules of the KRCM were designed in a manner that combines EPRI methodologies and KEPRI analysis technique. The KRCM is being applied to the three pilot system, chemical and volume control system, main steam system, and compressed air system of Yonggwang Units 1 and 2. In addition, the KRCM can be utilized as a tool to meet a part of the requirements of maintenance rule (MR) imposed by U.S. NRC. 3 refs., 4 figs. (Author)

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Alloy 600 is used as a material for steam generator tubing in pressurized water reactors(PWR) due to its high corrosion resistance under PWR environment. In spite of its corrosion resistance, stress corrosion cracking(SCC) has occurred on the primary side as well as the secondary side of the tubing. Oxide on steel surfaces in aqueous solution above 100 .deg. C is composed of duplex film structure. Inner layer of the oxide is dense and less porous, which is formed by growth of oxide layer on metal surface. Outer layer of the oxide is loose adhesive, which is formed by dissolution precipitation mechanism. Growth processes occur at the metal/oxide and oxide/electrolyte interfaces and are controlled by transport of the layer forming species through the layer, i.e. by the inward diffusion of oxygen including electrolyte species and the outward diffusion of metal cations. Understanding of basic electrochemical behaviors about anodic dissolution and ...

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As part of a coordinated program, AECL is developing a set of tools to aid with the prediction and management of steam generator performance. Although stress corrosion cracking (of Alloy 800) has not been detected in any operating steam generator, for life management it is necessary to develop mechanistic models to predict the conditions under which stress corrosion cracking is plausible. Therefore, constant extension rate tests were carried out for Alloy 800 under various steam generator crevice chemistry conditions at applied potentials. These tests were designed to evaluate the stress corrosion cracking susceptibility of Alloy 800 under CANDU( steam generator operating conditions. Based on the experimental results, the recommended electrochemical corrosion potential/pH zone for Alloy 800 determined by electrochemical polarization measurements was verified with the respect of stress corrosion cracking ...

International Nuclear Information System (INIS)

A detailed outline of the application of the MELCOR and RELAP5/MOD3.1 codes to the analysis of the thermohydraulic response and determination of other parameters of the medium on the floor is given for several classes of secondary coolant circuit accidents along with the description of the related facilities. An overview is presented of the maximum values and time behavior of the thermohydraulic parameters, pressure, temperature, relative humidity, and water level on the floor. Transverse rupture of the steam generator, main steam header, or main feedwater header piping during normal operation is considered as the initiating event. Pressure is only 10% higher than the atmospheric pressure. Air temperature attains a value as high as 100 degC. Relative humidity is 100%, persisting as long as the steam source is available. The water level is typically about 8 cm and never exceeds 15 ...

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This document gathers a series of very short articles concerning nuclear industry around the world. Areva company is investing 30 million euros in its Chalon-Saint-Marcel plant, it is the consequence of the extension of service life of nuclear power plants in the Usa. Areva holds 40% of the American market concerning the replacement of steam generators and 50% of that concerning the replacement of closure heads. The Obrigheim nuclear power plant was definitely closed down on may 2004, this decommissioning is a step forward in the German policy of progressively stepping out of nuclear energy. Chinese authorities are willing to construct 40 nuclear reactors in 15 years, despite that, the contribution of nuclear energy to the generation of electricity will reach only 4% in 2020. In 2007 Cea will begin the construction works of a new research ...

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Photodestruction has received much attention in recent years. In particular, titanium dioxide (TiO{sub 2}) and ozonolysis have attained a position of prominence. These technologies hold promise for the destruction of trace amounts of aqueous high explosives that are generated by load and pack operations, as well as demilitarization activities. Currently this water is treated by passing through a bed of activated carbon. The carbon is then steam regenerated and reused, thus creating a second waste stream which must be disposed of, or the carbon is burned directly. Recent trends in environmental regulation have shown that this may not be a viable option for process water remediation in the future. This talk will discuss efforts to employ alternate aqueous treatment techniques that not only remove the explosives compounds but are able to transform the parent compound into carbon dioxide and water. Titanium ...

