Device for starting a steam generator by heating sodium in a reactor

International Nuclear Information System (INIS)

Nakano, Hisao.

1975-01-01

Object: To enhance cooperation between ventilation and steam conditions of turbine and ventilation condition relative to a superheater at the time of starting a plant using a fast breeder, and to enhance safety with respect to failure of heat transmission tubes at the time of start. Structure: In a device in which steam generated in an evaporator is fed to a high pressure turbine through a super-heater and an outlet steam of high pressure turbine is reheated by means of a re-heater and fed into a turbine on the side of low pressure to drive the turbine for power generation, opening and closing valves are mounted on outlet and inlet pipes, respectively, of the heat transmission pipe in the super heater, said outlet and inlet pipes being connected by a bypass pipe. Upstream side of the opening and closing valve on the inlet pipe and the downstream side of the opening and closing valve on the outlet pipe and connected by a bypass pipe in the re-heater and said bypass pipe in the re-heater is provided with a steam heat exchanger to be heated by steam in the outlet of the superheater, and a steam line in an auxiliary boiler is connected to the side of re-heater from the opening and closing valve on the heat transmission pipe in the re-heater. (Hanada, M.)

Numerical investigation of passive heat removal system via steam generator in VVER 1200

International Nuclear Information System (INIS)

Dinh Anh Tuan; Duong Thanh Tung; Tran Chi Thanh; Nguyen Van Thai

2015-01-01

Passive heat removal system (PHRS) via Steam Generator is an important part in VVER design. In case of Design Basic Accidents such as blackout, failure of feed water supply to steam generator or coolant leakage with failure of emergency core cooling at high pressure. PHRS is designed to remove the residual heat from reactor core through steam generator to heat exchanger which is placed outside reactor vessel. In order to evaluate the passive system, a numerical investigation using a CFD code is performed. However, PHRS has complex geometry for using CFD simulation. Thus, RELAP5 is applied to provide the wall heat flux of tube in the heat exchanger tank. The natural convection in the heat exchanger tank is investigated in this report. Numerical results show temperature and velocity distribution in the heat exchanger tank are calculated with different wall heat flux corresponding to various transient conditions. The calculated results contribute to the capacity analysis of passive heat removal system and giving valuable information for safe operation of VVER 1200. (author)

Steam generators and waste heat boilers for process and plant engineers

CERN Document Server

Ganapathy, V

2014-01-01

Incorporates Worked-Out Real-World ProblemsSteam Generators and Waste Heat Boilers: For Process and Plant Engineers focuses on the thermal design and performance aspects of steam generators, HRSGs and fire tube, water tube waste heat boilers including air heaters, and condensing economizers. Over 120 real-life problems are fully worked out which will help plant engineers in evaluating new boilers or making modifications to existing boiler components without assistance from boiler suppliers. The book examines recent trends and developments in boiler design and technology and presents novel idea

Specification of steam generator, condenser and regenerative heat exchanger materials for nuclear applications

International Nuclear Information System (INIS)

Jovasevic, J.V.; Stefanovic, V.M.; Spasic, Z.LJ.

1977-01-01

The basic standards specifications of materials for nuclear applications are selected. Seamless Ni-Cr-Fe alloy Tubes (Inconel-600) for steam generators, condensers and other heat exchangers can be employed instead of austenitic stainless steal or copper alloys tubes; supplementary requirements for these materials are given. Specifications of Ni-Cr-Fe alloy plate, sheet and strip for steam generator lower sub-assembly, U-bend seamless copper-alloy tubes for heat exchanger and condensers are also presented. At the end, steam generator channel head material is proposed in the specification for carbon-steel castings suitable for welding

Thermal performances of molten salt steam generator

International Nuclear Information System (INIS)

Yuan, Yibo; He, Canming; Lu, Jianfeng; Ding, Jing

2016-01-01

Highlights: • Thermal performances of molten salt steam generator were experimentally studied. • Overall heat transfer coefficient reached maximum with optimal molten salt flow rate. • Energy efficiency first rose and then decreased with salt flow rate and temperature. • Optimal molten salt flow rate and temperature existed for good thermal performance. • High inlet water temperature benefited steam generating rate and energy efficiency. - Abstract: Molten salt steam generator is the key technology for thermal energy conversion from high temperature molten salt to steam, and it is used in solar thermal power station and molten salt reactor. A shell and tube type molten salt steam generator was set up, and its thermal performance and heat transfer mechanism were studied. As a coupling heat transfer process, molten salt steam generation is mainly affected by molten salt convective heat transfer and boiling heat transfer, while its energy efficiency is also affected by the heat loss. As molten salt temperature increased, the energy efficiency first rose with the increase of heat flow absorbed by water/steam, and then slightly decreased for large heat loss as the absorbed heat flow still rising. At very high molten salt temperature, the absorbed heat flow decreased as boiling heat transfer coefficient dropping, and then the energy efficiency quickly dropped. As the inlet water temperature increased, the boiling region in the steam generator remarkably expanded, and then the steam generation rate and energy efficiency both rose with the overall heat transfer coefficient increasing. As the molten salt flow rate increased, the wall temperature rose and the boiling heat transfer coefficient first increased and then decreased according to the boiling curve, so the overall heat transfer coefficient first increased and then decreased, and then the steam generation rate and energy efficiency of steam generator both had maxima.

IAEA Technical Meeting on Innovative Heat Exchanger and Steam Generator Designs for Fast Reactors. Working Material

International Nuclear Information System (INIS)

2011-01-01

The IAEA, within the framework of its Nuclear Energy Department’s Technical Working Group on Fast Reactors (TWG-FR), assists Member States activities in fast reactors technology development areas by providing an umbrella for information exchange [topical Technical Meetings (TMs), Workshops and large Conferences] and collaborative R&D [Coordinated Research Projects (CRPs)]. The Technical meeting on “Innovative Heat Exchanger and Steam Generator Designs for Fast Reactors” was held from 21 – 22 December 2011 in Vienna, addressing Member States’ expressed needs of information exchange in the field of advanced fast reactor design features, with particular attention to innovative heat exchangers and steam generators. The Objective of the TM is to provide a global forum for in-depth information exchange and discussion on the most advanced concepts of heat exchangers and steam generators for fast reactors. More specifically, the objectives are: · Review of the status of advanced fast reactor development activities with special emphasis on design and performance of heat exchangers and steam generators; · Discuss requirements for innovative heat exchangers and steam generators; · Present results of studies and conceptual designs for innovative heat exchangers and steam generators; · Provide recommendations for international collaboration under the IAEA aegis. The meeting agenda of the meeting is in Annex I

Steam generators, turbines, and condensers. Volume six

International Nuclear Information System (INIS)

Anon.

1986-01-01

Volume six covers steam generators (How steam is generated, steam generation in a PWR, vertical U-tube steam generators, once-through steam generators, how much steam do steam generators make?), turbines (basic turbine principles, impulse turbines, reaction turbines, turbine stages, turbine arrangements, turbine steam flow, steam admission to turbines, turbine seals and supports, turbine oil system, generators), and condensers (need for condensers, basic condenser principles, condenser arrangements, heat transfer in condensers, air removal from condensers, circulating water system, heat loss to the circulating water system, factors affecting condenser performance, condenser auxiliaries)

Analysis of Decay Heat Removal by Natural Convection in LMR with a Combined Steam Generator

International Nuclear Information System (INIS)

Kim, Eui Kwang; Eoh, Jae Hyuk; Han, Ji Woong; Lee, Tae Ho

2011-01-01

Liquid metal reactors (LMRs) conventionally employ an intermediate heat transport system (IHTS) to protect the nuclear core during a sodium-water reaction (SWR) event. However these SWR-related components increase plant construction costs. In order to eliminate the need for an IHTS, a combined steam generator, which is an integrated heat exchanger of a steam generator and intermediate heat exchanger (IHX), was proposed by the Korea Atomic Energy Research Institute (KAERI). The objective of this work is to analyze the natural circulation heat removal capability of the rector system using a combined steam generator. As a means of decay heat removal, a normal heat transport path is composed of a primary sodium system, intermediate lead-bismuth circuit combined with SG and steam/water system. This paper presents the results of the possible temperature and natural circulation flows in all circuits during a steady state for a given reactor power level varied as a function of time

Thermal hydraulic studies in steam generator test facility

International Nuclear Information System (INIS)

Vinod, V.; Suresh Kumar, V.A.; Noushad, I.B.; Ellappan, T.R.; Rajan, K.K.; Rajan, M.; Vaidyanathan, G.

2005-01-01

Full text of publication follows: A 500 MWe fast breeder reactor is being constructed at Kalpakkam, India. This is a sodium cooled reactor with two primary and two secondary sodium loops with total 8 steam generators. The typical advantage of fast breeder plants is the high operating temperature of steam cycles and the high plant efficiency. To produce this high pressure and high temperature steam, once through straight tube vertical sodium heated steam generators are used. The steam is generated from the heat produced in the reactor core and being transported through primary and secondary sodium circuits. The steam generator is a 25 m high middle supported steam generator with expansion bend and 23 m heat transfer length. Steam Generator Test Facility (SGTF) constructed at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam aims at performing various tests on a 5.5 MWt steam generator. This vertically simulated test article is similar in all respects to the proposed 157 MWt steam generator module for the Prototype Fast Breeder Reactor (PFBR), with reduced number of tubes. Heat transfer performance tests are done with this 19 tube steam generator at various load conditions. Sodium circuit for the SGTF is equipped with oil fired heater as heat source and centrifugal sodium pump, to pump sodium at 105 m 3 /hr flow rate. Other typical components like sodium to air heat exchanger, sodium purification system and hydrogen leak detection system is also present in the sodium circuit. High pressure steam produced in the steam generator is dumped in a condenser and recycled. Important tests planned in SGTF are the heat transfer performance test, stability test, endurance test and performance test of steam generator under various transients. The controlled operation of steam generator will be studied with possible control schemes. A steady state simulation of the steam generator is done with a mathematical model. This paper gives the details of heat transfer

Vertical steam generator

International Nuclear Information System (INIS)

Cuda, F.; Kondr, M.; Kresta, M.; Kusak, V.; Manek, O.; Turon, S.

1982-01-01

A vertical steam generator for nuclear power plants and dual purpose power plants consists of a cylindrical vessel in which are placed heating tubes in the form upside-down U. The heating tubes lead to the jacket of the cylindrical collector placed in the lower part of the steam generator perpendicularly to its vertical axis. The cylindrical collector is divided by a longitudinal partition into the inlet and outlet primary water sections of the heating tubes. One ends of the heating tube leads to the jacket of the collector for primary water feeding and the second ends of the heating tubes into the jacket of the collector which feeds and offtakes primary water from the heating tubes. (B.S.)

Stress analysis of HTR-10 steam generator heat exchanging tubes

International Nuclear Information System (INIS)

Dong Jianling; Zhang Xiaohang; Yin Dejian; Fu Jiyang

2001-01-01

Steam Generator (SG) heat exchanging tubes of 10 MW High Temperature Gas Cooled Reactor (HTR-10) are protective screens between the primary loop of helium with radioactivity and the secondary loop of feeding water and steam without radioactivity. Water and steam will enter into the primary loop when rupture of the heat exchanging tubes occurs, which lead to increase of the primary loop pressure and discharge of radioactive materials. Therefore it is important to guarantee the integrity of the tubes. The tube structure is spiral tube with small bending radius, which make it impossible to test with volumetric in-service detection. For such kind of spiral tube, using LBB concept to guarantee the integrity of the tubes is an important option. The author conducts stress analysis and calculation of HTR-10 SG heat exchanging tubes using the FEM code of piping stress analysis, PIPESTRESS. The maximum stress and the dangerous positions are obtained

High temperature technological heat exchangers and steam generators with helical coil assembly tube bundle

International Nuclear Information System (INIS)

Korotaev, O.J.; Mizonov, N.V.; Nikolaevsky, V.B.; Nazarov, E.K.

1990-01-01

Analysis of thermal hydraulics characteristics of nuclear steam generators with different tube bundle arrangements and waste heat boilers for ammonia production units was performed on the basis of operating experience results and research and development data. The present report involves the obtained information. The estimations of steam generator performances and repair-ability are given. The significant temperature profile of the primary and secondary coolant flows are attributed to all steam generator designs. The intermediate mixing is found to be an effective means of temperature profile overcoming. At present the only means to provide an effective mixing in heat exchangers of the following types: straight tubes, field tubes, platen tubes and multibank helical coil tubes (with complicated bend distribution along their length) are section arrangements in series in conjunction with forced and natural mixing in connecting lines. Development of the unificated system from mini helical coil assemblies allows to design and manufacture heat exchangers and steam generators within the wide range of operating conditions without additional expenses on the research and development work

LMFBR steam generator development: duplex bayonet tube steam generator. Volume II

International Nuclear Information System (INIS)

DeFur, D.D.

1975-04-01

This report represents the culmination of work performed in fulfillment of ERDA Contract AT(11-1)-2426, Task Agreement 2, in which alternate steam generator designs were developed and studied. The basic bayonet tube generator design previously developed by C-E under AEC Contract AT(11-1)-3031 was expanded by incorporating duplex heat transfer tubes to enhance the unit's overall safety and reliability. The effort consisted of providing and evaluating conceptual designs of the evaporator, superheater and reheater components for a large plant LMFBR steam generator (950 MWt per heat transport loop)

Diagnostic system of steam generator, especially molten metal heated steam generator

International Nuclear Information System (INIS)

Matal, O.; Martoch, J.

1986-01-01

A diagnostic system is described and graphically represented consisting of a leak detector, a medium analyzer and sensors placed on the piping connected to the indication sections of both tube plates. The advantage of the designed system consists in the possibility of detecting tube failure immediately on leak formation, especially in generators with duplex tubes. This shortens the period of steam generator shutdown for repair and reduces power losses. The design also allows to make periodical leak tests during planned steam generator shutdowns. (A.K.)

Test and evaluation of Alco/BLH prototype sodium-heated steam generator. Final report

Energy Technology Data Exchange (ETDEWEB)

Kaplan, C.J.; Auge, L.J.; Cho, S.M.; Hanna, R.W.; Prevost, J.R.; Steger, N.A.; Wagner, R.K.

1971-01-31

A 30-Mwt prototype sodium-to-sodium intermediate heat exchanger and a 30-Mwt prototype sodium-heated steam generator were tested in combined operation in its Sodium Components Test Installation. This report contains the results of test and evaluation of the steam generator. During plant performance tests, performance degradation was observed, which resulted in the initiation of a diagnostic test series. This test series revealed that under certain operating conditions, the thermohydraulic characteristic of the steam generator changed either suddenly or gradually, resulting in overall performance degradation. A structural failure, requiring retirement of the unit, occurred before the diagnostic test series and analytical support effort were completed. This report describes the thermohydraulic and structural performance, including the structural failures, and related evaluation analyses of the Alco/BLH prototype steam generator performed prior to termination of the test and evaluation program. In addition, the report presents a post-test examination plan to obtain data that could possibly explain the cause of performance anomalies and structural failures experienced during testing.

A methodology for the geometric design of heat recovery steam generators applying genetic algorithms

International Nuclear Information System (INIS)

Durán, M. Dolores; Valdés, Manuel; Rovira, Antonio; Rincón, E.

2013-01-01

This paper shows how the geometric design of heat recovery steam generators (HRSG) can be achieved. The method calculates the product of the overall heat transfer coefficient (U) by the area of the heat exchange surface (A) as a function of certain thermodynamic design parameters of the HRSG. A genetic algorithm is then applied to determine the best set of geometric parameters which comply with the desired UA product and, at the same time, result in a small heat exchange area and low pressure losses in the HRSG. In order to test this method, the design was applied to the HRSG of an existing plant and the results obtained were compared with the real exchange area of the steam generator. The findings show that the methodology is sound and offers reliable results even for complex HRSG designs. -- Highlights: ► The paper shows a methodology for the geometric design of heat recovery steam generators. ► Calculates product of the overall heat transfer coefficient by heat exchange area as a function of certain HRSG thermodynamic design parameters. ► It is a complement for the thermoeconomic optimization method. ► Genetic algorithms are used for solving the optimization problem

Heat exchanger, particularly liquid sodium heated steam generator

International Nuclear Information System (INIS)

Robin, Marcel; Tillequin, Jean.

1977-01-01

This invention relates to a liquid sodium heated steam generator the characteristic of which is an annular distribution chamber fed by two independent and diametrically opposed manifolds on a common horizontal axis, issuing respectively into two adjacent compartments made in the chambers on both sides of a vertical transversal partition containing the axis of the casing and extending perpendicularly to the manifolds, each compartment being itself divided into a number of adjacent sectors marked by folded metal sheets fixed to the distributor and shaped so as to present in pairs and with the chamber opposite the manifold issuing into a compartment two independent ducts for distributing the sodium flow [fr

International examples of steam generator replacement

International Nuclear Information System (INIS)

Wiechmann, K.

1993-01-01

Since 1979-1980 a total of twelve nuclear power plants world-wide have had their steam generators replaced. The replacement of the Combustion steam generators in the Millstone-2 plant in the United States was completed very recently. Steam generator replacement activities are going on at present in four plants. In North Anna, the steam generators have been under replacement since January 1990. In Japan, preparations have been started for Genkai-1. Since January 1992, the two projects in Beznau-1, Switzerland, and Doel-3, Belgium, have bee planned and executed in parallel. Why steam generator replacement? There are a number of defect mechanisms which give rise to the need for early steam generator replacement. One of the main reasons is the use of Inconel-600 as material for the heating tubes. Steam generator heating tubes made of Inconel-600 have been known to exhibit their first defects due to stress corrosion cracking after less than one year of operation. (orig.) [de

Temperature condition in decreasing heat transfer zone for NPP steam generators

International Nuclear Information System (INIS)

Kudryavtsev, I.S.; Paskar', B.L.; Sudakov, A.V.

1985-01-01

An experimental set-up is described and the results of temperature pulsation investigation are presented for coil steam generating channel surfaces of the NPP helium and sodium cooled HTGR. The investigations are carried out at the heat flux density of 350-900 kW/m 3 , the mass rate of 350-2000 kg/(m 2 Xs), the pressUre of 15 MPa. Temperature pulsations occur due to instability of heat transfer in the near-wall region. The results show that the critical region of burnout has a local character. Pulsation dependences on operating conditions are given. The required resource for the steam generating channel may be provided by chosing the ratio of heat flux to the mass rate, the ratio being equal to 0.5 kJ/kg for the channel with the internal diameter of 19 mm, made of the 12Kh2M steel, the wall thickness of 3 mm. In this case the maximum span of temperature pulsations doesn't exceed 25-30 K

Future aspects for liquid metal heated steam generators

International Nuclear Information System (INIS)

Jansing, W.; Ratzel, W.; Vinzens, K.