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A guidance document on the instrumentation for monitoring and control of cycle chemistry for the steam-water circuits of fossil-fired and combined-cycle power plants was developed within the IAPWS Power Cycle Chemistry Working Group. This technical guidance document has been authorized by the International Association for the Properties of Water and Steam (IAPWS) at its meeting in Doorwerth, The Netherlands, 6-11 September, 2009, for issue by its Secretariat. The members of the IAPWS are: Britain and Ireland, Canada, the Czech Republic, Denmark, France, Germany, Greece, Japan, Russia, and the United States of America, and the associate members Argentina and Brazil, Italy, and Switzerland. In order to achieve suitable chemical conditions in steam-water circuits it is essential to establish reliable monitoring of key parameters on every plant. This enables the demonstration of operation within cycle ...

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This article describes a new natural-gas-fired combined cycle cogeneration plant in Bellingham, Washington. The topics of the article include community impact, siting constraints, natural gas fuel, the flexibility provided by the steam turbine, the cooling tower and pumps, air-quality, noise, and cooling water system constraints, and community relations program.

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Full text: During JET operations, tritium contaminated waste is generated principally but not exclusively from 'intervention' work and from removing or replacing redundant items. It is essential for JET and for any future fusion plant to have available a route for managing each waste stream however large or small, both during operation and decommissioning of the plant. The long term outcome is to have for each tritiated waste stream from JET a route for its management leading to its eventual disposal or recycling (and thus to be available for similar waste streams which will be produced by ITER operations). Since several years SCK#centre dot#CEN has been developing techniques for the treatment of tritiated waste. Amongst them, technologies for water detritiation, for the treatment of tritiated organic liquids and for the decontamination of several types of solid tritiated waste. Our R and D focuses on the development of a system to decontaminate surface ...

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A new three-dimensional (3D) acoustic modelling method was developed using a first-order hyperbolic wave system which was solved with explicit finite dfferences. The numerical solution of the 3D wave system provides a useful method for simulating evolution of a pressure field corresponding to compressional type waves. Existing two-dimensional (2D) elastic modelling algorithms were modified and fine-tuned for computationally efficient and realistic wave propagation simulations in complex structures. An original formulation of the 3D reverse time migration method was developed which is very accurate, does not suffer from unwanted evenescent energy, can image dips beyond 90{degree}, and does not generate multiple energy. Two case studies were performed that involved steam stimulation projects in the Cold Lake deposit. Simulations were performed during different phases of the steam stimulation process to examine the relation between reservoir ...

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In view of the widespread cracks discovered in the reinforced-concrete circulating water culvert of Daya Bay Nuclear Power Station, cause analysis for the cracks is made in terms of construction and design. It is concluded that the cracks mainly resulted from shrinkage and temperature. Corresponding countermeasures are put forward thereafter for reference of similar projects

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This patent describes a nuclear system comprising a containment airspace in which a nuclear reactor pressure vessel is disposed there being a reactor core within the pressure vessel. It comprises a heat exchanger elevated a distance above the pressure vessel; a pool of water surrounding the heat exchanger; means for venting the pool of water to an environment outside the containment; a heat exchanger entry conduit within the containment, the entry conduit having an open lower end communicating with the containment space, and an upper end connected to the heat exchanger, water-containing heated fluid present in the containment airspace incident a pressure vessel loss of coolant event entering and flowing through the entry conduit into the heat exchanger for cooling the fluid to convert water vapor therein to a condensate and separate non-condensable gasses ...

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Published in summary form only.

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The SGTR project has been carried out within the fifth EURATOM Framework Programme (Contract No FIKS-CT-1999-0007). Its main objective was to provide an experimental database and to develop and/or verify models to support definition of accident management measures in the hypothetical case of a Steam Generator tube Rupture (SGTR) sequence. The project addressed both vertical and horizontal steam generator designs. This report summarises the main results obtained in the intermediate scale experimentation that addressed Western type steam generators. The specific goal of this test programme was to investigate aerosol retention in the break stage of the secondary side of a water-empty steam generator. The test matrix consisted of 12 tests that explored the influence of variables such as break type and orientation and inlet gas flow rate. This work was performed in the PECA facility of ...

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Alloy 800 has been used for steam generator (SG) tubing for more than 30 years, primarily in CANDU reactors worldwide and in reactors in Germany. Extensive laboratory testing and in-service experience suggest that the Alloy 800 tubing has excellent resistance to corrosion-related degradation under appropriate operating conditions. In planning refurbishment of nuclear plants stations, a key concern is the longevity of existing SGs up to the 60-year lifetime of the refurbished plant. The paper reviews an existing methodology based on the concept of the improvement factor, and refines its estimation based on data specific to CANDU operating conditions. The paper presents a more advanced Bayesian probabilistic approach to estimate the degradation free lifetime distribution of Alloy 800 tubing, which is used to quantify the probability of degradation during the service life and to evaluate the impact of potential occurrences of degradation on ...

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It is hypothesized that the good performance of Alloy 800 in steam generator service is due to its relative immunity to two distinct mechanisms of stress corrosion cracking; the argument also applies to intergranular corrosion. One mechanism operates in the high-nickel region (Alloy 600 and nearby model alloys) and is due to internal intergranular oxidation. The other operates in the low-nickel (stainless steel) region and is due to de-alloying of Fe and/or Cr. This latter mechanism may, under special conditions, operate in high-Ni, high-Cr alloys such as 690. Some essential features of the de-alloying mechanism are demonstrated using strong caustic solutions, and the prospect of extending this approach to dilute high-temperature environments is discussed. (author)

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Actinides other than the main uranium or plutonium isotopes take a growing part in the different stages of the nuclear cycle. For the French nuclear power program based on the development of light water reactors and fast breeders, many evaluations of the secondary actinides build up are made for the both reactor types using mainly the existing reactor codes. The comparison of these foreseen compositions with experimental results allows to perform some adjustments of the neutronic data. The secondary actinide compositions are given for some typical fuels and their consequences on the nuclear cycle are discussed. An hypothetical burning of these wastes in fast reactors has been studied and the main conclusions are reported.

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In connection with the Public Inquiry into the CEGB proposal to build a pressurised water reactor at Sizewell in Suffolk, the case for nuclear power is examined under the headings: the economics of nuclear power - how they would like them to be; systems analysis - net effective cost; CEGB prejudices the results (comparison with coal-fired plants; forecasting on various assumptions); discounting future costs; back-end costs soar (reprocessing); real reprocessing costs; AGR costs balloon.

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As part of a coordinated program, AECL is developing a set of tools to aid with the prediction and management of steam generator performance. Although stress corrosion cracking (of Alloy 800) has not been detected in any operating steam generator, for life management it is necessary to develop mechanistic models to predict the conditions under which stress corrosion cracking is plausible. Therefore, constant extension rate tests were carried out for Alloy 800 under various steam generator crevice chemistry conditions at applied potentials. These tests were designed to evaluate the stress corrosion cracking susceptibility of Alloy 800 under CANDU( steam generator operating conditions. Based on the experimental results, the recommended electrochemical corrosion potential/pH zone for Alloy 800 determined by electrochemical polarization measurements was verified with the respect of stress corrosion cracking ...

International Nuclear Information System (INIS)

Experiments have been performed with the PMK integral-type facility, a model of WWER-440 type PWRs, to investigate two-phase natural circulation behaviour. The phenomena to be expected in this reactor type are different from those in PWRs with vertical steam generators mainly due to the loop seal in the hot leg and the horizontal layout of the steam generator heat transfer tubes. The experiments showed that the system is repressurized when the water level drops to the hot leg elevation due to the effect of the loop seal. Opening of the loop seal can be smooth, but may lead to oscillations depending on the power and the mass inventory. Natural circulation recovers after the hot leg loop seal is opened, but then decreases with further mass inventory decrease. (orig.).

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Briefly described is knowledge of crevice corrosion, corrosion cracking and denting. In evaluating the corrosion resistance of steam generator materials it is necessary to distinguish corrosion problems caused by the primary coolant side and by the secondary circuit side. At present tubes are manufactured of 7 austenitic alloys of a different chemical composition, and available information shows that views on their corrosion resistance differ. Greatest attention has been devoted to corrosion cracking in the presence of NaOH. Findings related to I-600, I-800, I-690 and AISI 316 are given. Corrodibility by sulfur-containing products is now being studied, namely the intercrystalline corrosion cracking caused by the presence of H/sub 2/S/sub 4/O/sub 6/. Knowledge gained in this respect is summed up.