1975-01-01

The present status of steam generators is shown. The experience gained until now is expressed in form of basic points. The most important design criteria for steam generator systems are outlined. On the basis of these design criteria, two possible steam generator concepts are shown. Costs in relationship to the repair concepts of two modular steam generators (thermal output 156 and 625 MW) and a pool design of 625 MW are compared. (author)

Future aspects for liquid metal heated steam generators

Energy Technology Data Exchange (ETDEWEB)

Jansing, W; Ratzel, W; Vinzens, K

1975-07-01

The present status of steam generators is shown. The experience gained until now is expressed in form of basic points. The most important design criteria for steam generator systems are outlined. On the basis of these design criteria, two possible steam generator concepts are shown. Costs in relationship to the repair concepts of two modular steam generators (thermal output 156 and 625 MW) and a pool design of 625 MW are compared. (author)

Numerical simulation of dryout and post-dryout heat transfer in a straight-pipe once-through steam generator

International Nuclear Information System (INIS)

Shi, Jianxin; Sun, Baozhi; Han, Wenjing; Zhang, Guolei; Li, Yanjun; Yang, Longbin

2016-01-01

Highlights: • Two-fluid three-flow-field model is developed to predict dryout in steam generator. • The empirical correlation is used to correct dryout criterion. • The interactions between three-flow-fields and the wall are considered. • Dryout and post-dryout heat transfer mechanisms are discussed through the results. - Abstract: Accurately predicting dryout and post-dryout heat transfer characteristics is critical for proper design of once-through steam generators. This paper provides a reasonable and simple method for this prediction by introducing a two-fluid, three-flow-field mathematical model and improving the dryout criterion-critical quality, and conducts a numerical simulation of dryout and post-dryout heat transfer in a once-through steam generator to prove the model’s performance. The results show that the critical quality in a once-through steam generator is about 0.82, with the heat transfer capacity significantly reducing and the wall temperature sharply increasing in a non-linear form by approximately 30 K when dryout occurs. Part of the steam is superheated in the post-dryout region, resulting in a deviation from thermodynamic equilibrium between the vapor and liquid phases. Dryout and post-dryout heat transfer in the once-through steam generator operate between complete deviation from thermodynamic equilibrium and complete thermodynamic equilibrium. Therefore, the presence of droplets has a significant influence on the mass, momentum and energy transfer between the film and vapor phases.

Theoretical study on a multivariate feedback control of a sodium-heated steam generator

International Nuclear Information System (INIS)

Takahashi, R.; Maruyama, Y.; Oikawa, T.

1984-01-01

This paper applies the connection of a multivariate feedback controller with a state estimator to a 1-MW sodium-heated steam generator for LMFBR theoretically, to obtain a control strategy which emphasizes, from the view point of safety and availability of the FBR plant, that a superheat of 30 0 C should be required for the evaporator steam. This involves a trial to study the feasibility for the estimation of such an inaccessible variable as the dry-out location of tubes and utilize the state estimate to design a feedback controller of steam generators. The Kalman filter tested was found to generate reasonable estimates of the transient process variables of the steam generator and can provide a major advantage of regulating steam condition of the system even in the presence of contamination by a rather high level of measurement noise in the view point of economic uses of micro- and/or minicomputers. (orig.)

Evaluation of piping heat transfer, piping flow regimes, and steam generator heat transfer for the Semiscale Mod-1 isothermal tests

International Nuclear Information System (INIS)

French, R.T.

1975-08-01

Selected experimental data pertinent to piping heat transfer, transient fluid flow regimes, and steam generator heat transfer obtained during the Semiscale Mod-1 isothermal blowdown test series (Test Series 1) are analyzed. The tests in this first test series were designed to provide counterparts to the LOFT nonnuclear experiments. The data from the Semiscale Mod-1 intact and broken loop piping are evaluated to determine the surface heat flux and average heat transfer coefficients effective during the blowdown transient and compared with well known heat transfer correlations used in the RELAP4 computer program. Flow regimes in horizontal pipe sections are calculated and compared with data obtained from horizontal and vertical densitometers and with an existing steady state flow map. Effects of steam generator heat transfer are evaluated quantitatively and qualitatively. The Semiscale Mod-1 data and the analysis presented in this report are valuable for evaluating the adequacy and improving the predictive capability of analytical models developed to predict system response to piping heat transfer, piping flow regimes, and steam generator heat transfer during a postulated loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). 16 references. (auth)

Selection of values of design peak heat flux to reduce the risk of waterside corrosion in F.R. steam generators

Energy Technology Data Exchange (ETDEWEB)

Bolt, P R; Garnsey, R

1975-07-01

Attention is drawn to the high levels of peak heat Flux that can exist in sodium heated steam generators. The strength of the relationship between heat, flux and both deposition rate and the concentration of salts is discussed. Relevant steam generator operational experience obtained on the C.E.G.B. system is described and tentative proposals are made for limits to he to the peak heat flux values used in F.R. steam generator design. (author)

Selection of values of design peak heat flux to reduce the risk of waterside corrosion in F.R. steam generators

International Nuclear Information System (INIS)

Bolt, P.R.; Garnsey, R.

1975-01-01

Attention is drawn to the high levels of peak heat Flux that can exist in sodium heated steam generators. The strength of the relationship between heat, flux and both deposition rate and the concentration of salts is discussed. Relevant steam generator operational experience obtained on the C.E.G.B. system is described and tentative proposals are made for limits to he to the peak heat flux values used in F.R. steam generator design. (author)

Critical heat flux and transition boiling characteristics for a sodium-heated steam generator tube for LMFBR applications

Energy Technology Data Exchange (ETDEWEB)

Wolf, S.; Holmes, D.H.

1977-04-01

An experimental program was conducted to characterize critical heat flux (CHF) in a sodium-heated steam generator tube model at a proposed PLBR steam generator design pressure of 7.2 MPa. Water was circulated vertically upward in the tube and the heating sodium was flowing counter-current downward. The experimental ranges were: mass flux, 110 to 1490 kg/s.m/sup 2/ (0.08 to 1.10 10/sup 6/ lbm/h.ft/sup 2/); critical heat flux, 0.16 to 1.86 MW/m/sup 2/ (0.05 to 0.59 10/sup 6/ Btu/h.ft/sup 2/); and critical quality, 0.48 to 1.0. The CHF phenomenon for the experimental conditions is determined to be dryout as opposed to departure from nucleate boiling (DNB). The data are divided into high- and low-mass flux regions.

Liquid metal steam generator

International Nuclear Information System (INIS)

Wolowodiuk, W.

1975-01-01

A liquid metal heated steam generator is described which in the event of a tube failure quickly exhausts out of the steam generator the products of the reaction between the water and the liquid metal. The steam is generated in a plurality of bayonet tubes which are heated by liquid metal flowing over them between an inner cylinder and an outer cylinder. The inner cylinder extends above the level of liquid metal but below the main tube sheet. A central pipe extends down into the inner cylinder with a centrifugal separator between it and the inner cylinder at its lower end and an involute deflector plate above the separator so that the products of a reaction between the liquid metal and the water will be deflected downwardly by the deflector plate and through the separator so that the liquid metal will flow outwardly and away from the central pipe through which the steam and gaseous reaction products are exhausted. (U.S.)

Model studies of the vertical steam generator thermal-hydraulic characteristics

International Nuclear Information System (INIS)

Desyatun, V.F.; Moskvichev, V.F.; Ulasov, V.M.; Morozov, V.G.; Burkov, V.K.; Grebennikov, V.N.

1984-01-01

Results of investigations conducted to clarify the calculation technique and to test the workability of the main elements and units of the PGV-250 vertical steam generator of saturated steam are considered. The steam generating capacity of the plant is 1486 t/h, thermal power is 792 MW. Steam generation follows a multiple circulation scheme. The heat surface comprises 330-shields. The investigations are carried out with a model which reproduces all the main elements of the steam generator xcluding the economizer section. The flow rates of feed water, generated steam and coolant of the first circuit as well as temperature, pressure and humidity of the generated steam past the separator are determined. The average heat transfer factors of the heat surface are calculated on the base of the data obtained and a conclusion is drawn on the correctness of the thermohydraulic calculation technique used in development of the PGV-250 steam generator design. Temperature pulsations and heat surface steaming are not observed. The steam humidity at the outlet and steam capture into sinking tubes are within permissible values

Scale model test results for an inverted U-tube steam generator with comparisons to heat transfer correlations

International Nuclear Information System (INIS)

Boucher, T.J.

1987-01-01

To provide data for assessment and development of thermal-hydraulic computer codes, bottom main feedwater-line-break transient simulations were performed in a scale model (Semiscale Mod-2C) of a pressurized water reactor (PWR) with conditions typical of a PWR (15.0 MPa primary pressure, 600 K steam generator inlet plenum fluid temperatures, 6.2 MPa secondary pressure). The state-of-the-art measurements in the scale model (Type III) steam generator allow for the determination of U-tube steam generator allow for the determination of U-tube steam generator secondary component interactions, tube bundle local radial heat transfer, and tube bundle and riser vapor void fractions for steady state and transient operations. To enhance the understanding of the observed phenomena, the component interactions, local heat fluxes, local secondary convective heat transfer coefficients and local vapor void fractions are discussed for steady state, full-power and transient operations. Comparisons between the measurement-derived secondary convective heat transfer coefficients and those predicted by a number of correlations, including the Chen correlation currently used in thermal-hydraulic computer codes, show that none of the correlations adequately predict the data and points out the need for the formulation of a new correlation based on this experimental data. The unique information presented herein should be of the interest to anyone involved in modeling inverted U-tube steam generator thermal-hydraulics for forced convection boiling/vaporization heat transfer. 5 refs., 13 figs., 1 tab

Testing installation for a steam generator

International Nuclear Information System (INIS)

Dubourg, M.

1985-01-01

The invention proposes a testing installation for a steam generator associated to a boiler, comprising a testing exchanger connected to a feeding circuit in secondary fluid and to a circuit to release the steam produced, and comprising a heating-tube bundle connected to a closed circuit of circulation of a primary coolant at the same temperature and at the pressure than the primary fluid. The heating-tube bundle of the testing exchanger has the same height than the primary bundle of the steam generator and the testing exchanger is at the same level and near the steam generator and is fed by the same secondary fluid such as it is subject to the same operation phases during a long period. The in - vention applies, more particularly, to the steam generators of pressurized water nuclear power plants [fr

Steam generation device with heat exchange between a liquid metal coolant and the feedwater

International Nuclear Information System (INIS)

Malaval, C.

1983-01-01

The invention is particularly applicable to a liquid metal fast breeder reactor plant, the liquid metal being sodium. The steam generation device is described in detail, it allows to get an upper liquid metal level without turbulence and an easier passage for the shock wave towards the steam generator up to the liquid metal level without being laterally reflected back to the intermediate heat exchangers [fr

Experimental investigation of reflux condensation heat transfer in PWR steam generator tubes in the presence of noncondensible gases

Energy Technology Data Exchange (ETDEWEB)

Vierow, Karen; Wu, Tiejun [Purdue Univ., West Lafayette (United States); Nagae, Takashi [Institute of Nuclear Safety System, Tokyo (Japan)

2003-07-01

Under certain circumstances in a Pressurized Water Reactor (PWR), the coolant system may be in a partially drained state and reflux condensation in the steam generator U-tubes can be the major heat removal mechanism. Noncondensable gases may be present and would degrade the heat transfer rate. If heat removal rates are insufficient, this situation could lead to core boil-off, fuel rod heatup, and eventually core damage. The Institute of Nuclear Safety System, Inc. (INSS) and the Nuclear Heat Transfer Systems Laboratory at Purdue University have begun a cooperative research program to investigate the effectiveness of reflux condensation in PWR steam generator U-tubes in the presence of noncondensable gases. The final objectives are to provide local heat transfer data for development of methods to analyze reflux condensation in PWR steam generator U-tubes and to investigate the potential for flooding. Key features of the experimental data reported herein are that they are local data under laminar steam/gas mixture and condensate film flow and they are taken from a test section with dimensions similar to an actual steam generator tube. Steady state data were obtained under various steam and air inlet flow rates and pressures. The data show the significant degrading effect of noncondensable gas on heat transfer coefficients. From the data, correlations for the reflux condensation local heat transfer coefficient and the local Nusselt number under laminar conditions were derived. These experiments are providing essential and unique fundamental data for development of methods to analyze reflux condensation.

Effect of heat- and steam-generating sheet on daily activities of living in patients with osteoarthritis of the knee: randomized prospective study.

Science.gov (United States)

Seto, Hiroaki; Ikeda, Hiroshi; Hisaoka, Hidehiko; Kurosawa, Hisashi

2008-05-01

Thermotherapy is widely known to be effective for osteoarthritis of the knee (knee OA), but most treatment methods make use of dry heat. We developed a sheet that generates heat and steam simultaneously. In this prospective randomized study, we evaluated the effectiveness of this sheet. Of 41 female patients with knee OA randomized to use the heat/steam-generating sheet or the dry heat-generating sheet, 37 patients (20 using the heat/steam-generating sheet and 17 using the dry heat-generating sheet) who used the sheets continuously for 4 weeks were studied. Outcome measures included the Western Ontario and McMaster Universities Osteoarthritis Index (WOMAC) and Japan Orthopaedic Association (JOA) scores, which were applied at baseline and after 2 and 4 weeks of use. Significant improvement of the total WOMAC score was observed at 2 and 4 weeks (compared to baseline) in the heat/steam-generating sheet group, but no significant change was observed in the dry heat-generating sheet group. Among the JOA scores, the gait ability score was also improved significantly only in the heat/steam-generating sheet group. The effects were still seen 6 weeks after completion of treatment. The present study provided evidence that the heat/steam-generating sheet that we developed is effective for alleviating pain and is especially superior in regard to improving stiffness and gait impairment in patients with knee OA. Furthermore, the effect persists for at least 6 weeks after application.

Thermodynamic performance analysis and algorithm model of multi-pressure heat recovery steam generators (HRSG) based on heat exchangers layout

International Nuclear Information System (INIS)

Feng, Hongcui; Zhong, Wei; Wu, Yanling; Tong, Shuiguang

2014-01-01

Highlights: • A general model of multi-pressure HRSG based on heat exchangers layout is built. • The minimum temperature difference is introduced to replace pinch point analysis. • Effects of layout on dual pressure HRSG thermodynamic performances are analyzed. - Abstract: Changes of heat exchangers layout in heat recovery steam generator (HRSG) will modify the amount of waste heat recovered from flue gas; this brings forward a desire for the optimization of the design of HRSG. In this paper the model of multi-pressure HRSG is built, and an instance of a dual pressure HRSG under three different layouts of Taihu Boiler Co., Ltd. is discussed, with specified values of inlet temperature, mass flow rate, composition of flue gas and water/steam parameters as temperature, pressure etc., steam mass flow rate and heat efficiency of different heat exchangers layout of HRSG are analyzed. This analysis is based on the laws of thermodynamics and incorporated into the energy balance equations for the heat exchangers. In the conclusion, the results of the steam mass flow rate, heat efficiency obtained for three heat exchangers layout of HRSGs are compared. The results show that the optimization of heat exchangers layout of HRSGs has a great significance for waste heat recovery and energy conservation

Economic impact of latent heat thermal energy storage systems within direct steam generating solar thermal power plants with parabolic troughs

International Nuclear Information System (INIS)

Seitz, M.; Johnson, M.; Hübner, S.

2017-01-01

Highlights: • Integration of a latent heat thermal energy storage system into a solar direct steam generation power cycle. • Parametric study of solar field and storage size for determination of the optimal layout. • Evaluation of storage impact on the economic performance of the solar thermal power plant. • Economic comparison of new direct steam generation plant layout with state-of-the-art oil plant layout. - Abstract: One possible way to further reduce levelized costs of electricity of concentrated solar thermal energy is to directly use water/steam as the primary heat transfer fluid within a concentrated collector field. This so-called direct steam generation offers the opportunity of higher operating temperatures and better exergy efficiency. A technical challenge of the direct steam generation technology compared to oil-driven power cycles is a competitive storage technology for heat transfer fluids with a phase change. Latent heat thermal energy storages are suitable for storing heat at a constant temperature and can be used for direct steam generation power plants. The calculation of the economic impact of an economically optimized thermal energy storage system, based on a latent heat thermal energy storage system with phase change material, is the main focus of the presented work. To reach that goal, a thermal energy storage system for a direct steam generation power plant with parabolic troughs in the solar field was thermally designed to determine the boundary conditions. This paper discusses the economic impact of the designed thermal energy storage system based on the levelized costs of electricity results, provided via a wide parametric study. A state-of-the-art power cycle with a primary and a secondary heat transfer fluid and a two-tank thermal energy storage is used as a benchmark technology for electricity generation with solar thermal energy. The benchmark and direct steam generation systems are compared to each other, based respectively

On the method of heat exchange calculation for critical and postcritical regimes in sodium-water steam generators

International Nuclear Information System (INIS)

Khudasko, V.V.; Kardash, D.Yu.; Grachev, N.S.

1986-01-01

Technique for calculating heat exchange in sodium-water steam generators with provisions for steam-water flow non-equilibrium character and moisture additional evaporation in pipes is suggested. Zone of heat exchange crisis representing the zone of transition from developed boiling to postcritical zone is considered. Comparison of estimated and experimental data performed for the following ranges of steame generator parameters: pressure p=7.8-14.0 MPa, coolant flow rate ρw=350-1000 kg/(m 2 xs), inlet sodium temperature T=590-825 K shows their good agreement

Implications of small water leak reactions on sodium heated steam generator design

Energy Technology Data Exchange (ETDEWEB)

Smedley, J A

1975-07-01

Various types of sodium water reactions have been looked on as possibly causing hazard conditions in sodium heated steam generator units ranging from the very improbable boiler tube double ended guillotine fracture to the almost certain occurrence of micro-leaks. Within this range small water leaks reactions have attracted particular interest and the present paper looks at the principles of associating the reactions with detection and protection systems for Commercial Fast Reactors. A method is developed for assessing whether adequate protection has been provided against the effects of small water leak reactions in a steam generator unit. (author)

Investigations of the gas-side heat transfer and flow characteristics of steam generators in AGR stations

Energy Technology Data Exchange (ETDEWEB)

Lis, J [Central Electricity Research Laboratories, Leatherhead, Surrey (United Kingdom)

1984-07-01

This paper describes the experimental and analytical investigations of the gas-side heat transfer and flow characteristics of steam generators in the AGR stations carried out by CERL. The majority of the experimental work on heat transfer and flow characteristics of close-packed tube arrangements in cross-flow of gases is carried out in a pressurised heat exchanger rig. The rig is operated on-line by a dedicated PDP 11/40 computer over the range of Reynolds number 10{sup 4} to 3x10{sup 5}. Atmospheric wind tunnels employing either small or large scale models of the specific sections of steam generators are used for a variety of supplementary and development studies. Various measurements techniques and, in particular, LDA and hot wire anemometry employed in these studies are described. The more important aspects of various investigations are illustrated by typical results. In order to ensure the efficient operation and integrity of steam generators under asymmetric boundary conditions a MIX suite of 2-dimensional codes has been developed. The codes calculate the gas and water/steam flow and temperature distributions in each channel of the steam generator taking into account thermal mixing in the gas as it passes through the generator. Application of the MIX codes to the solution of various operational problems is illustrated by typical examples and the continuing exercise of validating the codes against plant operational data is discussed. (author)

Investigations of the gas-side heat transfer and flow characteristics of steam generators in AGR stations

International Nuclear Information System (INIS)

Lis, J.