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A thermal analysis of the Omnium-G receiver is presented and the technique is shown to be generally applicable to solar thermal receivers utilizing a directly heated thermal mass. The thermal loss coefficient, including reradiation losses, is calculated and shown to agree quite well with the experimentally measured thermal loss coefficient. The rate of heat transfer to the working fluid is also analyzed and the analysis is used to show that the Omnium-G receiver is well matched to the water/steam working fluid because the steam outlet temperature is almost the same as the receiver temperature. A general procedure for calculating receiver performance is presented. With this procedure, the energy delivery to any working fluid, the delivered temperature of the working fluid, and the pressure drop through the receiver can be determined. An example of the calculation is also presented.

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As a three-year joint university-industry effort, development of a generalized boiling transition analysis method has been started in 2002 aiming at enhanced capabilities of subchannel analysis for a wide variety of BWR-type fuel bundle geometry from ordinary BWR to tight lattice fuel bundles. For this purpose, five dominant factors affecting boiling transition phenomena have been identified on which our efforts of experimentation and numerical analyses are focused. In this report, as the first-year achievement, we will describe a master plan of the development and contents for experimental approaches to construct thermal-hydraulic databases. The databases will be utilized for the developments of constitutive equations to describe the basic characteristics of the elementary processes. The planned experiments are divided into two groups. One is air-water experiments at atmospheric pressure, and the other is steam-water experiments up to 1 MPa. ...

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During 17 years since the start of operation of the first commercial LWR in Japan, many LWRs have experienced various corrosion damages, but the causes of them were clarified, and the counter-measures were executed effectively in actual plants, as the results, the cause of corrosion damage decreased remarkably, and now, the high rate of operation has become to be maintained. In this paper, the major cases of corrosion damage experienced in LWRs in Japan and foreign countries, the causes of them and the countermeasures, the problems of hereafter and so on are described. The corrosion damage of metallic materials in the environment of LWRs occurs in the parts in contact with high temperature, high pressure water and steam, such as stainless steel piping in the primary cooling system of BWRs, and nickel alloy heating tubes of steam generators, carbon steel feed water piping and zirconium alloy fuel ...

International Nuclear Information System (INIS)

Purpose: To stably control the reactor water level so as not to cause excess water feeding in a BWR type reactor. Constitution: A flow control valve is disposed to the exit of a feedwater pump for a nuclear reactor and the valve is controlled by a flow regulator to maintain the water level constant in the reactor. A signal from a water level controller is inputted to the flow regulator to thereby control the flow rate control valve. In this case, the flow regulator remains in a saturated state just after the starting of the feedwater pump, in which the pump flowrate is at 100% to result in an excess water feeding condition. In view of the above, a feedback circuit is provided to the flow regulator so that the saturated state is eliminated and the water feeding can be controlled directly from the water level controller. ...

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In 1972 a consortium consisting of Gebruder Sulzer A.G., Winterthur and EVT, Energie- und Verfahrenstechnik G.m.b.H., Stuttgart, was awarded the contract by HBR, Hochtemperatur Reaktorbau G.m.b.H., Koln/Mannheim, for the design, manufacture and installation of the 6 steam generator units for the prototype 300 MWe Thorium High Temperature Reactor (THTR) at Schmehausen, Germany. The design and manufacture of the 6 units is done by Sulzer; installation of the units as well as the manufacture of the external pipework and headers is done by EVT. The units are built under semiclean conditions in a special workshop at CCM-Sulzer in mantes, France. This workshop had been used previously for the manufacture of the steam generators for the nuclear power plants St. Laurent I, II, Vandellos (500 Mwe each) and EL-4 (70 Mwe). The present paper defines the operating conditions and the chemical composition of the materials used, including ...

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Inherent boron dilution can occur in case of a Small Break LOCA when low borated water is mainly accumulated in the U-legs due to reflux boiling in the Steam Generator tubes after the loss of natural circulation. The restart of the natural circulation may lead to criticality because of the injection of these low borated slugs towards the core. To evaluate this potential risk, the boron concentration at the core inlet has to be known which makes necessary to estimate the mixing phenomena in the cold leg, in the downcomer and in the lower plenum: CFD calculations are required. First of all the validation of CFX5 CFD code on the relevant phenomena of inherent boron dilution has been established (UPTF TRAM C3 test). Then, an application to the 900 MW French Pressurized Water Reactor series has been performed. (authors)

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SMART (System integrated Modular Advanced ReacTor) , is under development at the Korea Atomic Energy Research Institute (KAERI). SMART is an integral type pressurized water reactor which contains a pressurizer, 4 reactor coolant pumps (RCPs), and 8 steam generator cassettes(S/Gs) in a single reactor vessel. This reactor has substantially enhanced its safety with an integral layout of its major components, 4 trains of safety injection system (SIS), and an adoption of 4 trains of passive residual heat removal system (PRHRS) instead of an active auxiliary feedwater system . The thermal power is 330 MWth. During the conceptual design stage, a preliminary PSA was performed. PSA results identified that a steam generator tube rupture (SGTR) is one of the most important initiating events which results in a high core damage frequency. Clear understanding of accident progression with various combinations of the safety systems helps ...

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The presentation summarises the highlights of experimental results obtained for VVER type horizontal steam generator heat transfer, primary side flow pattern, and mixing in the hot collector during secondary side boil-off with primary at single-phase natural circulation. The experiments were performed using the PACTEL facility with Large Diameter (LD) steam generator models, with collector instrumentation designed specifically for these tests. The key findings are as follows: (1) the primary to secondary heat transfer degrades as the secondary water inventory is depleted, following closely the wetted tube area; (2) a circulatory flow pattern exists in the tube bundle, resulting in reversed flow (from cold to the hot collector) in the lower part of the tube bundle, and continuous flow through the upper part, including the tubes that have already dried out; and (3) mixing of the hot leg flow entering the hot collector and ...

International Nuclear Information System (INIS)

The application of chemical cleaning for dissolving and removing scale and sludge is being planned in the Japanese pressurized water reactor (PWR) plant in order to maintain high heat transfer performance and to prevent steam generator (SG) tube degradation. In this paper, the effectiveness of the Electric Power Research Institute (EPRI) and German Kraftwerk Union (KWU) processes on the integrity of structural materials other than SG tubes and the comprehensive applicability of chemical cleaning are discussed. The integrity of structural materials such as carbon steel, low-alloy steel and stainless steel was maintained after the EPRI and KWU processes. KWU chemical cleaning tailored for crevice cleaning has been studied to improve its cleaning effectiveness in crevices and to control the corrosion depth of structural materials less than the criterion for corrosion depth. (author)

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The Phebus FP project in an international reactor safety project. Its main objective is to study the release, transport and retention of fission products in a severe accident of a Light Water Reactor (LWR). The FPT4 test was performed with a fuel debris bed geometry, to look at late phase core degradation and the releases of low volatile fission products and actinides. Post Test Analyses results indicate that releases of noble gases (Xe, Kr) and high-volatile fission products (Cs, I) were nearly complete and comparable to those obtained during Phebus tests performed with a fuel bundle geometry (FPT1, FPT2). Volatile fission products such as Mo, Te, Rb, Sb were released significantly as in previous tests. Ba integral release was greater than that observed during FPT1. Release of Ru was comparable to that observed during FPT1 and FPT2. As in other Phebus tests, the Ru distribution suggests Ru volatilization followed by fast redeposition in the fuelled section. The ...

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The Idaho Nuclear Technology and Engineering Center (INTEC) was home to nuclear fuel reprocessing activities for decades at the Idaho National Engineering and Environmental Laboratory. As a result of the reprocessing activities, INTEC has accumulated approximately one million gallons of acidic, radioactive, sodium-bearing waste (SBW). The purpose of this demonstration was to investigate a reforming technology, offered by ThermoChem Waste Remediation, LLC, (TWR) for treatment of SBW into a ''road ready'' waste form that would meet the waste acceptance criteria for the Waste Isolation Pilot Plant (WIPP). TWR is the licensee of Manufacturing Technology Conservation International (MTCI) steam-reforming technology in the field of radioactive waste treatment. A non-radioactive simulated SBW was used based on the known composition of waste tank WM-180 at INTEC. Rhenium was included as a ...