1984-01-01

This paper describes the experimental and analytical investigations of the gas-side heat transfer and flow characteristics of steam generators in the AGR stations carried out by CERL. The majority of the experimental work on heat transfer and flow characteristics of close-packed tube arrangements in cross-flow of gases is carried out in a pressurised heat exchanger rig. The rig is operated on-line by a dedicated PDP 11/40 computer over the range of Reynolds number 10 4 to 3x10 5 . Atmospheric wind tunnels employing either small or large scale models of the specific sections of steam generators are used for a variety of supplementary and development studies. Various measurements techniques and, in particular, LDA and hot wire anemometry employed in these studies are described. The more important aspects of various investigations are illustrated by typical results. In order to ensure the efficient operation and integrity of steam generators under asymmetric boundary conditions a MIX suite of 2-dimensional codes has been developed. The codes calculate the gas and water/steam flow and temperature distributions in each channel of the steam generator taking into account thermal mixing in the gas as it passes through the generator. Application of the MIX codes to the solution of various operational problems is illustrated by typical examples and the continuing exercise of validating the codes against plant operational data is discussed. (author)

Model of reverse steam generator

International Nuclear Information System (INIS)

Malasek, V.; Manek, O.; Masek, V.; Riman, J.

1987-01-01

The claim of Czechoslovak discovery no. 239272 is a model designed for the verification of the properties of a reverse steam generator during the penetration of water, steam-water mixture or steam into liquid metal flowing inside the heat exchange tubes. The design may primarily be used for steam generators with a built-in inter-tube structure. The model is provided with several injection devices configured in different heat exchange tubes, spaced at different distances along the model axis. The design consists in that between the pressure and the circumferential casings there are transverse partitions and that in one chamber consisting of the circumferential casings, pressure casing and two adjoining partitions there is only one passage of the injection device through the inter-tube space. (Z.M.). 1 fig

Experimental investigations on waterside corrosion problems of sodium-heated steam generators

International Nuclear Information System (INIS)

Buescher, E.; Haubold, W.; Jansing, W.; Vinzens, K.

1975-01-01

Based on experimental results, it is shown that the formation of a protective magnetite layer in sodium-heated steam generators - operated with neutral demineralized water - proceeds very fast. The stability of the magnetite layer is excellent even at sudden load changes and at the start-up or shut-down operation. (author)

Experimental investigations on waterside corrosion problems of sodium-heated steam generators

Energy Technology Data Exchange (ETDEWEB)

Buescher, E; Haubold, W; Jansing, W; Vinzens, K

1975-07-01

Based on experimental results, it is shown that the formation of a protective magnetite layer in sodium-heated steam generators - operated with neutral demineralized water - proceeds very fast. The stability of the magnetite layer is excellent even at sudden load changes and at the start-up or shut-down operation. (author)

Monitoring method for steam generator operation

International Nuclear Information System (INIS)

Tamaoki, Tetsuo

1991-01-01

In an LMFBR plant having an once-through steam generator, reduction of life of a heat transfer pipe caused by heat cycle fatigue is monitored by early finding for the occurrence of abnormality in the inside of the steam generator and by continuous monitoring for the position of departure from nucleate boiling (DNB), which are difficult with existent static characteristic analysis codes. That is, RMS values of fluctuations in temperature signals sent from thermocouples for measuring the fluid temperature in the vicinity of heat transfer pipe disposed along a primary channel of the once-through type steam generator. The abnormality in heat transfer performance is monitored by the distribution change of the RMS values. Subsequently, DNB point on the side of water and steam is determined by the distribution of the RMS value. Then, accumulated values of the product between the time in which the starting point stays in the DNB region and a life consumption amount per unit time given in accordance with the operation condition are monitored. Accordingly, thermal fatigue failure of the heat transfer pipe due to temperature fluctuation in the DNB region is monitored. (I.S.)

Tritium permeation through helium-heated steam generators of ceramic breeder blankets for DEMO

International Nuclear Information System (INIS)

Fuetterer, M.A.; Raepsaet, X.; Proust, E.

1994-01-01

The specifications of permeation barriers, tritium recovery process maintaining a very low tritium activity in the coolant, and control of the coolant chemistry, required the evaluation of the tritium losses through the steam generators and include the definition of its operating conditions by thermodynamic cycle calculations and its thermal-hydraulic design. For both tasks specific computer tools were developed. The obtained geometry, surface area, and temperature profiles along the heat exchanger tubes were then used to estimate the daily tritium permeation into the steam cycle. Steam oxidized Incoloy 800 austenitic stainless steel was identified as the best suited existing material; in nominal steady-state operation, the tritium escape into the steam cycle could be restricted to less than 10 Ci/d. Tritium permeation during temperature and pressure transients in the steam generator (destruction and possible self-healing of the permeation barrier) is identified to bear a large tritium release potential. Solutions are proposed. (from authors). 4 figs., 1 tab

Tritium permeation through helium-heated steam generators of ceramic breeder blankets for DEMO

Energy Technology Data Exchange (ETDEWEB)

Fuetterer, M A; Raepsaet, X; Proust, E

1994-12-31

The specifications of permeation barriers, tritium recovery process maintaining a very low tritium activity in the coolant, and control of the coolant chemistry, required the evaluation of the tritium losses through the steam generators and include the definition of its operating conditions by thermodynamic cycle calculations and its thermal-hydraulic design. For both tasks specific computer tools were developed. The obtained geometry, surface area, and temperature profiles along the heat exchanger tubes were then used to estimate the daily tritium permeation into the steam cycle. Steam oxidized Incoloy 800 austenitic stainless steel was identified as the best suited existing material; in nominal steady-state operation, the tritium escape into the steam cycle could be restricted to less than 10 Ci/d. Tritium permeation during temperature and pressure transients in the steam generator (destruction and possible self-healing of the permeation barrier) is identified to bear a large tritium release potential. Solutions are proposed. (from authors). 4 figs., 1 tab.

Corrosion aspects in steam generators of nuclear power plants

International Nuclear Information System (INIS)

Visoni, E.; Santos Pinto, M. dos

1988-01-01

Steam generators of pressurized water reactors (PWR), transfer heat from a primary coolant system to a secondary coolant system. Primary coolant water is heated in the core and passes through the steam generator that transfer heat to the secondary coolant water. However, the steam generator is dead for ionic impurities, corrosion products and fabrication/maintenence residues. These impurities concentrate between crevice and cracks. Many types of degradation mechanisms affect the tubes. The tubes are dented, craked, ovalized, wasted, etc. This paper describes the main corrosion problems in steam generators and includes the corrective actions to considered to reduce or eliminate these corrosion problems. (author) [pt

Main results of assessing integrity of RNPP-3 steam generator heat exchange tubes in accident management

International Nuclear Information System (INIS)

Shugajlo, Al-j P.; Mustafin, M.A.; Shugajlo, Al-r P.; Ryzhov, D.I.; Zhabin, O.I.

2017-01-01

Tubes integrity evaluation under accident conditions considering drain of SG and current technical state of steam exchange tubes is an important question regarding SG long-term operation and improvement of accident management strategy.The main investigation results prepared for heat exchange surface of RNPP-3 steam generator are presented in this research aimed at assessing integrity of heat exchange tubes under accident conditions, which lead to full or partial drain of heat exchange surface, in particular during station blackout.

An evaluation of analytical heat transfer area with various boiling heat transfer correlations in steam generator thermal sizing

International Nuclear Information System (INIS)

Jung, B. R.; Park, H. S.; Chung, D. M.; Baik, S. J.

1999-01-01

The computer program SAFE has been used to size and analyze the performance of a steam generator which has two types of heat transfer regions in Korean Standard Nuclear Power Plants (KSNP) and Korean Next Generation Reactor (KNGR) design. The SAFE code calculates the analytical boiling heat transfer area using the modified form of the saturated nucleate pool boiling correlation suggested by Rohsenow. The predicted heat transfer area in the boiling region is multiplied by a constant to obtain a final analytical heat transfer area. The inclusion of the multiplier in the analytical calculation has some disadvantage of loss of complete correlation by the governing heat transfer equation. Several comparative analyses have been performed quantitatively to evaluate the possibility of removing the multiplier in the analytical calculation in the SAFE code. The evaluation shows that the boiling correlation and multiplier used in predicting the boiling region heat transfer area can be replaced with other correlations predicting nearly the same heat transfer area. The removal of multiplier included in the analytical calculation will facilitate a direct use of a set of concerned analytical sizing values that can be exactly correlated by the governing heat transfer equation. In addition this will provide more reasonable basis for the steam generator thermal sizing calculation and enhance the code usability without loss of any validity of the current sizing procedure. (author)

Flow-induced vibration analysis of heat exchanger and steam generator designs

International Nuclear Information System (INIS)

Pettigrew, M.J.; Sylvestre, Y.; Campagna, A.O.

1977-08-01

Tube and shell heat exchange components such as steam generators, heat exchangers and condensers are essential parts of CANDU nuclear power stations. Excessive flow-induced vibration may cause tube failures by fatigue or more likely by fretting-wear. Such failures may lead to station shutdowns that are very undesirable in terms of lost production. Hence good performance and reliability dictate a thorough flow-induced vibration analysis at the design stage. This paper presents our approach and techniques in this respect. (author)

Design specifications to ensure flow-induced vibration and fretting-wear performance in CANDU steam generators and heat exchangers

International Nuclear Information System (INIS)

Janzen, V.P.; Han, Y.; Pettigrew, M.J.

2009-01-01

Preventing flow-induced vibration and fretting-wear problems in steam generators and heat exchangers requires design specifications that bring together specific guidelines, analysis methods, requirements and appropriate performance criteria. This paper outlines the steps required to generate and support such design specifications for CANDU nuclear steam generators and heat exchangers, and relates them to typical steam-generator design features and computer modeling capabilities. It also describes current issues that are driving changes to flow-induced vibration and fretting-wear specifications that can be applied to the design process for component refurbishment, replacement or new designs. These issues include recent experimental or field evidence for new excitation mechanisms, e.g., the possibility of in-plane fluidelastic instability of U-tubes, the demand for longer reactor and component lifetimes, the need for better predictions of dynamic properties and vibration response, e.g., two-phase random-turbulence excitation, and requirements to consider system 'excursions' or abnormal scenarios, e.g., a main steam line break in the case of steam generators. The paper describes steps being taken to resolve these issues. (author)

Dynamic simulation of steam generator failures

Energy Technology Data Exchange (ETDEWEB)

Meister, G [Institut fuer Nukleare Sicherheitsforschung, Kernforschungsanlage Juelich GmbH, Juelich (Germany)

1988-07-01

A computer program will be described which is capable to simulate severe transients in a gas heated steam generator. Such transients may arise in the safety analysis of accidents resulting from failures in the heat removal system of an HTGR power plant. Important failure modes which have to be considered are ruptures of one or more steam generator tubes leading to water or steam ejection into the primary system or anomalous operating conditions which my cause damage due to excessive thermal stress. Examples are the complete dryout as a consequence of feedwater interrupt in connection with continuing gas heating and the reflooding of the secondary channel with cold feedwater after dryout. The steam generator program which is capable to simulate accidents of this type is written as a module which can be implemented into a program system fur the simulation of the total heat rejection system. It based on an advanced mathematical model for the two phase flow taking deviations from thermal equilibrium into account. Mass, energy and momentum balances for the primary and secondary fluid and the heat diffusion equations for the heat exchanging wall form a system of coupled differential equations which is solved numerically by an algorithm which is stiffly stable and suppresses effectively oscillations of numerical origin. Results of the simulation of transients of the type mentioned above will be presented and discussed. (author)

Dynamic simulation of steam generator failures

International Nuclear Information System (INIS)

Meister, G.

1988-01-01

A computer program will be described which is capable to simulate severe transients in a gas heated steam generator. Such transients may arise in the safety analysis of accidents resulting from failures in the heat removal system of an HTGR power plant. Important failure modes which have to be considered are ruptures of one or more steam generator tubes leading to water or steam ejection into the primary system or anomalous operating conditions which my cause damage due to excessive thermal stress. Examples are the complete dryout as a consequence of feedwater interrupt in connection with continuing gas heating and the reflooding of the secondary channel with cold feedwater after dryout. The steam generator program which is capable to simulate accidents of this type is written as a module which can be implemented into a program system fur the simulation of the total heat rejection system. It based on an advanced mathematical model for the two phase flow taking deviations from thermal equilibrium into account. Mass, energy and momentum balances for the primary and secondary fluid and the heat diffusion equations for the heat exchanging wall form a system of coupled differential equations which is solved numerically by an algorithm which is stiffly stable and suppresses effectively oscillations of numerical origin. Results of the simulation of transients of the type mentioned above will be presented and discussed. (author)

Thermal-hydraulics in recirculating steam generators

International Nuclear Information System (INIS)

Carver, M.B.; Carlucci, L.N.; Inch, W.W.R.

1981-04-01

This manual describes the THIRST code and its use in computing three-dimensional two-phase flow and heat transfer in a steam generator under steady state operation. The manual is intended primarily to facilitate the application of the code to the analysis of steam generators typical of CANDU nuclear stations. Application to other steam generator designs is also discussed. Details of the assumptions used to formulate the model and to implement the numerical solution are also included

Comparison of Heat Transfer Fluid and Direct Steam Generation technologies for Integrated Solar Combined Cycles

International Nuclear Information System (INIS)

Rovira, Antonio; Montes, María José; Varela, Fernando; Gil, Mónica

2013-01-01

At present time and in the medium term, Solar Thermal Power Plants are going to share scenario with conventional energy generation technologies, like fossil and nuclear. In such a context, Integrated Solar Combined Cycles (ISCCs) may be an interesting choice since integrated designs may lead to a very efficient use of the solar and fossil resources. In this work, different ISCC configurations including a solar field based on parabolic trough collectors and working with the so-called Heat Transfer Fluid (HTF) and Direct Steam Generation (DSG) technologies are compared. For each technology, four layouts have been studied: one in which solar heat is used to evaporate part of the high pressure steam of a bottoming Rankine cycle with two pressure levels, another that incorporates a preheating section to the previous layout, the third one that includes superheating instead of preheating and the last one including both preheating and superheating in addition to the evaporation. The analysis is made with the aim of finding out which of the different layouts reaches the best performance. For that purpose, three types of comparisons have been performed. The first one assesses the benefits of including a solar steam production fixed at 50 MW th . The second one compares the configurations with a standardised solar field size instead of a fixed solar steam production. Finally, the last one consists on an even more homogeneous comparison considering the same steam generator size for all the configurations as well as standardised solar fields. The configurations are studied by mean of exergy analyses. Several figures of merit are used to correctly assess the configurations. Results reveal that the only-evaporative DSG configuration becomes the best choice, since it benefits of both low irreversibility at the heat recovery steam generator and high thermal efficiency in the solar field. Highlights: ► ISCC configurations with DSG and HTF technologies are compared. ► Four

Cleanliness criteria to improve steam generator performance

International Nuclear Information System (INIS)

Schwarz, T.; Bouecke, R.; Odar, S.

2005-01-01

High steam generator performance is a prerequisite for high plant availability and possible life time extension. The major opponent to that is corrosion and fouling of the heating tubes. Such steam generator degradation problems arise from the continuous ingress of non-volatile contaminants, i.e. corrosion products and salt impurities may accumulate in the steam generators. These impurities have their origin in the secondary side systems. The corrosion products generally accumulate in the steam generators and form deposits not only in the flow restricted areas, such as on top of tube sheet and tube support structure, but also build scales on the steam generator heating tubes. In addition, the tube scales in general affect the steam generator thermal performance, which ultimately causes a reduction of power output. The most effective ways of counteracting all these degradation problems, and thus of improving the steam generator performance is to keep them in clean conditions or, if judged necessary, to plan cleaning measures such as mechanical tube sheet lancing or chemical cleaning. This paper presents a methodology how to assess the cleanliness condition of a steam generator by bringing together all available operational and inspection data such as thermal performance and water chemistry data. By means of this all-inclusive approach the cleanliness condition is quantified in terms of a fouling index. The fouling index allows to monitor the condition of a specific steam generator, compare it to other plants and, finally, to serve as criterion for cleaning measures such as chemical cleaning. The application of the cleanliness criteria and the achieved field results with respect to improvements of steam generator performance will be presented. (author)

Method of processing results of tests of heating surfaces of a steam generator on a digital computer

Energy Technology Data Exchange (ETDEWEB)

Glusker, B.N.

1975-03-01

At present, processing of information obtained by testing steam generators in high-capacity generating units is carried out manually. This takes a long time and does not always permit one to process all the information obtained, which impoverishes the results of experimental work. In addition, this kind of processing of experimental results is as a rule done after completion of a considerable part of the tests, and occasionally after completion of all the tests. In this case, it is impossible to conduct a better directed, corrected experiment, and this leads to duplication of experiments and to increasing the period of adjusting and exploratory work on industrial plants. An algorithm was developed for automated processing of the hydraulic and temperature conditions of the heating surfaces in steam generators on digital computers, which is a part of the general algorithm of processing of results of thermal tests of steam generators. It includes calculation of all characteristics determining the thermal and hydraulic conditions of the heating surfaces. The program of processing includes a subprogram: determination of the thermophysical and thermodynamic properties of the water and steam.

Steady-state heat transfer in an inverted U-tube steam generator

International Nuclear Information System (INIS)

Boucher, T.J.

1986-01-01

Experimental results are presented involving U-tube steam generator tube bundle local heat transfer and fluid conditions during steady-state, full-power operations performed at high temperatures and pressures with conditions typical of a pressurized water reactor (15.0 MPa primary pressure, 600 K hot-leg fluid temperatures, 6.2 MPa secondary pressure). The MOD-2C facility represents the state-of-the-art in measurement of tube local heat transfer data and average tube bundle secondary fluid density at several elevations, which allows an estimate of the axial heat transfer and void distributions during steady-state and transient operations. The method of heat transfer data reduction is presented and the heat flux, secondary convective heat transfer coefficient, and void fraction distributions are quantified for steady-state, full-power operations

Corrosion Evaluation and Corrosion Control of Steam Generators

International Nuclear Information System (INIS)

Maeng, W. Y.; Kim, U. C.; Sung, K. W.; Na, J. W.; Lee, Y. H.; Lee, D. H.; Kim, K. M.

2008-06-01

Corrosion damage significantly influences the integrity and efficiency of steam generator. Corrosion problems of steam generator are unsolved issues until now even though much effort is made around world. Especially the stress corrosion cracking of heat exchange materials is the first issue to be solved. The corrosion protection method of steam generator is important and urgent for the guarantee of nuclear plant's integrity. The objectives of this study are 1) to evaluate the corrosion properties of steam generator materials, 2) to optimize the water chemistry of steam generator and 3) to develop the corrosion protection method of primary and secondary sides of steam generator. The results will be reflected to the water chemistry guideline for improving the integrity and efficiency of steam generator in domestic power plants

Results of Steam-Water-Oxygen Treatment of the Inside of Heating Surfaces in Heat-Recovery Steam Generators of the PGU-800 Power Unit at the Perm' District Thermal Power Station

Science.gov (United States)

Ovechkina, O. V.; Zhuravlev, L. S.; Drozdov, A. A.; Solomeina, S. V.

2018-05-01

Prestarting, postinstallation steam-water-oxygen treatment (SWOT) of the natural circulation/steam reheat heat-recovery steam generators (HRSG) manufactured by OAO Krasny Kotelshchik was performed at the PGU-800 power unit of the Perm District Thermal Power Station (GRES). Prior to SWOT, steam-oxygen cleaning, passivation, and preservation of gas condensate heaters (GCH) of HRSGs were performed for 10 h using 1.3MPa/260°C/70 t/h external steam. After that, test specimens were cut out that demonstrated high strength of the passivating film. SWOT of the inside of the heating surfaces was carried out during no-load operation of the gas turbine unit with an exhaust temperature of 280-300°C at the HRSG inlet. The steam turbine was shutdown, and the generated steam was discharged into the atmosphere. Oxygen was metered into the discharge pipeline of the electricity-driven feed pumps and downcomers of the evaporators. The behavior of the concentration by weight of iron compounds and the results of investigation of cutout specimens by the drop or potentiometric method indicate that the steam-water-oxygen process makes it possible to remove corrosion products and reduce the time required to put a boiler into operation. Unlike other processes, SWOT does not require metal-intensive cleaning systems, temporary metering stations, and structures for collection of the waste solution.

Layout of an internally heated gas generator for the steam gasification of coal

International Nuclear Information System (INIS)

Feistel, P.P.; Duerrfeld, R.; Heck, K.H. van; Juentgen, H.

1975-01-01

Industrial-scale steam gasification of coal using heat from high temperature reactors requires research and development on allothermal gas generators. Bergbau-Forschung GmbH, Essen, does theoretical and experimental work in this field. The experiments deal with reaction kinetics, heat transfer and material tests. Their significance for the layout of a full-scale gas generator is shown. Including material specifications, the feasibility of a gasifier, characterized by a fluid bed volume of 318 m 3 and a heat transferring area of 4000 m 2 , results. The data, now available, are used to determine the gasification throughput from the heat balance, i.e. the equality of heat consumed and heat transferred. Throughputs of about 50 t/hr of coal are possible for a single gas generator, the helium outlet temperature of the HTR being 950 0 C/ Bergbau-Forschung has commissioned a medium-scale pilot plant (200 kg/hr). (Auth.)

How can the heat transfer correlations for finned-tubes influence the numerical simulation of the dynamic behavior of a heat recovery steam generator?

International Nuclear Information System (INIS)

Walter, H.; Hofmann, R.

2011-01-01

This paper presents the results of a theoretical investigation on the influence of different heat transfer correlations for finned-tubes to the dynamic behavior of a heat recovery steam generator (HRSG). The investigation was done for a vertical type natural circulation HRSG with 3 pressure stages under hot start-up and shutdown conditions. For the calculation of the flue gas-side heat transfer coefficient the well known correlations for segmented finned-tubes according to Schmidt, VDI and ESCOA TM (traditional and revised) as well as a new correlation, which was developed at the Institute for Energy Systems and Thermodynamics, are used. The simulation results show a good agreement in the overall behavior of the boiler between the different correlations. But there are still some important differences found in the detail analysis of the boiler behavior. - Research highlights: → Numerical simulation is performed to explore the influence of different heat transfer correlations for finned-tubes to the dynamic behavior of a heat recovery steam generator. → Differences in the steam generator behavior are found. → In the worst case the boiler can lead to unfavorable operation conditions, e.g. reverse flow.

Steam generating system in LMFBR type reactors

International Nuclear Information System (INIS)

Kurosawa, Katsutoshi.

1984-01-01

Purpose: To suppress the thermal shock loads to the structures of reactor system and secondary coolant system, for instance, upon plant trip accompanying turbine trip in the steam generation system of LMFBR type reactors. Constitution: Additional feedwater heater is disposed to the pipeway at the inlet of a steam generator in a steam generation system equipped with a closed loop extended from a steam generator by way of a gas-liquid separator, a turbine and a condensator to the steam generator. The separated water at high temperature and high pressure from a gas-liquid separator is heat exchanged with coolants flowing through the closed loop of the steam generation system in non-contact manner and, thereafter, introduced to a water reservoir tank. This can avoid the water to be fed at low temperature as it is to the steam generator, whereby the thermal shock loads to the structures of the reactor system and the secondary coolant system can be suppressed. (Moriyama, K.)

Corrosion Evaluation and Corrosion Control of Steam Generators

Energy Technology Data Exchange (ETDEWEB)

Maeng, W. Y.; Kim, U. C.; Sung, K. W.; Na, J. W.; Lee, Y. H.; Lee, D. H.; Kim, K. M

2008-06-15

Corrosion damage significantly influences the integrity and efficiency of steam generator. Corrosion problems of steam generator are unsolved issues until now even though much effort is made around world. Especially the stress corrosion cracking of heat exchange materials is the first issue to be solved. The corrosion protection method of steam generator is important and urgent for the guarantee of nuclear plant's integrity. The objectives of this study are 1) to evaluate the corrosion properties of steam generator materials, 2) to optimize the water chemistry of steam generator and 3) to develop the corrosion protection method of primary and secondary sides of steam generator. The results will be reflected to the water chemistry guideline for improving the integrity and efficiency of steam generator in domestic power plants.

Improvements in steam cycle electric power generating plants

International Nuclear Information System (INIS)

Bienvenu, Claude.

1973-01-01

The invention relates to a steam cycle electric energy generating plants of the type comprising a fossil or nuclear fuel boiler for generating steam and a turbo alternator group, the turbine of which is fed by the boiler steam. The improvement is characterized in that use is made of a second energy generating group in which a fluid (e.g. ammoniac) undergoes a condensation cycle the heat source of said cycle being obtained through a direct or indirect heat exchange with a portion of the boiler generated steam whereby it is possible without overloading the turbo-alternator group, to accomodate any increase of the boiler power resulting from the use of another fuel while maintaining a maximum energy output. This can be applied to electric power stations [fr

Regulation of ageing steam generators

International Nuclear Information System (INIS)

Jarman, B.L.; Grant, I.M.; Garg, R.

1998-01-01

Recent years have seen leaks and shutdowns of Canadian CANDU plants due to steam generator tube degradation by mechanisms including stress corrosion cracking, fretting and pitting. Failure of a single steam generator tube, or even a few tubes, would not be a serious safety related event in a CANDU reactor. The leakage from a ruptured tube is within the makeup capacity of the primary heat transport system, so that as long as the operator takes the correct actions, the off-site consequences will be negligible. However, assurance that no tubes deteriorate to the point where their integrity could be seriously breached as result of potential accidents, and that any leakage caused by such an accident will be small enough to be inconsequential, can only be obtained through detailed monitoring and management of steam generator condition. This paper presents the AECB's current approach and future regulatory directions regarding ageing steam generators. (author)

IAEA activities on steam generator life management

International Nuclear Information System (INIS)

Gueorguiev, B.; Lyssakov, V.; Trampus, P.

2002-01-01

The International Atomic Energy Agency (IAEA) carries out a set of activities in the field of Nuclear Power Plant (NPP) life management. Main activities within this area are implemented through the Technical Working Group on Life Management of NPPs, and mostly concentrated on studies of understanding mechanisms of degradation and their monitoring, optimisation of maintenance management, economic aspects, proven practices of and approaches to plant life management including decommissioning. The paper covers two ongoing activities related to steam generator life management: the International Database on NPP Steam Generators and the Co-ordinated Research Project on Verification of WWER Steam Generator Tube Integrity (WWER is the Russian designed PWR). The lifetime assessment of main components relies on an ability to assess their condition and predict future degradation trends, which to a large extent is dependent on the availability of relevant data. Effective management of ageing and degradation processes requires a large amount of data. Several years ago the IAEA started to work on the International Database on NPP Life Management. This is a multi-module database consisting of modules such as reactor pressure vessels materials, piping, steam generators, and concrete structures. At present the work on pressure vessel materials, on piping as well as on steam generator is completed. The paper will present the concept and structure of the steam generator module of the database. In countries operating WWER NPPs, there are big differences in the eddy current inspection strategy and practice as well as in the approach to steam generator heat exchanger tube structural integrity assessment. Responding to the need for a co-ordinated research to compare eddy current testing results with destructive testing using pulled out tubes from WWER steam generators, the IAEA launched this project. The main objectives of the project are to summarise the operating experiences of WWER

Dismantling of the 50 MW steam generator test facility

International Nuclear Information System (INIS)

Nakai, S.; Onojima, T.; Yamamoto, S.; Akai, M.; Isozaki, T.; Gunji, M.; Yatabe, T.

1997-01-01

We have been dismantling the 50MW Steam Generator Test Facility (50MWSGTF). The objectives of the dismantling are reuse of sodium components to a planned large scale thermal hydraulics sodium test facility and the material examination of component that have been operated for long time in sodium. The facility consisted of primary sodium loop with sodium heater by gas burner as heat source instead of reactor, secondary sodium loop with auxiliary cooling system (ACS) and water/steam system with steam temperature and pressure reducer instead of turbine. It simulated the 1 loop of the Monju cooling system. The rated power of the facility was 50MWt and it was about 1/5 of the Monju power plant. Several sodium removal methods are applied. As for the components to be dismantled such as piping, intermediate heat exchanger (IHX), air cooled heat exchangers (AC), sodium is removed by steam with nitrogen gas in the air or sodium is burned in the air. As for steam generators which material tests are planned, sodium is removed by steam injection with nitrogen gas to the steam generator. The steam generator vessel is filled with nitrogen and no air in the steam generator during sodium removal. As for sodium pumps, pump internal structure is pulled out from the casing and installed into the tank. After the installation, sodium is removed by the same method of steam generator. As for relatively small reuse components such as sodium valves, electromagnet flow meters (EMFs) etc., sodium is removed by alcohol process. (author)

Experimental investigations of heat exchange and hydrodynamics on models of a VG-400 steam generator tube bundle made up of small diameter helicoils

International Nuclear Information System (INIS)

Golovko, V.F; Ivaskov, N.A.; Obukhov, P.I.; Pospelov, V.N.; Sergeev, A.I.

1988-01-01

Features of HTGR steam generators having heat exchange surface made up of small diameter helicoils are discussed in the paper. A general approach to optimization of thermohydraulic characteristics BΓW-400 steam generator design backed by calculation and experiment are given. Main results of steam generator assembly's model aerodynamic test are presented. Data of thermohydraulic tests of a single tube model in a helium heated test rig are discussed. (author)

A heat transfer study for vertical straight-tube steam generators heated by liquid metal

International Nuclear Information System (INIS)

Valette, M.

1984-04-01

A single-tube mockup of a vertical straight-tube steam generator heated by sodium-potassium alloy NaK was submitted to thermal and hydraulic testing in conditions representative of fast breeder reactor operation. The mockup consisted of a 10mm I.D. ferritic steel heat exchange tube centered inside a cylindrical stainless steel shell. The complete assembly was 20.9 meters long. Water flowed upward inside the exchange tube, and NaK flowed downward in the annular gap between the tube and the shell. The steam outlet pressure ranged from 90 to 195 bars, while the liquid metal temperature at the mockup inlet was between 480 and 580 0 C. The water flowrate in the tube ranged from 153 to 2460 kg.m -2 .s -1 . During the tests the fluid inlet and outlet temperatures, flowrate and pressures were measured, as was the NaK temperature profile over the full length of the device. The test results were subsequently compared with heat exchange and pressure drop values calculated using the standard formulas for straight-tube heat exchangers. The heat exchange coefficients predicted by these correlations in the boiling zone were found to be largely overestimated, while the calculated pressure drop values proved satisfactory. A set of modified correlations is proposed to account for the observed phenomena, and for use in designing commercial units, provided the sodium flow in the tube bundle is adequately distributed

Intermediate heat exchanger and steam generator designs for the HYLIFE-II fusion power plant using molten salts

International Nuclear Information System (INIS)

Lee, Y.T.; Hoffman, M.A.

1992-01-01

The HYLIFE-II fusion power plant employs the molten salt, Flibe, for the liquid jets which form the self-healing 'first wall' of the reactor. The molten salt, sodium fluoroborate then transports the heat from the IHX's to the steam generators. The design and optimization of the IHX's and the steam generators for use with molten salts has been done as part of the HYLIFE-II conceptual design study. The results of this study are described, and reference designs of these large heat exchangers are selected to minimize the cost of electricity while satisfying other important constraints

Reconstruction of steam generators super emergency feadwater supply system (SHNC) and steam dump stations to the atmosphere system PSA

International Nuclear Information System (INIS)

Kuzma, J.

2001-01-01

Steam Generators Super Emergency Feadwater Supply System (SHNC) and Steam Dump Stations to the Atmosphere System (PSA) are two systems which cooperate to remove residual heat from reactor core after seismic event. SHNC assure feeding of the secondary site of steam generator (Feed) where after heat removal.from primary loops, is relieved to the atmosphere by PSA (Bleed) in form of steam. (author)

Pressurized-water coolant nuclear reactor steam generator

International Nuclear Information System (INIS)

Mayer, H.; Schroder, H.J.

1975-01-01

A description is given of a pressurized-water coolant nuclear reactor steam generator having a vertical housing for the steam generating water and containing an upstanding heat exchanger to which the pressurized-water coolant passes and which is radially surrounded by a guide jacket supporting a water separator on its top. By thermosiphon action the steam generating water flows upward through and around the heat exchanger within the guide chamber to the latter's top from which it flows radially outwardly and downwardly through a down draft space formed between the outside of the jacket and the housing. The water separator discharges separated water downwardly. The housing has a feedwater inlet opening adjacent to the lower portion of the heat exchanger, providing preheating of the introduced feedwater. This preheated feedwater is conveyed by a duct upwardly to a location where it mixes with the water discharged from the water separator

Horizontal steam generator PGV-1000 thermal-hydraulic analysis

Energy Technology Data Exchange (ETDEWEB)

Ubra, O. [Skoda Company, Prague (Switzerland); Doubek, M. [Czech Technical Univ., Prague (Switzerland)

1995-12-31

A computer program for the steady state thermal-hydraulic analysis of horizontal steam generator PGV-1000 is presented. The program provides the capability to analyze steam generator PGV-1000 primary side flow and temperature distribution, primary side pressure drops, heat transfer between the primary and secondary sides and multidimensional heat flux distribution. A special attention is paid to the thermal-hydraulics of the secondary side. The code predicts 3-D distribution of the void fraction at the secondary side, mass redistribution under the submerged perforated sheet and the steam generator level profile. By means of developed computer program a detailed thermal-hydraulic study of the PGV-1000 has been carried out. A wide range of calculations has been performed and a set of important steam generator characteristics has been obtained. Some of them are presented in the paper. (orig.). 5 refs.

Horizontal steam generator PGV-1000 thermal-hydraulic analysis

International Nuclear Information System (INIS)

Ubra, O.; Doubek, M.

1995-01-01

A computer program for the steady state thermal-hydraulic analysis of horizontal steam generator PGV-1000 is presented. The program provides the capability to analyze steam generator PGV-1000 primary side flow and temperature distribution, primary side pressure drops, heat transfer between the primary and secondary sides and multidimensional heat flux distribution. A special attention is paid to the thermal-hydraulics of the secondary side. The code predicts 3-D distribution of the void fraction at the secondary side, mass redistribution under the submerged perforated sheet and the steam generator level profile. By means of developed computer program a detailed thermal-hydraulic study of the PGV-1000 has been carried out. A wide range of calculations has been performed and a set of important steam generator characteristics has been obtained. Some of them are presented in the paper. (orig.)

Horizontal steam generator PGV-1000 thermal-hydraulic analysis

Energy Technology Data Exchange (ETDEWEB)

Ubra, O [Skoda Company, Prague (Switzerland); Doubek, M [Czech Technical Univ., Prague (Switzerland)

1996-12-31

A computer program for the steady state thermal-hydraulic analysis of horizontal steam generator PGV-1000 is presented. The program provides the capability to analyze steam generator PGV-1000 primary side flow and temperature distribution, primary side pressure drops, heat transfer between the primary and secondary sides and multidimensional heat flux distribution. A special attention is paid to the thermal-hydraulics of the secondary side. The code predicts 3-D distribution of the void fraction at the secondary side, mass redistribution under the submerged perforated sheet and the steam generator level profile. By means of developed computer program a detailed thermal-hydraulic study of the PGV-1000 has been carried out. A wide range of calculations has been performed and a set of important steam generator characteristics has been obtained. Some of them are presented in the paper. (orig.). 5 refs.

Heat transfer in the over-crisis region in a steam-generating channel with steam-droplet flow

International Nuclear Information System (INIS)

Nigmatulin, B.I.; Kukharenko, V.N.

1991-01-01

Statement and results of numerical solution of the problem on heat transfer in the over-crisis region under stationary and nonstationary conditions are studied. Two-temperature model, accounting for influence of phase slipping on heat exchange between steam and droplets and direct heat interaction with heated surface is used for describing steam-droplet flow. Comparison is made of calculational results and experimental data published in the literature on heated surface temperature within broad range of parameters: pressure - p=0.3-18.5 MPa; specific mass flow rate of the mixture - G=30-5000 kg/m 2 xs; specific heat flows - q=0.02-3 MW/m 2 and tube diameter - D=4-20 mm

Steam generator for PWR type reactor

International Nuclear Information System (INIS)

Baba, Iwao; Hiyama, Nobuyuki.