Energy Technology Data Exchange (ETDEWEB)

Steam generator tubes in nuclear power plants are periodically checked by means of eddy current probes. The output of a probe is composed of three types of signals: known events (rolling zone, support plates, U-bend part), noise (mainly metallurgical noise) and possible flaws. The latter are random transients, both in arrival time and in shape: they have to be detected and then estimated, before to be fed to the high level stages of a diagnostic system. The objective of the study presented is to develop a semi-automatic system, which could manage and process more than 1 M-bytes of data per tube and provide an operator with reliable diagnostics proposals within a few minutes. This can be achieved only by cooperation of several digital signal processing techniques: detection, segmentation, estimation, noise subtraction, adaptive filtering, modelization, pattern recognition. The paper describes some of these items.

Energy Technology Data Exchange (ETDEWEB)

The Alloy 800 tubes used in CANDU 6 steam generators have not experienced significant corrosion damage to date, which may be attributed to successful water chemistry control strategies. However, it is known that Alloy 800, like other steam generator (SG) tubing materials, is not immune to corrosion, especially pitting, under some plausible but off-specification operating scenarios. Electrochemical measurements provide information on corrosion susceptibility and rate, which are known to be a function of water chemistry. Using laboratory data in combination with chemistry monitoring and diagnostic software it is possible to assess the impact of plant operating conditions on SG tube corrosion for plant life management (PLIM). In this context, this paper discusses the results of electrochemical measurements made to elucidate the corrosion behaviour of Alloy 800 SG tubes under conditions simulating those ...

International Nuclear Information System (INIS)

The symposium covers papers under different sections namely, (i) Core physics and Fuel management, (ii) Commissioning of facilities and systems, (iii) Operational experience and Human resource development, (iv) Fuel handling, Maintenance management and Surveillance, (v) Instrumentation and Control and Power supply systems, (vi) Analysis, modifications and developments for enhancing operational safety, (vii) Chemistry control and Effluent management, (viii) Radiation and industrial safety and (ix) Steam generators, Turbo-generators and other auxiliaries. Papers relevant to INIS are indexed separately. (author)

International Nuclear Information System (INIS)

During a hypothetical severe incident in a nuclear power plant with core meltdown a large part of radioactive material is present as aerosol particles in the reactor containment. In current severe accident containment codes the potential influences of hydrogen combustions on the behaviour of aerosols are not considered. Among other effects dry resuspension can increase the aerosol concentration in the atmosphere. Already deposited aerosol material can be re-released into the containment atmosphere by atmospheric currents induced by hydrogen deflagrations or by other phenomena like steam explosions. The objective is to assess the possible influence of this dry resuspension effect on the radioactive source term. (author)

Energy Technology Data Exchange (ETDEWEB)

This study deals with the application of wavelet analysis to detection and characterization of defects from eddy-current and ultrasonic testing signals of a low signal-to-noise ratio. Presented in this paper are the methods for processing eddy-current testing signals of heat exchanger tubes of a steam generator in a nuclear power plant. The results of processing eddy-current testing signals of tube test pieces with artificial flaws show that the flaw signals corrupted by noise and/or non-defect signals can be effectively detected and characterized by using the wavelet methods. (author)

International Nuclear Information System (INIS)

Water chemistry of nuclear power plants has played an important role in reduction of personnel doses, structural materials and fuel integrity assurance, and reduction of radioactive wastes production. Further contributions are requested for advanced utilization of the LWR, advanced fuels and aging management of plants. Since water chemistry has an effect on all structure and materials immersed and at the same time affected by them, the optimum control not sticking to specific issues and covering the whole plant is required for these requests. Taking account of roles and activities of the industry, governmental institutes and academia, road maps on research and development plans for water chemistry were compiled into identified eleven items with targets and counter measures taken, such as common basic technologies, dose reduction, SCC mitigation, fuel cans corrosion/hydrogen absorption mitigation, ...