1994-01-01

A steam generator of the present invention comprises a primary coolant chamber having primary coolants circulating therein, a secondary coolants chamber having secondary coolants and steams circulating therein, which are isolated from each other by a partition wall, and heat pipes disposed being passed through the partition wall. The heat pipes are disposed having an evaporation portion in the primary coolants chamber, a condensation portion in the secondary coolants chamber, and an intermediate heat insulating portion in the partition wall. Since the primary coolants containing radioactivity and the secondary coolants not containing radioactivity does not transfer heat directly by a heat transfer wall, a leakage accident of radioactivity to the secondary coolants can be prevented. Moreover, since the heat pipes are used, a great amount of heat can be transferred by a slight temperature difference by using steams of the heat transfer medium itself, latent heat due to coagulation, and capillary phenomenon. Since neither transferring power nor pumps are required, heat of the primary coolants can effectively be transferred to the secondary coolants. (N.H.)

Wastage of Steam Generator Tubes by Sodium-Water Reaction

International Nuclear Information System (INIS)

Jeong, Ji Young; Kim, Jong Man; Kim, Tae Joon; Choi, Jong Hyeun; Kim, Byung Ho; Lee, Yong Bum; Park, Nam Cook

2010-01-01

The Korea Advanced LIquid MEtal Reactor (KALIMER) steam generator is a helical coil, vertically oriented, shell-and-tube type heat exchanger with fixed tube-sheet. The conceptual design and outline drawing of the steam generator are shown in Figure 1. Flow is counter-current, with sodium on the shell side and water/steam on the tube side. Sodium flow enters the steam generator through the upper inlet nozzles and then flows down through the tube bundle. Feedwater enters the steam generator through the feedwater nozzles at the bottom of steam generator. Therefore, if there is a hole or a crack in a heat transfer tube, a leakage of water/steam into the sodium may occur, resulting in a sodium-water reaction. When such a leak occurs, so-called 'wastage' is the result which may cause damage to or a failure of the adjacent tubes. If a steam generator is operated for some time in this condition, it is possible that it might create an intermediate leak state which would then give rise to the problems of a multi-target wastage in a very short time. Therefore, it is very important to predict these phenomena quantitatively from the view of designing a steam generator and its leak detection systems. For this, multi-target wastage tests for modified 9Cr-1Mo steel tube bundle by intermediate leaks are being prepared

Improvements of reforming performance of a nuclear heated steam reforming process

International Nuclear Information System (INIS)

Hada, Kazuhiko

1996-10-01

Performance of an energy production process by utilizing high temperature nuclear process heat was not competitive to that by utilizing non-nuclear process heat, especially fossil-fired process heat due to its less favorable chemical reaction conditions. Less favorable conditions are because a temperature of the nuclear generated heat is around 950degC and the heat transferring fluid is the helium gas pressurized at around 4 MPa. Improvements of reforming performance of nuclear heated steam reforming process were proposed in the present report. The steam reforming process, one of hydrogen production processes, has the possibility to be industrialized as a nuclear heated process as early as expected, and technical solutions to resolve issues for coupling an HTGR with the steam reforming system are applicable to other nuclear-heated hydrogen production systems. The improvements are as follows: As for the steam reformer, (1) increase in heat input to process gas by applying a bayonet type of reformer tubes and so on, (2) increase in reforming temperature by enhancing heat transfer rate by the use of combined promoters of orifice baffles, cylindrical thermal radiation pipes and other proposal, and (3) increase in conversion rate of methane to hydrogen by optimizing chemical compositions of feed process gas. Regarding system arrangement, a steam generator and superheater are set in the helium loop as downstream coolers of the steam reformer, so as to effectively utilize the residual nuclear heat for generating feed steam. The improvements are estimated to achieve the hydrogen production rate of approximately 3800 STP-m 3 /h for the heat source of 10 MW and therefore will provide the potential competitiveness to a fossil-fired steam reforming process. Those improvements also provide the compactness of reformer tubes, giving the applicability of seamless tubes. (J.P.N.)

Heat transfer efficient thermal energy storage for steam generation

International Nuclear Information System (INIS)

Adinberg, R.; Zvegilsky, D.; Epstein, M.

2010-01-01

A novel reflux heat transfer storage (RHTS) concept for producing high-temperature superheated steam in the temperature range 350-400 deg. C was developed and tested. The thermal storage medium is a metallic substance, Zinc-Tin alloy, which serves as the phase change material (PCM). A high-temperature heat transfer fluid (HTF) is added to the storage medium in order to enhance heat exchange within the storage system, which comprises PCM units and the associated heat exchangers serving for charging and discharging the storage. The applied heat transfer mechanism is based on the HTF reflux created by a combined evaporation-condensation process. It was shown that a PCM with a fraction of 70 wt.% Zn in the alloy (Zn70Sn30) is optimal to attain a storage temperature of 370 deg. C, provided the heat source such as solar-produced steam or solar-heated synthetic oil has a temperature of about 400 deg. C (typical for the parabolic troughs technology). This PCM melts gradually between temperatures 200 and 370 deg. C preserving the latent heat of fusion, mainly of the Zn-component, that later, at the stage of heat discharge, will be available for producing steam. The thermal storage concept was experimentally studied using a lab scale apparatus that enabled investigating of storage materials (the PCM-HTF system) simultaneously with carrying out thermal performance measurements and observing heat transfer effects occurring in the system. The tests produced satisfactory results in terms of thermal stability and compatibility of the utilized storage materials, alloy Zn70Sn30 and the eutectic mixture of biphenyl and diphenyl oxide, up to a working temperature of 400 deg. C. Optional schemes for integrating the developed thermal storage into a solar thermal electric plant are discussed and evaluated considering a pilot scale solar plant with thermal power output of 12 MW. The storage should enable uninterrupted operation of solar thermal electric systems during additional hours

Steam-Generator Integrity Program/Steam-Generator Group Project

International Nuclear Information System (INIS)

1982-10-01

The Steam Generator Integrity Program (SGIP) is a comprehensive effort addressing issues of nondestructive test (NDT) reliability, inservice inspection (ISI) requirements, and tube plugging criteria for PWR steam generators. In addition, the program has interactive research tasks relating primary side decontamination, secondary side cleaning, and proposed repair techniques to nondestructive inspectability and primary system integrity. The program has acquired a service degraded PWR steam generator for research purposes. This past year a research facility, the Steam Generator Examination Facility (SGEF), specifically designed for nondestructive and destructive examination tasks of the SGIP was completed. The Surry generator previously transported to the Hanford Reservation was then inserted into the SGEF. Nondestructive characterization of the generator from both primary and secondary sides has been initiated. Decontamination of the channelhead cold leg side was conducted. Radioactive field maps were established in the steam generator, at the generator surface and in the SGEF

On the heat exchange tube failures in steam generators at NPPs with WWER reactors

International Nuclear Information System (INIS)

Titov, V.F.; Banyuk, G.F.; Brykov, S.I.

1992-01-01

Data on dynamics of failed heat exchanging tube closing in steam generators of NPPs with WWER type reactors for the whole period of their operation are presented. It is shown that the main cause of the tube failures consists in their corrosion cracking under stresses. The effect of chlorine ions on tubes is intensified by the presence of porous sediments on heat exchaning surfaces in quantities exceeding 150 g/m 2

Characterization of elevated temperature properties of heat exchanger and steam generator alloys

International Nuclear Information System (INIS)

Wright, J.K.; Carroll, L.J.; Cabet, C.; Lillo, T.M.; Benz, J.K.; Simpson, J.A.; Lloyd, W.R.; Chapman, J.A.; Wright, R.N.

2012-01-01

The Next Generation Nuclear Plant project is considering Alloy 800H and Alloy 617 for steam generator and intermediate heat exchangers. It is envisioned that a steam generator would operate with reactor outlet temperatures from 750 to 800 °C, while an intermediate heat exchanger for primary to secondary helium would operate up to an outlet temperature of 950 °C. Although both alloys are of interest due in part to their technical maturity, a number of specific properties require further characterization for design of nuclear components. Strain rate sensitivity of both alloys has been characterized and is found to be significant above 600 °C. Both alloys also exhibit dynamic strain aging, characterized by serrated flow, over a wide range of temperatures and strain rates. High temperature tensile testing of Alloy 617 and Alloy 800H has been conducted over a range of temperatures. Dynamic strain aging is a concern for these materials since it is observed to result in reduced ductility for many solid solution alloys. Creep, fatigue, and creep–fatigue properties of Alloy 617 have been measured as well, with the goal of determining the influence of the temperature, strain rate and atmosphere on the creep–fatigue life of Alloy 617. Elevated temperature properties and implications for codification of the alloys will be described.

Thermo hydrodynamical analyses of steam generator of nuclear power plant

International Nuclear Information System (INIS)

Petelin, S.; Gregoric, M.

1984-01-01

SMUP computer code for stationary model of a U-tube steam generator of a PWR nuclear power plant was developed. feed water flow can enter through main and auxiliary path. The computer code is based on the one dimensional mathematical model. Among the results that give an insight into physical processes along the tubes of steam generator are distribution of temperatures, water qualities, heat transfer rates. Parametric analysis permits conclusion on advantage of each design solution regarding heat transfer effects and safety of steam generator. (author)

PMK-2. Experimental study on steam generator behaviour

Energy Technology Data Exchange (ETDEWEB)

Ezsoel, G.; Szabados, L.; Trosztel, I. [KFKI Atomic Energy Research Inst., Budabest (Hungary)

1995-12-31

The PMK-2 is a full pressure scaled-down model of the Paks Nuclear Power Plant, with a 1:2070 scaling ratio for the volume and power. It has a steam generator model which is a vertical section of the horizontal steam generator. The model has hot and cold collectors similarly to the steam generators of the plant. The heat transfer tubes are horizontal tubes. There are 82 rows of tubes and the elevations, as well as the heat transfer surface distribution is the same as in the plant. The elevation of the feed water supply is similar to that of the plant. To study the temperature distribution in both the primary and the secondary side several thermocouples are built in, in addition to the overall instrumentation of the loop which has again a high number of measurement channels. Paper gives a description and results of SPE-4, with special respect to the steam generator behaviour in both steady state and transient conditions. Axial distribution of coolant and feedwater temperatures are given for the primary and the secondary side of hot and cold collectors and the temperature distribution in the centre of steam generator. (orig.).

PMK-2. Experimental study on steam generator behaviour

Energy Technology Data Exchange (ETDEWEB)

Ezsoel, G; Szabados, L; Trosztel, I [KFKI Atomic Energy Research Inst., Budabest (Hungary)

1996-12-31

The PMK-2 is a full pressure scaled-down model of the Paks Nuclear Power Plant, with a 1:2070 scaling ratio for the volume and power. It has a steam generator model which is a vertical section of the horizontal steam generator. The model has hot and cold collectors similarly to the steam generators of the plant. The heat transfer tubes are horizontal tubes. There are 82 rows of tubes and the elevations, as well as the heat transfer surface distribution is the same as in the plant. The elevation of the feed water supply is similar to that of the plant. To study the temperature distribution in both the primary and the secondary side several thermocouples are built in, in addition to the overall instrumentation of the loop which has again a high number of measurement channels. Paper gives a description and results of SPE-4, with special respect to the steam generator behaviour in both steady state and transient conditions. Axial distribution of coolant and feedwater temperatures are given for the primary and the secondary side of hot and cold collectors and the temperature distribution in the centre of steam generator. (orig.).

Steam generator for pressurized-water reactors

International Nuclear Information System (INIS)

Michel, E.

1971-01-01

In the steam generator for a PWR the central fall space of a U-tube bundel heat exchanger is used as a preliminary cyclon separator. The steam escaping upwards, which is largely free of water, can flow through the residual heating surface, i.e. the U-tube turns. In this way substantial drying and less superheating by the heat still added becomes possible. In its upper part the central fall space for the water separated in the preliminary separator, enclosed by a cylindrical guide wall and the U-tube bundle, is provided with tangential inlet slots. Through these, the water-steam mixture steams out of the section of the vertical legs of the U-tube bundle into the fall space. Above the inlet slots the rising space is closed by means of a turn-round plate. At the lower end of the guide wall outlet, slots are provided for the water flowing downwards and radially outwards into the unfilled space. (DG/PB) [de

Design and manufacture of steam generators for replacement

International Nuclear Information System (INIS)

Hirano, Hiroshi; Kuri, Syuhei

1995-01-01

The basic specification of the steam generators for replacement as heat exchangers (the pressure, temperature, flow rate and thermal output on primary and secondary sides) is set same as that of steam generators before replacement, but the latest design reflecting the operation experience obtained so far and taking the countermeasures for preventing heating tube damage in it is adopted, such as the heating tubes made of TT 690 alloy, the tube support plates with four-lobe shape tube holes made of stainless steel, the stainless steel rest fittings of three in one set and so on. After the heating tube break accident in Mihama No. 2 plant, the quality control was further strengthened. The comparison of the specifications of the steam generators of respective plants before and after the replacement is shown. The main objective of improving steam generators is the heightening of the reliability of heating tubes against intergranular attack and primary water stress corrosion cracking. The improvements of heating tube material, tube support plate material, secondary side heat flow, the shape of tube holes of tube support plates, the method of expanding heating tubes, and vibration-controlling fittings are explained. As to the manufacturing procedure and quality control, the manufacture of raw materials, the build-up welding of tube plates, the manufacture of lower half shell plates, the tube hole making of support plates, the insection of outer cylinder, flow rate distribution plate. Support plates and heating tubes, the sealing welding and expanding of heating tubes, the fixing of rest fittings, the manufacture and fixing of water chamber cover, the manufacture of upper half shell, the fixing of parts inside it, the final joint and inspection are described. (K.I.)

Dynamic and control of a once through steam generator

International Nuclear Information System (INIS)

Gomes, Arivaldo Vicente

1979-01-01

This paper presents a non linear distributed parameter model for the dynamics and feedback control of a large countercurrent heat exchanger used as a once through steam generator for a breeder reactor power plant. A convergent, implicit method has been developed to solve simultaneously the equations of conservation of mass, momentum and energy. The model, applicable to heat exchanger systems in general, has been used specifically to study the performance of a once-through steam generator with respect to its load following ability and stability of throttle steam temperature and pressure. (author)

Steam 80 steam generator instrumentation

International Nuclear Information System (INIS)

Carson, W.H.; Harris, H.H.

1980-01-01

This paper describes two special instrumentation packages in an integral economizer (preheater) steam generator of one of the first System 80 plants scheduled to go into commercial operation. The purpose of the instrumentation is to obtain accurate operating information from regions of the secondary side of the steam generator inaccessible to normal plant instrumentation. In addition to verification of the System 80 steam generator design predictions, the data obtained will assist in verification of steam generator thermal/hydraulic computer codes developed for generic use in the industry

Mushrooms as Efficient Solar Steam-Generation Devices.

Science.gov (United States)

Xu, Ning; Hu, Xiaozhen; Xu, Weichao; Li, Xiuqiang; Zhou, Lin; Zhu, Shining; Zhu, Jia

2017-07-01

Solar steam generation is emerging as a promising technology, for its potential in harvesting solar energy for various applications such as desalination and sterilization. Recent studies have reported a variety of artificial structures that are designed and fabricated to improve energy conversion efficiencies by enhancing solar absorption, heat localization, water supply, and vapor transportation. Mushrooms, as a kind of living organism, are surprisingly found to be efficient solar steam-generation devices for the first time. Natural and carbonized mushrooms can achieve ≈62% and ≈78% conversion efficiencies under 1 sun illumination, respectively. It is found that this capability of high solar steam generation is attributed to the unique natural structure of mushroom, umbrella-shaped black pileus, porous context, and fibrous stipe with a small cross section. These features not only provide efficient light absorption, water supply, and vapor escape, but also suppress three components of heat losses at the same time. These findings not only reveal the hidden talent of mushrooms as low-cost materials for solar steam generation, but also provide inspiration for the future development of high-performance solar thermal conversion devices. © 2017 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Kind of approximate theoretical calculating formula of heat-exchange area for the vertical U-bend tube natural-circuit steam generator

International Nuclear Information System (INIS)

Luo Mingkun; Wang Fei; Huang Wei; Zhang Wenqi; Zhao Shan; Lu Lianghong

2001-01-01

A kind of approximate theoretical calculating formula of the vertical U-bend tube natural-circuit steam generator is deduced by using an approximate method, the results of this formula is compared with the heat exchanging areas of the real vertical U-bend tube natural-circuit steam generators, the absolute errors of them are below 8%

Combined heat and power considered as a virtual steam cycle heat pump

International Nuclear Information System (INIS)

Lowe, Robert

2011-01-01

The first aim of this paper is to shed light on the thermodynamic reasons for the practical pursuit of low temperature operation by engineers involved in the design and the operation of combined heat and power (CHP) and district heating (DH) systems. The paper shows that the steam cycle of a combined heat and power generator is thermodynamically equivalent to a conventional steam cycle generator plus an additional virtual steam cycle heat pump. This apparently novel conceptualisation leads directly to (i) the observed sensitivity of coefficient of performance of CHP to supply and return temperatures in associated DH systems, and (ii) the conclusion that the performance of CHP will tend to be significantly higher than real heat pumps operating at similar temperatures. The second aim, which is pursued more qualitatively, is to show that the thermodynamic performance advantages of CHP are consistent with the goal of deep, long-term decarbonisation of industrialised economies. As an example, estimates are presented, which suggest that CHP based on combined-cycle gas turbines with carbon capture and storage has the potential to reduce the carbon intensity of delivered heat by a factor of ∼30, compared with a base case of natural gas-fired condensing boilers. - Highlights: → Large-scale CHP systems are thermodynamically equivalent to virtual steam cycle heat pumps. → COPs of such virtual heat pumps are necessarily better than the Carnot limit for real heat pumps. → COPs can approach 9 for plant matched to district heating systems with flow temperatures of 90 deg. C. → CHP combined with CCGT and CCS can reduce the carbon intensity of delivered heat ∼30-fold.

Energy balance and flow in steam generator part with sodium-water reaction

International Nuclear Information System (INIS)

Matal, O.

1980-01-01

Relations were derived for the calculation of heat liberated during the sodium water reaction in a tube failure in different parts of a steam generator. The results are graphically shown in i-T diagrams. Heat removal is described from the reaction zone to water and steam in undisturbed tubes and to the steam generator metal structure. (author)

Characterization of Elevated Temperature Properties of Heat Exchanger and Steam Generator Alloys

International Nuclear Information System (INIS)

Wright, J.K.; Carroll, L.J.; Benz, J.K.; Simpson, J.A.; Wright, R.N.; Lloyd, W.R.; Chapman, J.A.

2010-01-01

The Next Generation Nuclear Plant project is considering Alloy 800H and Alloy 617 for steam generator and intermediate heat exchangers. It is envisioned that a steam generator would operate with reactor outlet temperatures from 750 to 800 C, while an intermediate heat exchanger for primary to secondary helium would operate up to an outlet temperature of 950 C. Although both alloys are of interest due in part to their technical maturity, a number of specific properties require further characterization for design of nuclear components. Strain rate sensitivity of both alloys has been characterized and is found to be significant above 600 C. Both alloys also exhibit dynamic strain aging, characterized by serrated flow, over a wide range of temperatures and strain rates. In general dynamic strain aging is observed to begin at higher temperatures and serrated flow persists to higher temperatures in Alloy 617 compared to Alloy 800H. Dynamic strain aging is a concern for these materials since it is observed to result in reduced ductility for many solid solution alloys. The role of dynamic strain aging in the creep-fatigue behavior of Alloy 617 at temperatures of 800 C and above has also been examined in detail. Serrated flow is found to persist in cyclic stress-strain curves up to nearly the cycle to failure in some temperature and strain regimes. Results of those experiments and implications for creep-fatigue testing protocols will be described.

Strategy for assessment of WWER steam generator tube integrity. Report prepared within the framework of the coordinated research project on verification of WWER steam generator tube integrity

International Nuclear Information System (INIS)

2007-12-01

Steam generator heat exchanger tube degradations happen in WWER Nuclear Power Plant (NPP). The situation varies from country to country and from NPP to NPP. More severe degradation is observed in WWER-1000 NPPs than in case of WWER-440s. The reasons for these differences could be, among others, differences in heat exchanger tube material (chemical composition, microstructure, residual stresses), in thermal and mechanical loadings, as well as differences in water chemistry. However, WWER steam generators had not been designed for eddy current testing which is the usual testing method in steam generators of western PWRs. Moreover, their supplier provided neither adequate methodology and criteria nor equipment for planning and implementing In-Service Inspection (ISI). Consequently, WWER steam generator ISI infrastructure was established with delay. Even today, there are still big differences in the eddy current inspection strategy and practice as well as in the approach to steam generator heat exchanger tube structural integrity assessment (plugging criteria for defective tubes vary from 40 to 90% wall thickness degradation). Recognizing this situation, the WWER operating countries expressed their need for a joint effort to develop methodology to establish reasonable commonly accepted integrity assessment criteria for the heat exchanger tubes. The IAEA's programme related to steam generator life management is embedded into the systematic activity of its Technical Working Group on Life Management of Nuclear Power Plants (TWG-LMNPP). Under the advice of the TWG-LMNPP, an IAEA coordinated research project (CRP) on Verification of WWER Steam Generator Tube Integrity was launched in 2001. It was completed in 2005. Thirteen organizations involved in in-service inspection of steam generators in WWER operating countries participated: Croatia, Czech Republic, Finland, France, Hungary, Russian Federation, Slovakia, Spain, Ukraine, and the USA. The overall objective was to

A performance analysis of integrated solid oxide fuel cell and heat recovery steam generator for IGFC system

DEFF Research Database (Denmark)

Rudra, Souman; Lee, Jinwook; Rosendahl, Lasse

2010-01-01

efficiencies can be achieved. The outputs from SOFC can be utilized by heat recovery steam generator (HRSG), which drives the steam turbine for electricity production. The SOFC stack model was developed using the process flow sheet simulator Aspen Plus, which is of the equilibrium type. Various ranges...... of syngas properties gathered from different literature were used for the simulation. The results indicate a trade-off efficiency and power with respect to a variety of SOFC inputs. The HRSG located after SOFC was included in the current simulation study with various operating parameters. This paper...... describes IGFC power plants, particularly the optimization of HRSG to improve the efficiency of the heat recovery from the SOFC exhaust gas and to maximize the power production in the steam cycle in the IGFC system. HRSG output from different pressure levels varies depending on the SOFC output. The steam...

Steam generator development in France for the Super Phenix project

International Nuclear Information System (INIS)

Robin, M.G.

1975-01-01

'Steam Generator Development for Super Phenix Project'. The development program of steam generators studied by Fives-Cail Babcock and Stein Industrie Companies, jointly with CEA end EDF, for the Super Phenix 1200 MWe Fast Breeder Power Plant, is presented. The main characteristics of both sodium heated steam generators are emphasized and experimental studies related to their key features are reported. (author)

Thermohydraulic verification during THTR steam generator commissioning

International Nuclear Information System (INIS)

Henry, C.; Elter, C.

1988-01-01

In one of the six THTR 300 steam generators thermocouples are installed inside the heat transfer tube bundles for measuring the gas and steam temperatures. Fluid temperature distribution measurements along and across the helix bundle have been recorded in its first months of operation over a load range of 40% up to 100% for steady state and transient conditions. Using these measurements as well as the rest of the operating instrumentation. the computer programs for the design of heat exchanger heat transfer areas are verified. The temperature measurements for steady state conditions are compared with predictions obtained in the design stage. In these codes. the heat transferred from the outside helium gas to the water/steam inside the tubes is determined in discrete steps along the heating surface by one- and two-phase heat transfer correlations. The degree of conformity between prediction and measurement is discussed and compared with more recent correlations. (author)

Steam and sodium leak simulation in a fluidized-bed steam generator

International Nuclear Information System (INIS)

Vaux, W.G.; Keeton, A.R.; Keairns, D.L.

1977-01-01

A fluidized-bed steam generator for the liquid metal fast breeder reactor enhances plant availability and minimizes the probability of a water/sodium reaction. An experimental test program was conceived to assess design criteria and fluidized-bed operation under conditions of water, steam, and sodium leaks. Sodium, steam, and water were leaked into helium-fluidized beds of metal and ceramic particles at 900 F. Test results show the effects of leaks on the heat transfer coefficient, quality of fluidization, leak detection, and cleanup procedures

Rapid Generation of Superheated Steam Using a Water-containing Porous Material

Science.gov (United States)

Mori, Shoji; Okuyama, Kunito

Heat treatment by superheated steam has been utilized in several industrial fields including sterilization, desiccation, and cooking. In particular, cooking by superheated steam is receiving increased attention because it has advantages of reducing the salt and fat contents in foods as well as suppressing the oxidation of vitamin C and fat. In this application, quick startup and cut-off responses are required. Most electrically energized steam generators require a relatively long time to generate superheated steam due to the large heat capacities of the water in container and of the heater. Zhao and Liao (2002) introduced a novel process for rapid vaporization of subcooled liquid, in which a low-thermal-conductivity porous wick containing water is heated by a downward-facing grooved heating block in contact with the upper surface of the wick structure. They showed that saturated steam is generated within approximately 30 seconds from room-temperature water at a heat flux 41.2 kW⁄m2. In order to quickly generate superheated steam of approximately 300°C, which is required for cooking, the heat capacity of the heater should be as small as possible and the imposed heat flux should be so high enough that the porous wick is able to dry out in the vicinity of the contact with the heater and that the resulting heater temperature becomes much higher than the saturation temperature. The present paper proposes a simple structured generator to quickly produce superheated steam. Only a fine wire heater is contacted spirally on the inside wall in a hollow porous material. The start-up, cut-off responses and the rate of energy conversion for input power are investigated experimentally. Superheated steam of 300°C is produced in approximately 19 seconds from room-temperature water for an input power of 300 W. The maximum rate of energy conversion in the steady state is approximately 0.9.

Evaluation of heat transfer tube failure propagation due to sodium-water reaction in steam generator

International Nuclear Information System (INIS)

Nei, Hiromichi

1978-01-01

An evaluation was made of heat transfer tube failure propagation due to sodium-water reaction wastage in a sodium heated steam generator, by comparing an empirically derived wastage equation with leak detector responses. The experimental data agreed well with the wastage equation even for different values of distance-to-nozzle diameter ratio, though the formula had been based on wastage data obtained for only one given distance. The time taken for failure propagation was estimated for a prototype steam generator, and compared with the responses characteristics of acoustic detectors and level gages. It was found that there exists a range of leak rate between 0.5 and 100 g/sec, where the level gage can play a useful role in leak detection. The acoustic detector can be expected to respond more rapidly than the cover gas pressure gage, if noise is kept below ten times the value observed in an experimental facility, SWAT-2. (auth.)

Thermo-hydraulic stability study of a steam generator

International Nuclear Information System (INIS)

Magni, M C; Marcel, C P; Delmastro, D F

2012-01-01

In this work a mathematical model developed to investigate the thermalhydraulic stability of a helically coiled steam generator is presented. Such a steam generator is prone to experiment density wave oscillations. The model is therefore used to analyze the stability of the CAREM-25 reactor steam generators. The model is linear, numerically non-diffusive and nodal. In addition, it is able to represent non-uniform heat transfer fluxes between the primary and secondary coolant circuits. By using this model the marginal stability condition is found by varying the inlet friction coefficient for different conditions. The results are then compared with those obtained with a different model for which a simple uniform heat flux profiled is assumed. It is found that with such simplification the density waves instability mechanism is overestimated in a wide range of operating powers. For very low powers, in the contrary, the so-called uniform model underestimates the stabilizing inlet friction and therefore it gives non-conservative results. With the use of the more realistic non-uniform power profile model, it was possible to determine that, for a CAREM-25 steam generator, the most stable conditions is found at 60MW when the reactor operates at nominal pressure. Moreover, it is found that at high power levels the stability performance is dominated by the two-phase friction component while at low power levels the friction component originated in the over heated steam region prevail (author)

Continuous-Wave Radar to Detect Defects Within Heat Exchangers and Steam Generator Tubes; Revised September 3, 2003

International Nuclear Information System (INIS)

Rochau, Gary E.; Caffey, Thurlow W.H.; Bahram Nassersharif; Garcia, Gabe V.; Jedlicka, Russell P.

2003-01-01

OAK B204 Continuous-Wave Radar to Detect Defects Within Heat Exchangers and Steam Generator Tubes ; Revised September 3, 2003. A major cause of failures in heat exchangers and steam generators in nuclear power plants is degradation of the tubes within them. The tube failure is often caused by the development of cracks that begin on the outer surface of the tube and propagate both inwards and laterally. A new technique was researched for detection of defects using a continuous-wave radar method within metal tubing. The technique is 100% volumetric, and may find smaller defects, more rapidly, and less expensively than present methods. The project described in this report was a joint development effort between Sandia National Laboratories (SNL) and New Mexico State University (NMSU) funded by the US Department of Energy. The goal of the project was to research, design, and develop a new concept utilizing a continuous wave radar to detect defects inside metallic tubes and in particular nuclear plant steam generator tubing. The project was divided into four parallel tracks: computational modeling, experimental prototyping, thermo-mechanical design, and signal detection and analysis

Continuous-Wave Radar to Detect Defects Within Heat Exchangers and Steam Generator Tubes ; Revised September 3, 2003

Energy Technology Data Exchange (ETDEWEB)

Gary E. Rochau and Thurlow W.H. Caffey, Sandia National Laboratories, Albuquerque, NM 87185-0740; Bahram Nassersharif and Gabe V. Garcia, Department of Mechanical Engineering, New Mexico State University, Las Cruces, NM 88003-8001; Russell P. Jedlicka, Klipsch School of Electrical and Computer Engineering, New Mexico State University, Las Cruces, NM 88003-8001

2003-05-01

OAK B204 Continuous-Wave Radar to Detect Defects Within Heat Exchangers and Steam Generator Tubes ; Revised September 3, 2003. A major cause of failures in heat exchangers and steam generators in nuclear power plants is degradation of the tubes within them. The tube failure is often caused by the development of cracks that begin on the outer surface of the tube and propagate both inwards and laterally. A new technique was researched for detection of defects using a continuous-wave radar method within metal tubing. The technique is 100% volumetric, and may find smaller defects, more rapidly, and less expensively than present methods. The project described in this report was a joint development effort between Sandia National Laboratories (SNL) and New Mexico State University (NMSU) funded by the US Department of Energy. The goal of the project was to research, design, and develop a new concept utilizing a continuous wave radar to detect defects inside metallic tubes and in particular nuclear plant steam generator tubing. The project was divided into four parallel tracks: computational modeling, experimental prototyping, thermo-mechanical design, and signal detection and analysis.

Themoeconomic optimization of triple pressure heat recovery steam generator operating parameters for combined cycle plants

Directory of Open Access Journals (Sweden)

Mohammd Mohammed S.

2015-01-01

Full Text Available The aim of this work is to develop a method for optimization of operating parameters of a triple pressure heat recovery steam generator. Two types of optimization: (a thermodynamic and (b thermoeconomic were preformed. The purpose of the thermodynamic optimization is to maximize the efficiency of the plant. The selected objective for this purpose is minimization of the exergy destruction in the heat recovery steam generator (HRSG. The purpose of the thermoeconomic optimization is to decrease the production cost of electricity. Here, the total annual cost of HRSG, defined as a sum of annual values of the capital costs and the cost of the exergy destruction, is selected as the objective function. The optimal values of the most influencing variables are obtained by minimizing the objective function while satisfying a group of constraints. The optimization algorithm is developed and tested on a case of CCGT plant with complex configuration. Six operating parameters were subject of optimization: pressures and pinch point temperatures of every three (high, intermediate and low pressure steam stream in the HRSG. The influence of these variables on the objective function and production cost are investigated in detail. The differences between results of thermodynamic and the thermoeconomic optimization are discussed.

Advanced life-cycle management for an increased steam generator performance

International Nuclear Information System (INIS)

Beck, J.; Schwarz, T.; Bouecke, R.; Schneider, V.

2006-01-01

High steam generators performance is a prerequisite for high plant availability and possible life time extension. During operation, the performance is reduced by fouling of the heating tubes and by corrosion, resulting on a reduction of the heat-exchange area. Such steam generator degradation problems arise from mechanical degradation and a continuous ingress of non-volatile contaminants, i.e. corrosion products and salt impurities accumulated in the steam generators. In addition, the tube scales in general affect the steam generator thermal performance, which ultimately cause a reduction of power output. AREVA applied an integrated service for utilities to evaluate all operational parameters influencing the steam generator performance. The evaluation is assisted by a systematic approach to evaluate the major steam generator operational data. The different data are structured and indexed in a Cleanling-Matrix. The result of this matrix is a quantified, dimensionless figure, given as the Fouling Index. The Fouling Index allows to monitor the condition of steam generators, compare it to other plants and, in combination with a life-time management applied at several German utilities, it allows verified statements on the past operation. Based on these data, an extrapolation of the potential additional life-time of the component is possible. As such, the Fouling Index is a valuable tool concerning life-time extension considerations. The application of the cleanliness criteria in combination with operational data with respect to life-time monitoring and improvements of steam generator performance are presented. (author)

Steam generator assessment for sustainable power plant operation

International Nuclear Information System (INIS)

Drexler, Andreas; Fandrich, Joerg; Ramminger, Ute; Montaner-Garcia, Violeta

2012-09-01

Water and steam serve in the water-steam cycle as the energy transport and work media. These fluids shall not affect, through corrosion processes on the construction materials and their consequences, undisturbed plant operation. The main objectives of the steam water cycle chemistry consequently are: - The metal release rates of the structural materials shall be minimal - The probability of selective / localized forms of corrosion shall be minimal. - The deposition of corrosion products on heat transfer surfaces shall be minimized. - The formation of aggressive media, particularly local aggressive environments under deposits, shall be avoided. These objectives are especially important for the steam generators (SGs) because their condition is a key factor for plant performance, high plant availability, life time extension and is important to NPP safety. The major opponent to that is corrosion and fouling of the heating tubes. Effective ways of counteracting all degradation problems and thus of improving the SG performance are to keep SGs in clean conditions or if necessary to plan cleaning measures such as mechanical tube sheet lancing or chemical cleaning. Based on more than 40 years of experience in steam-water cycle water chemistry treatment AREVA developed an overall methodology assessing the steam generator cleanliness condition by evaluating all available operational and inspection data together. In order to gain a complete picture all relevant water chemistry data (e.g. corrosion product mass balances, impurity ingress), inspection data (e.g. visual inspections and tube sheet lancing results) and thermal performance data (e.g. heat transfer calculations) are evaluated, structured and indexed using the AREVA Fouling Index Tool Box. This Fouling Index Tool Box is more than a database or statistical approach for assessment of plant chemistry data. Furthermore the AREVA's approach combines manufacturer's experience with plant data and operates with an

Thermoelectric generation coupling methanol steam reforming characteristic in microreactor

International Nuclear Information System (INIS)

Wang, Feng; Cao, Yiding; Wang, Guoqiang

2015-01-01

Thermoelectric (TE) generator converts heat to electric energy by thermoelectric material. However, heat removal on the cold side of the generator represents a serious challenge. To address this problem and for improved energy conversion, a thermoelectric generation process coupled with methanol steam reforming (SR) for hydrogen production is designed and analyzed in this paper. Experimental study on the cold spot character in a micro-reactor with monolayer catalyst bed is first carried out to understand the endothermic nature of the reforming as the thermoelectric cold side. A novel methanol steam reforming micro-reactor heated by waste heat or methanol catalytic combustion for hydrogen production coupled with a thermoelectric generation module is then simulated. Results show that the cold spot effect exists in the catalyst bed under all conditions, and the associated temperature difference first increases and then decreases with the inlet temperature. In the micro-reactor, the temperature difference between the reforming and heating channel outlets decreases rapidly with an increase in thermoelectric material's conductivity coefficient. However, methanol conversion at the reforming outlet is mainly affected by the reactor inlet temperature; while at the combustion outlet, it is mainly affected by the reactor inlet velocity. Due to the strong endothermic effect of the methanol steam reforming, heat supply of both kinds cannot balance the heat needed at reactor local areas, resulting in the cold spot at the reactor inlet. When the temperature difference between the thermoelectric module's hot and cold sides is 22 K, the generator can achieve an output voltage of 55 mV. The corresponding molar fraction of hydrogen can reach about 62.6%, which corresponds to methanol conversion rate of 72.6%. - Highlights: • Cold spot character of methanol steam reforming was studied through experiment. • Thermoelectric generation Coupling MSR process has been

Heat exchanging tube behaviour in steam generators of pressurized water reactors

International Nuclear Information System (INIS)

Pastor, D.; Oertel, K.