International Nuclear Information System (INIS)

On 24th of July 2007 Westinghouse Electric Co. signed landmark contracts with China's State Nuclear Power Technology Corporation (SNPTC), to provide four AP1000 nuclear power plants in China. The AP1000 is a two-loop 1117 MWe Pressurized Water Reactor (PWR). It is based on proven technology, but with an emphasis on safety features that rely on natural driving forces, such as pressurized gas, gravity flow, natural circulation flow and convection. Ansaldo Nucleare has provided a significant support to the passive plant technology development and, starting from 2000, is cooperating with Westinghouse to development of the AP1000 Plant. In the frame of the AP1000 Chinese agreement, Ansaldo Nucleare, in Joint Venture with Mangiarotti Nuclear, has signed a contract with Westinghouse for the design and the supply of innovative components to be installed in the first ...

Energy Technology Data Exchange (ETDEWEB)

Water injection into a vapor-dominated geothermal reservoir is an effective method of sustaining steam production from the field. Injection puts additional water to the reservoir and raises the prevailing reservoir pressure. This process improves the field`s productivity. However, the increased pressure also increases the water retention capacity of the reservoir rocks through the effects of adsorption and capillary condensation. Due to the significant costs associated with water injection programs, optimizing injection not only involves maximizing the energy yield from the resource but also the present worth of the project. Two crucial parameters that need to be established are: (1) how much to inject; and, (2) when to inject it. This study investigated the optimal design of these parameters. It was found that comparable energy yield can be attained for injection programs that are ...

International Nuclear Information System (INIS)

In General Electric's design of the Simplified Boiling Water Reactor (SBWR), The depressurization valves (DPV) are installed in the reactor pressure boundary: four are connected to the reactor vessel by means of nozzles, and two more are located on the main steam pipes (one DPV for each line), which act during particular transients and/or loss of coolant accidents (LOCA), consequently providing the reactor vessel with a safe quick depressurization system. Once the vessel is de pressurised, the passive gravity-driven cooling system (GDCS) starts to operate, permitting the injection of water required for continuous core cooling. DPVs are leak tight, with welded flaps, actuated by a [striker[hammer***] which is activated by an explosive mixture. The dynamic loads that open these valves include, in addition to those produced by steam (typical in any thermodynamic transient with open/close valves), other ...

International Nuclear Information System (INIS)

Purpose: To effectively cool the reactor core in a steam atmosphere by upwardly directing several of spray nozzles attached to a ring header thereby increasing the flying distance of the spray. Constitution: Ring headers in two upper and lower stages are disposed above the outer circumference of a reactor core and each of the ring headers is mounted with spray nozzles. Among the spray nozzles, at least several nozzles mounted to the ring header at the lower stage are directed such that the center axis for each of the nozzle is raised above the horizontal axis and other several nozzles are mounted with the nozzle center axis directed downwardly from the horizontal axis. Accordingly, even if collapsing phenomenon occurs in the jetting stream due to the condensation in the steams that forms the operation atmosphere of the reactor core spray cooling device, a sufficient amount of emergency cooling water can be distributed over ...

Energy Technology Data Exchange (ETDEWEB)

The gross calorific value (H{sub o}) is the amount of heat which is generated by total combustion of a type of fuel. It also includes that part of heat which is generated by steam from evaporated water contained in heating gas. In conventional heating boilers, this heat portion is not being utilized at all. To utilize the gross calorific value, it is necessary to extract from the exhaust gas the evaporation heat bound in steam and to return this to the heating system. This means that the exhaust gas has to be chilled by the return water of the heating system and condensed in suitable heat exchangers to well below the dew point. (orig.) [Deutsch] Waermelieferanten im Bereich der Nahwaermeversorgung koennen durch die Anwendung verfuegbarer und bewaehrter Brennwerttechnik ihre Kosten senken und einen zusaetzlichen Beitrag zum Umweltschutz leisten. Mit dem richtigen Brennwertkessel amortisieren sich ...

International Nuclear Information System (INIS)

SMART (System-Integrated Modular Advanced Reactor) is a first-of-the-kind integral reactor with 330 MW thermal power under active development by Korea Atomic Energy Research Institute (KAERI) for power generation and seawater desalination. SMART employs various design features that are not typically found in other nuclear power plants. Examples include a unique passive residual heat removal system (PRHRS), and enclosure of a pressurizer, eight helical steam generators, and eight canned reactor coolant pumps inside the reactor pressure vessel. This paper presents risk insights on the SMART reactor gained during the development of a regulatory PSA model by Korea Institute of Nuclear Safety (KINS)