1979-01-01

Based on a comprehensive failure statistics, materials corrosion chemistry and thermohydraulics problems of the tubings of steam generators are considered. A historical review of failures in the tubings of steam generators in pressurized water reactors reflects the often successless measures by designers, manufacturers and operating organizations for preventing failures, especially with regard to materials selection and water regime. It is stated that laboratory tests could not give sufficient information about safe and stable operation of nuclear steam generators unless real constructive, hydrodynamic, thermodynamical and chemical conditions of operation had been taken into account. (author)

Certification of materials for steam generator condensor and regeneration heat exchanger for nuclear plant

International Nuclear Information System (INIS)

Stevanovicj, M.V.; Jovashevicj, V.J.; Jovashevicj, V.D.J.; Spasicj, Zh.Lj.

1977-01-01

In the construction of a nuclear power plant almost all known materials are used. The choice depends on working conditions. In this work standard specifications of contemporary materials that take part in larger quantities in the following components of the secondary circuit of PWR-type nuclear power plant are proposed: steam generator with moisture separator, condensor and regenerative heat eXchanger

Inverted Steam Generators for Sodium Cooled Fast Reactors

International Nuclear Information System (INIS)

Matal, Oldřich; Šimo, Tomáš; Matal, Oldřich Jr.

2013-01-01

Conclusions: Two inverted steam generators of the Czech industry provenience have still been in successful operation with no water into sodium leaks at BOR 60 (RIAR Dimitrovgrad, Russian Federation). Micromodular inverted steam generator (MMISG) since 1981 and modular inverted steam generator (MISG) since 1991. In the framework of the CP ESFR project predesign studies of 100 MW (thermal) ISG modules were performed with the consideration of MMISG and MISG design, operational and safety benefits and experience. Development of material and technology for sodium heated steam generators components reflecting contemporary domestic industrial conditions in the Czech Republic was restarted in the years 2003 to 2004 and supported in the years 2008 to 2011 by the European CP ESFR project and by the Ministry of Industry and Trade of the Czech Republic

Fuqing nuclear power of nuclear steam turbine generating unit No.1 at the implementation and feedback

International Nuclear Information System (INIS)

Cao Yuhua; Xiao Bo; He Liu; Huang Min

2014-01-01

The article introduces the Fuqing nuclear power of nuclear steam turbine generating unit no.l purpose, range of experience, experiment preparation, implementation, feedback and response. Turn of nuclear steam turbo-generator set flush, using the main reactor coolant pump and regulator of the heat generated by the electric heating element and the total heat capacity in secondary circuit of reactor coolant system (steam generator secondary side) of saturated steam turbine rushed to 1500 RPM, Fuqing nuclear power of nuclear steam turbine generating unit no.1 implementation of the performance of the inspection of steam turbine and its auxiliary system, through the test problems found in the clean up in time, the nuclear steam sweep turn smooth realization has accumulated experience. At the same time, Fuqing nuclear power of nuclear steam turbine generating unit no.1 at turn is half speed steam turbine generator non-nuclear turn at the first, with its smooth realization of other nuclear power steam turbine generator set in the field of non-nuclear turn play a reference role. (authors)

Tests of an experimental steam generator heated by sodium-potassium flow

International Nuclear Information System (INIS)

Robin, M.G.; Duchatelle, L.; Nucheze, L. de

1974-01-01

The first boiler heated by liquid alkaline metal flow, installed at the C.E.A. Heat Test Centre at Grand-Quevilly, near Rouen, was a double-walled 5MW model fed with water at 220 deg C and producing superheated steam at 545 deg C and at an effective pressure of 125 bar. From 1965 to 1967, in the course of more than 8,000 hours of operation under varied conditions (specific water flow rates between 110 and 1,200kg/sq.m./sec., steam pressures of 85 and 125 bar), some 200 results of stationary operating conditions were obtained. These were compared with calculated predictions. After specifying the laws of exchange utilized, the authors discuss the heat transfer results and pressure loss values [fr

Soviet steam generator technology: fossil fuel and nuclear power plants

International Nuclear Information System (INIS)

Rosengaus, J.

1987-01-01

In the Soviet Union, particular operational requirements, coupled with a centralized planning system adopted in the 1920s, have led to a current technology which differs in significant ways from its counterparts elsewhere in the would and particularly in the United States. However, the monograph has a broader value in that it traces the development of steam generators in response to the industrial requirements of a major nation dealing with the global energy situation. Specifically, it shows how Soviet steam generator technology evolved as a result of changing industrial requirements, fuel availability, and national fuel utilization policy. The monograph begins with a brief technical introduction focusing on steam-turbine power plants, and includes a discussion of the Soviet Union's regional power supply (GRES) networks and heat and power plant (TETs) systems. TETs may be described as large central co-generating stations which, in addition to electricity, provide heat in the form of steam and hot water. Plants of this type are a common feature of the USSR today. The adoption of these cogeneration units as a matter of national policy has had a central influence on Soviet steam generator technology which can be traced throughout the monograph. The six chapters contain: a short history of steam generators in the USSR; steam generator design and manufacture in the USSR; boiler and furnace assemblies for fossil fuel-fired power stations; auxiliary components; steam generators in nuclear power plants; and the current status of the Soviet steam generator industry. Chapters have been abstracted separately. A glossary is included containing abbreviations and acronyms of USSR organizations. 26 references

Chemical control and design considerations for CANDU-PHW steam generators

International Nuclear Information System (INIS)

Frost, C.R.; Churchill, B.R.

1978-01-01

Ontario Hydro presently operates eight nuclear power units with a total capacitiy of about 4000 MW(e) net. Operating experience has been with Monel-400 and with Inconel-600 tubed steam generators using sodium phosphate or all volatile control of the boiler steam and water system. With a heavy water Heat Transport System, steam generator tube integrity is an essential ingredient of economical power production. Only three steam generator tube failures have occurred so far in about 40 unit-years operation. None was attributable to corrosion. Factors in the good reliability are, careful engineering design, good quality control at all stages of tubing and steam generator manufacture and close chemical control. The continuing evolution of our steam generator design means that future requirements will be more stringent. (author)

On the evaluation of lifetime of evaporative tubes of once-through steam generators at steam-generating surface temperature oscillations in the burnout region

International Nuclear Information System (INIS)

Vorob'ev, V.A.; Loshchinin, V.M.; Remizov, O.V.

1978-01-01

Suggested is a method for evaluation of a stressed state of evaporation tubes of once-through steam generators at temperature oscillations in the burnout region. Calculated is the amplitude of steam-generating surface temperature oscillations in the burnout region depending on the frequency of a liquid-steam boundary transfer and on this basis determined are thermal stresses in a tube wall. Knowing a fatigue curve gives the possibility to evaluate a heat transfer tube lifetime

Modular steam generator for use in nuclear power plants

International Nuclear Information System (INIS)

Cella, A.

1979-01-01

An improved steam generator for a PWR is described. A turbine generator is driven by the steam output of the steam generator to provide electrical power. The improvement provides vertically assemblable modules which are removably mounted together in sealing relationship. The modules comprising a base module, a tube bundle module removably mountable on the base module in sealing relationship, and an uppermost dryer module removably mountable on the tube bundle module in sealing relationship. Ready access to and removal of the tube bundle module in situ from the nuclear power plant steam generator is facilitated. The dryer module contains moisture separator for drying the generated steam. The base module, upon which the associated weight of the vertically assembled dryer module and tube bundle module are supported, contains the inlet and outlet for the heat exchange fluid. The tube bundle module contains the tube bundle through which the heat exchange fluid flows as well as an inlet for feedwater. The tube sheet serves as a closure flange for the tube bundle module, with the associated weight of the vertically assembled dryer module and tube bundle module on the tube sheet closure flange effectuating the sealing relationship between the base module and the tube bundle module for facilitating closure

HTGR steam generator development

International Nuclear Information System (INIS)

Schuetzenduebel, W.G.; Hunt, P.S.; Weber, M.

1976-01-01

More than 40 gas-cooled reactor plants have produced in excess of 400 reactor years of operating experience which have proved a reasonably high rate of gas-cooled reactor steam generator availability. The steam generators used in these reactors include single U-tube and straight-tube steam generators as well as meander type and helically wound or involute tube steam generators. It appears that modern reactors are being equipped with helically wound steam generators of the once-through type as the end product of steam generator evolution in gas-cooled reactor plants. This paper provides a general overview of gas-cooled reactor steam generator evolution and operating experience and shows how design criteria and constraints, research and development, and experience data are factored into the design/development of modern helically wound tube steam generators for the present generation of gas-cooled reactors

Leakage experiences with 1 MW steam generator

International Nuclear Information System (INIS)

Kanamori, A.; Kawara, M.; Sano, A.

1975-01-01

An 1 MW steam generator was tested from October, 1971 and completed with the first series of experiments by May, 1972 after 3600 hours of operation. During these tests, unextraordinary heat absorption was experienced in the downcomer region, which led to shortage of heat transfer area to attain the rated steam temperature and to one of the reasons of flow instabilities. The steam generator was disassembled to get test pieces for structure as well as material examinations and then it was reassembled to proceed the second series of tests. Before it was done, a modification was provided to insulate the downcomer region by putting a gas space around the downcomer tube. The gas space was provided by a dual tube and spacers were welded on the inner tube and an end plate was welded on upper parts between the two to seal the gap by means of fillet welding. After the modified steam generator was put into operation, water happened to leak into a sodium side two times through these additional welding spots for the gas insulation. This paper presents operating conditions and behaviors of monitors at the time of the leakages, identifications of leaked spots, an evaluation of causes and a treatment or a precaution for them

Use of virtual steam generator cassette for tube spatial design and SGC assembling procedure

International Nuclear Information System (INIS)

Kim, Y. W.; Kim, J. I.; Ji, S. K.

2003-01-01

A method of determining spatial arrangement of tube connection and assembling procedure of once-through helical steam generator cassette utilizing three dimensional virtual steam generator cassette has been developed on the basis of recent 3-D modelling technology. One ends of the steam generator tubes are connected to the module feed water header and the other sides are connected to the module steam header. Due to the complex geometry of tube arrangement, it is very difficult to connect the tubes to the module headers without the help of a physical engineering mock up. A comparative study has been performed at each design step for the tube arrangement and heat transfer area. Heat transfer area computed from thermal sizing was 4% less than that of measured. Heat transfer area calculated from the virtual steam generator cassette mock up has only 0.2% difference with that of measured. Assembling procedure of the steam generator cassette also, can be developed in the design stage

Energy Analysis of Cascade Heating with High Back-Pressure Large-Scale Steam Turbine

Directory of Open Access Journals (Sweden)

Zhihua Ge

2018-01-01

Full Text Available To reduce the exergy loss that is caused by the high-grade extraction steam of traditional heating mode of combined heat and power (CHP generating unit, a high back-pressure cascade heating technology for two jointly constructed large-scale steam turbine power generating units is proposed. The Unit 1 makes full use of the exhaust steam heat from high back-pressure turbine, and the Unit 2 uses the original heating mode of extracting steam condensation, which significantly reduces the flow rate of high-grade extraction steam. The typical 2 × 350 MW supercritical CHP units in northern China were selected as object. The boundary conditions for heating were determined based on the actual climatic conditions and heating demands. A model to analyze the performance of the high back-pressure cascade heating supply units for off-design operating conditions was developed. The load distributions between high back-pressure exhaust steam direct supply and extraction steam heating supply were described under various conditions, based on which, the heating efficiency of the CHP units with the high back-pressure cascade heating system was analyzed. The design heating load and maximum heating supply load were determined as well. The results indicate that the average coal consumption rate during the heating season is 205.46 g/kWh for the design heating load after the retrofit, which is about 51.99 g/kWh lower than that of the traditional heating mode. The coal consumption rate of 199.07 g/kWh can be achieved for the maximum heating load. Significant energy saving and CO2 emission reduction are obtained.

Predicting steam generator crevice chemistry

International Nuclear Information System (INIS)

Burton, G.; Strati, G.

2006-01-01

'Full text:' Corrosion of steam cycle components produces insoluble material, mostly iron oxides, that are transported to the steam generator (SG) via the feedwater and deposited on internal surfaces such as the tubes, tube support plates and the tubesheet. The build up of these corrosion products over time can lead to regions of restricted flow with water chemistry that may be significantly different, and potentially more corrosive to SG tube material, than the bulk steam generator water chemistry. The aim of the present work is to predict SG crevice chemistry using experimentation and modelling as part of AECL's overall strategy for steam generator life management. Hideout-return experiments are performed under CANDU steam generator conditions to assess the accumulation of impurities in hideout, and return from, model crevices. The results are used to validate the ChemSolv model that predicts steam generator crevice impurity concentrations, and high temperature pH, based on process parameters (e.g., heat flux, primary side temperature) and blowdown water chemistry. The model has been incorporated into ChemAND, AECL's system health monitoring software for chemistry monitoring, analysis and diagnostics that has been installed at two domestic and one international CANDU station. ChemAND provides the station chemists with the only method to predict SG crevice chemistry. In one recent application, the software has been used to evaluate the crevice chemistry based on the elevated, but balanced, SG bulk water impurity concentrations present during reactor startup, in order to reduce hold times. The present paper will describe recent hideout-return experiments that are used for the validation of the ChemSolv model, station experience using the software, and improvements to predict the crevice electrochemical potential that will permit station staff to ensure that the SG tubes are in the 'safe operating zone' predicted by Lu (AECL). (author)

Accident alarm equipment for steam generator, especially liquid sodium heated steam generator

International Nuclear Information System (INIS)

Matal, O.; Jung, J.; Banovec, J.

1982-01-01

The alarm equipment consists of a system of sensors mounted onto the steam generator and its accessories. Each of the sensors is used for a different accident characteristic, such as the flow of sodium, the acoustic spectrum, the concentration of hydrogen in sodium. The system of sensors is connected to the common accident alarm system. The equipment will not issue the alarm signal if it receives a message from only one sensor, only when the message is confirmed from other sensors. This excludes false alarm. (M.D.)

Enrichment of water solube substances at heat transfer surfaces in steam boilers and steam generators - a literature survey

International Nuclear Information System (INIS)

Kelen, T.

1975-03-01

A literature survey has been made to determine the possible need for investigations into enrichment at heat transport surfaces. The survey shows that enrichment in furnace tubes of carbon steel has led both to hydrogen embrittlement as a result of a reduction of pH and to local tube wall thinning and alkaline stress corrosion following an increase of pH. Damage caused by enrichment has occurred also in the steam generators of PWR's. Information concerning measured enrichment, observed hide-out, occurrence of normally easily soluble deposits as well as of corrosion damage requiring high concentrations lead to the conclusion that enrichment factors of the order of 10 000 are to be expected in some instances. The relationship between operational conditions and enrichment in specific cases is, however, poorly documented. The tendency to hide out varies for different substances. This is however not evidently equivalent to variation of the corresponding enrichment factor in the solution. Enrichment at heat transfer surfaces can arise as the result of one or several of the following effects: - formation of porous deposits - poor circulation - high densities of heat flow rate - departure from nucleate boiling leading to dry-out - steam blanketing. The conclusion reached from the literature survey is that knowledge of the degree of enrichment produced under specific operational conditions is extremely inadequate. (author)

Design and construction features of steam generators at a nuclear power station

International Nuclear Information System (INIS)

Chakrabarti, A.K.; Gupta, K.N.; Bapat, C.N.; Sharma, V.K.

1996-01-01

The Indian nuclear power programme is based on Pressurised Heavy Water Reactors (PHWRs) using natural uranium as fuel and heavy water as reactor coolant as well as moderator. The nuclear heat is generated in the fuel located in the pressure tubes. Pressurised heavy water in the primary heat transport (PHT) system is circulated through the tubes which picks up the heat from the fuel and transfers it to ordinary water in steam generators (SGs) to produce steam. The steam is used for providing power to the turbine. The steam generator is a critical equipment in the nuclear steam supply system (NSSS) of a nuclear reactor. SG tube surface area constitute about 80% of total primary circuit surface area. A typical value in a 220 MWe reactor is 9000 m 2 which can release considerable amount of corrosion products unless very low corrosion rates are achieved by proper design, material selection and water chemistry control. Design and construction features of SGs are given. 1 tab

Embalse steam generators - status in 2009

International Nuclear Information System (INIS)

Luna, P.; Yetisir, M.; Roy, S.; MacEacheron, R.

2009-01-01

The Embalse Nuclear Generating Station (ENGS) is a CANDU 6, a pressurized heavy water plant, with a net capacity of 648 MW. The primary heat transport system at Embalse includes four Steam Generators (SGs) manufactured by Babcock and Wilcox Canada (B and W). These steam generators are vertical recirculating heat exchangers with Incoloy 800 inverted U-tubes and an integral preheater. Embalse SGs performed very well until the late 1990s, when an increase in tube fretting was noticed in the U-bend region. In-service inspection in 2002 and 2004 confirmed that the cause of the tube fretting was flow accelerated corrosion (FAC) damage of scallop bar supports in the U-bend region. The straight leg tube support plates (TSPs) have also been degrading. Degradation was worst at the top support plates, and it was in the form of material loss on the cold leg. The hot leg TSPs were heavily fouled with deposits and flow areas were blocked. Visual inspections and subsequent studies showed that the cause of the TSP degradation was also FAC. The Embalse SGs have carbon steel supports that make them susceptible to FAC. To mitigate the effects of degraded tube support structures, three additional sets of anti-vibration bars were installed in the U-bend regions of all four steam generators in 2004. In 2007, an improved secondary-side chemistry specification was implemented to reduce the FAC rate and the hot leg TSPs was waterlanced. A root cause analysis and condition assessment was performed for the tube supports in 2007. Fitness for Service (FFS) evaluation was completed using the Canadian Industry Guidelines for steam generator tubes. The steam generators were returned to service and the plan has operated without another forced outage to date. The FAC degradation of the carbon steel U-bend tube support systems has had the most significant impact on the plant operation causing a number of forced outages. The discovery of the extent of TSP degradation and difficulties to repair TSPs

Results of the secondary side chemical cleaning of the steam generators

International Nuclear Information System (INIS)

Doma, A.; Patek, G.

2001-01-01

A significant amount of deposit has developed on the secondary side of the heat transfer tubes of the steam generators (SG) of the Paks Nuclear Power Plant units in course of the years. More than 99.5% of the deposit is made up of magnetite (Fe 3 O 4 ) generated in the secondary circuit of the power plant. Those deposits lead to the decrease of the heat transfer. Even more important is its role from the point of view of operational reliability of the steam generators, leak tightness between the primary and secondary sides. The first series of cleaning took place following 8-9 years of operation of the units. Following the first cleaning cycle the transport of the corrosion products into the steam generators did not change, and thus obviously new cleaning was required. Periodical cleaning of the steam generators shall be assured. (R.P.)

Steam generator life-management, reliability, maintenance and refurbishment

International Nuclear Information System (INIS)

Spekkens, P.

2012-01-01

SGC 2012 is a different kind of a conference - it has its own focus, initiatives and objectives and differs from its predecessors. It originated as the Steam Generator and Heat Exchanger Conference in 1990 - a time when premature degradation of steam generators with Alloy 600 tubes was rampant world-wide, and some CANDU steam generators had started to experience significant fouling and corrosion issues. The six previous steam generator conferences were held on a regular cycle, in a very similar format and with a similar theme. We are now in a different era in steam generators. The Alloy 600 tubing has been largely replaced by more robust materials, and the CANDU steam generators have been brought under much more intense and effective life cycle management. Performance of steam generators has improved greatly, and they are no longer considered at risk of limiting the life of the units. Indeed, most Incoloy 800 steam generators in CANDU units are considered to be capable of operating reliably through the 'second life' of the units and are not being replaced during refurbishments. Given this changing environment, the scope of this conference has been expanded from one to three areas: steam generators and heat exchangers as before, but also; controls, valves and pumps, and; reactor components and systems, Programs A, B and C, respectively. The conference is targeting to address the needs and interests of the operating utilities, and to 'focus on what needs attention'. As a means of 'focusing on what needs attention' an 'Issue-Identification and Definition' program was initiated last winter. The Issue-Identification Team operating with COG President Bob Morrison as its Executive Lead, worked to identify issues requiring attention in the three areas of interest. Of the many issues identified by the Team and elaborated on by the Program Developers of this conference, four were recommended for special attention: A. 'Operate Clean - Build Clean - Plant Wide': Despite their

Integration of steam injection and inlet air cooling for a gas turbine generation system

International Nuclear Information System (INIS)

Wang, F.J.; Chiou, J.S.

2004-01-01

The temperature of exhaust gases from simple cycle gas turbine generation sets (GENSETs) is usually very high (around 500 deg. C), and a heat recovery steam generator (HRSG) is often used to recover the energy from the exhaust gases and generate steam. The generated steams can be either used for many useful processes (heating, drying, separation etc.) or used back in the power generation system for enhancing power generation capacity and efficiency. Two well-proven techniques, namely steam injection gas turbine (STIG) and inlet air cooling (IAC) are very effective features that can use the generated steam to improve the power generation capacity and efficiency. Since the energy level of the generated steam needed for steam injection is different from that needed by an absorption chiller to cool the inlet air, a proper arrangement is required to implement both the STIG and the IAC features into the simple cycle GENSET. In this study, a computer code was developed to simulate a Tai power's Frame 7B simple cycle GENSET. Under the condition of local summer weather, the benefits obtained from the system implementing both STIG and IAC features are more than a 70% boost in power and 20.4% improvement in heat rate

Sodium and steam leak simulation studies for fluidized bed steam generators

International Nuclear Information System (INIS)

Keeton, A.R.; Vaux, W.G.; Lee, P.K.; Witkowski, R.E.

1976-01-01

An experimental program is described which was conducted to study the effects of sodium or steam leaking into an operating fluidized bed of metal or ceramic particles at 680 to 800 0 K. This effort was part of the early development studies for a fluidized-bed steam generator concept using helium as the fluidizing gas. Test results indicated that steam and small sodium leaks had no effect on the quality of fluidization, heat transfer coefficient, temperature distribution, or fluidizing gas pressure drop across the bed. Large sodium leaks, however, immediately upset the operation of the fluidized bed. Both steam and sodium leaks were detected positively and rapidly at an early stage of a leak by instruments specifically selected to accomplish this

SWAAM code development, verification and application to steam generator design

International Nuclear Information System (INIS)

Shin, Y.W.; Valentin, R.A.

1990-01-01

This paper describes the family of SWAAM codes developed by Argonne National Laboratory to analyze the effects of sodium/water reactions on LMR steam generators. The SWAAM codes were developed as design tools for analyzing various phenomena related to steam generator leaks and to predict the resulting thermal and hydraulic effects on the steam generator and the intermediate heat transport system (IHTS). The theoretical foundations and numerical treatments on which the codes are based are discussed, followed by a description of code capabilities and limitations, verification of the codes by comparison with experiment, and applications to steam generator and IHTS design. (author). 25 refs, 14 figs

Summary of results from sodium-heated steam generator test program

Energy Technology Data Exchange (ETDEWEB)

McDonald, J S

1975-07-01

A 28 MWt sodium-heated steam generator test unit developed and fabricated by Atomics International was operated in the Sodium Component Test Installation. The SCTI is located at the Liquid Metal Engineering Center which is operated for the Atomic Energy Commission by Atomics International, Reviewed in this paper are the results of the test operations and the findings of the post-test examination of the module. Testing was performed to assure the mechanical integrity of the unit over a wide range of simulated plant operating conditions and to develop a variety of performance data. Specific tests conducted included preheat, vibration, startup-shutdown, pressurization, steady state and parametric performance mapping, endurance, simulated leak injection, low- flow stability and simulated plant transients. (author)

Stability study in one step steam generators

International Nuclear Information System (INIS)

Abbate, P.

1990-01-01

The TWO program is presented developed for the behaviour limit calculation stable in one step steam generators for the case of Density Waves phenomenom. The program is based on a nodal model which, using Laplace transformation equations, allows to study the system's transfer functions and foresee the beginning of the unstable behaviour. This program has been satisfactorily validated against channels data uniformly heated in the range from 4.0 to 6.0 Mpa. Results on the CAREM reactor's steam generator analysis are presented. (Author) [es

Steam turbines for nuclear power stations in Czechoslovakia and their use for district heating

International Nuclear Information System (INIS)

Drahy, J.

1989-01-01

The first generation of nuclear power stations in Czechoslavakia is equipped with 440 MW e pressurized water reactors. Each reactor supplies two 220 MW, 3000 rpm condensing type turbosets operating with saturated steam. After the completion of heating water piping systems, all of the 24 units of 220 MW in Czechoslovak nuclear power stations will be operated as dual purpose units, delivering both electricity and heat. At the present time, second-generation nuclear power stations, with 1000 MW e PWRs, are being built. Each such plant is equipped with one 1000 MW full-speed saturated steam turbine. The turbine is so designed as to permit the extraction of steam corresponding to the following quantities of heat: 893 MJ/s with three-stage water heating (150/60 0 C); and 570 MJ/s with two-stage water heating (120/60 0 C). The steam is taken from uncontrolled steam extraction points. (author)

Influence of recycling ratio on steam generator thermal recycling

International Nuclear Information System (INIS)

Bassel, W.S.; Rodrigues, V.G.

1989-01-01

A mathematical model was developed to simulate thermal performance of steam generator. The simulation was done with 3 control volumes. The coupled non-linear algebric equations, where the heat transfer was calculated with logarithmic meam temperature difference, was solved by iterative method. The developed model is suitable for calculation the parameters which effect the performance of steam generator. (author) [pt

Development of steam generators for combustion of biofuels up to 10 t/h

Energy Technology Data Exchange (ETDEWEB)

Bentzin, H

1985-01-01

Combustion parameters are compared for raw brown coal, rice hulls and coconut shells as fuel in small steam generators. Combustion of native biofuel is seen as a power generation alternative in developing countries. Experiments were conducted on a 6.5 t/h moving grate steam generator with a firing grate surface of 7.2 m/sup 2/. Combustion results are shown in a table. Technological modifications carried out in adapting brown coal-fired steam generators to biofuels are also listed. A series of small steam generators for combustion of brown coal, biofuels including wood chips, as well as heating oil as back-up has been developed by the Karl-Marx-Stadt Dampfkesselbau Plant, GDR, with steam capacities ranging from 3.2 to 10 t/h. Technical specifications and diagrams of this series design (DGK-3, DGK-45, DWK 2S) are given. A larger steam generator with 20 t/h steam capacity for combustion of raw brown coal, bagasse, wood chips with heating oil and for rice hulls as support fuels is being developed by the Berlin Dampferzeugerbau Plant, GDR. 5 references.

Wasteless combined aggregate-coal-fired steam-generator/melting-converter

International Nuclear Information System (INIS)

Pioro, L.S.; Pioro, I.L.

2003-01-01

A method of reprocessing coal sludge and ash into granulate for the building industry in a combined wasteless aggregate-steam-generator/melting-converter was developed and tested. The method involves melting sludge and ash from coal-fired steam-generators of power plants in a melting-converter installed under the steam-generator, with direct sludge drain from the steam generator combustion chamber. The direct drain of sludge into converter allows burnup of coal with high ash levels in the steam-generator without an additional source of ignition (natural gas, heating oil, etc.). Specific to the melting process is the use of a gas-air mixture with direct combustion inside a melt. This feature provides melt bubbling and helps to achieve maximum heat transfer from combustion products to the melt, to improve mixing, to increase rate of chemical reactions and to improve the conditions for burning the carbon residue from the sludge and ash. The 'gross' thermal efficiency of the combined aggregate is about 93% and the converter capacity is about 18 t of melt in 100 min. The experimental data for different aspects of the proposed method are presented. The effective ash/charging materials feeding system is also discussed. The reprocessed coal ash and sludge in the form of granules can be used as fillers for concretes and as additives in the production of cement, bricks and other building materials

Pressure drop-flow rate curves for single-phase steam in Combustion Engineering type steam generator U-tubes during severe accidents

Energy Technology Data Exchange (ETDEWEB)

Fynan, Douglas A.; Ahn, Kwang-Il, E-mail: kiahn@kaeri.re.kr

2016-12-15

Highlights: • Pressure drop-flow rate curves for superheated steam in U-tubes were generated. • Forward flow of hot steam is favored in the longer and taller U-tubes. • Reverse flow of cold steam is favored in short U-tubes. • Steam generator U-tube bundle geometry and tube diameter are important. • Need for correlation development for natural convention heat transfer coefficient. - Abstract: Characteristic pressure drop-flow rate curves are generated for all row numbers of the OPR1000 steam generators (SGs), representative of Combustion Engineering (CE) type SGs featuring square bend U-tubes. The pressure drop-flow rate curves are applicable to severe accident natural circulations of single-phase superheated steam during high pressure station blackout sequences with failed auxiliary feedwater and dry secondary side which are closely related to the thermally induced steam generator tube rupture event. The pressure drop-flow rate curves which determine the recirculation rate through the SG tubes are dependent on the tube bundle geometry and hydraulic diameter of the tubes. The larger CE type SGs have greater variation of tube length and height as a function of row number with forward flow of steam favored in the longer and taller high row number tubes and reverse flow favored in the short low row number tubes. Friction loss, natural convection heat transfer coefficients, and temperature differentials from the primary to secondary side are dominant parameters affecting the recirculation rate. The need for correlation development for natural convection heat transfer coefficients for external flow over tube bundles currently not modeled in system codes is discussed.

Solar energy for steam generation in Brazil

Energy Technology Data Exchange (ETDEWEB)

De Carvalho, Jr, A V; Orlando, A DeF; Magnoli, D

1979-05-01

Steam generation is a solar energy application that has not been frequently studied in Brazil, even though for example, about 10% of the national primary energy demand is utilized for processing heat generation in the range of 100 to 125/sup 0/C. On the other hand, substitution of automotive gasoline by ethanol, for instance, has received much greater attention even though primary energy demand for process heat generation in the range of 100 to 125/sup 0/C is of the same order of magnitude than for total automotive gasoline production. Generation of low-temperature steam is analyzed in this article using distributed systems of solar collectors. Main results of daily performance simulation of single flat-plate collectors and concentrating collectors are presented for 20/sup 0/S latitude, equinox, in clear days. Flat plate collectors considered are of the aluminum roll-bond absorber type, selective surface single or double glazing. Considering feedwater at 20/sup 0/C, saturated steam at 120/sup 0/C and an annual solar utilization factor of 50%, a total collector area of about 3,000 m/sup 2/ is necessary for the 10 ton/day plant, without energy storage. A fuel-oil back-up system is employed to complement the solar steam production, when necessary. Preliminary economic evaluation indicates that, although the case-study shows today a long payback period relative to subsidized fuel oil in the domestic market (over 20 years in the city of Rio de Janeiro), solar steam systems may be feasible in the medium term due to projected increase of fuel oil price in Brazil.

Thermal-hydraulic experiments for the PCHE type steam generator

International Nuclear Information System (INIS)

Shin, C. W.; No, H. C.

2015-01-01

Printed circuit heat exchanger (PCHE) manufactured by HEATRIC is a compact type of the mini-channel heat exchanger. The PCHE is manufactured by diffusion bonding of the chemically-etched plates, and has high heat transfer rate due to a large surface. Therefore, the size of heat exchanger can be reduced by 1/5 - 1/6 and PCHE can be operated under high pressure, high temperature and multi-phase flow. Under such merits, it is used as heat exchanger with various purposes of gas cycle and water cycle. Recently, it is newly suggested as an application of a steam generator. IRIS of MIT and FASES of KAIST conceptually adopted PCHE as a steam generator. When using boiling condition of micro-channel, flow instability is one of the critical issues. Instability may cause unstable mass flow rate, sudden temperature change and system control failure. However instability tests of micro channels using water are very limited because the previous studies were focused on a single tube or other fluid instead of water. In KAIST, we construct the test facility to study the thermal hydraulics and fluid dynamics of the heat exchanger, especially occurrence of instability. By inducing the pressure drop of inlet water, amplitude of oscillation declined by 90%. Finally, the throttling effect was experimentally confirmed that PCHE could be utilized as a steam generator

Steam generator waterlancing at DNGS

International Nuclear Information System (INIS)

Seppala, D.; Malaugh, J.

1995-01-01

Darlington Nuclear Generating Station (DNGS) is a four 900 MW Unit nuclear station forming part of the Ontario Hydro East System. There are four identical steam generators(SGs) per reactor unit. The Darlington SGs are vertical heat exchangers with an inverted U-tube bundle in a cylindrical shell. The DNGS Nuclear Plant Life Assurance Group , a department of DNGS Engineering Services have taken a Proactive Approach to ensure long term SG integrity. Instead of waiting until the tubesheets are covered by a substantial and established hard deposit; DNGS plan to clean each steam generator's tubesheet, first half lattice tube support assembly and bottom of the thermal plate every four years. The ten year business plan provides for cleaning and inspection to be conducted on all four SGs in each unit during maintenance outages (currently scheduled for every four years)

Steam generator

International Nuclear Information System (INIS)

Fenet, J.-C.

1980-01-01

Steam generator particularly intended for use in the coolant system of a pressurized water reactor for vaporizing a secondary liquid, generally water, by the primary cooling liquid of the reactor and comprising special arrangements for drying the steam before it leaves the generator [fr

SWAAM-code development and verification and application to steam generator designs

International Nuclear Information System (INIS)

Shin, Y.W.; Valentin, R.A.

1990-01-01

This paper describes the family of SWAAM codes which were developed by Argonne National Laboratory to analyze the effects of sodium-water reactions on LMR steam generators. The SWAAM codes were developed as design tools for analyzing various phenomena related to steam generator leaks and the resulting thermal and hydraulic effects on the steam generator and the intermediate heat transport system (IHTS). The paper discusses the theoretical foundations and numerical treatments on which the codes are based, followed by a description of code capabilities and limitations, verification of the codes and applications to steam generator and IHTS designs. 25 refs., 14 figs

Heat recovery optimization in a steam-assisted gravity drainage (SAGD) plant

International Nuclear Information System (INIS)

Ashrafi, Omid; Navarri, Philippe; Hughes, Robin; Lu, Dennis

2016-01-01

Pinch Analysis was used to improve the energy performance of a typical steam-assisted gravity drainage (SAGD) process. The objective of this work was to reduce the amount of natural gas used for steam generation in the plant and the associated greenhouse gas emissions. The INTEGRATION software was used to analyze how heat is being used in the existing design and identify inefficient heat exchanges causing excessive use of energy. Several modifications to improve the base case heat exchanger network (HEN) were identified. The proposed retrofit projects reduced the process heating demands by improving the existing heat recovery system and by recovering waste heat and decreased natural gas consumption in the steam production unit by approximately 40 MW, representing approximately 8% of total consumption. As a result, the amount of glycol used to transfer energy across the facility was also reduced, as well as the electricity consumption related to glycol pumping. It was shown that the proposed heat recovery projects reduced natural gas costs by C$3.8 million/y and greenhouse gas emissions by 61,700 t/y of CO_2. - Highlights: • A heat integration study using Pinch analysis was performed in a SAGD process. • Several modifications are suggested to improve the existing heat recovery system. • Heat recovery projects increased boiler feed water and combustion air temperatures. • The proposed modifications reduced natural gas use for steam generation. • Heat recovery significantly reduced operating costs and greenhouse gas emissions.

Vapor generator steam drum spray heat

International Nuclear Information System (INIS)

Fasnacht, F.A. Jr.

1978-01-01

A typical embodiment of the invention provides a combination feedwater and cooldown water spray head that is centrally disposed in the lower portion of a nuclear power plant steam drum. This structure not only discharges the feedwater in the hottest part of the steam drum, but also increases the time required for the feedwater to reach the steam drum shell, thereby further increasing the feedwater temperature before it contacts the shell surface, thus reducing thermal shock to the steam drum structure

Computer codes for simulation of Angra 1 reactor steam generator

International Nuclear Information System (INIS)

Pinto, A.C.

1978-01-01

A digital computer code is developed for the simulation of the steady-state operation of a u-tube steam generator with natural recirculation used in Pressurized Water Reactors. The steam generator is simulated with two flow channel separated by a metallic wall, with a preheating section with counter flow and a vaporizing section with parallel flow. The program permits the changes in flow patterns and heat transfer correlations, in accordance with the local conditions along the vaporizing section. Various sub-routines are developed for the determination of steam and water properties and a mathematical model is established for the simulation of transients in the same steam generator. The steady state operating conditions in one of the steam generators of ANGRA 1 reactor are determined utilizing this programme. Global results obtained agree with published values [pt

Evaluation on mechanical and corrosion properties of steam generator tubing materials

International Nuclear Information System (INIS)

Kim, In Sup; Lee, Byong Whi; Lee, Sang Kyu; Lee, Young Ho; Kim, Jun Whan; Lee, Ju Seok; Kwon, Hyuk Sang; Kim, Su Jung

1998-06-01

Steam generator is one of the major components of nuclear reactor pressure boundary. It's main function os transferring heat which generated in the reactor to turbine generator through steam generator tube. In these days, steam generator tubing materials of operating plant are used Inconel 600 alloys. But according to the operation time, there are many degradation phenomena which included mechanical damage due to flow induced vibration and corrosion damage due to PWSCC, IGA/SCC and pitting etc. Recently Inconel 690 alloys are selected as new and replacement steam generator tubes for domestic nuclear power plant. But there are few study about mechanical and corrosion properties of Inconel 600 and 690. The objectives of this study is to evaluate and compare mechanical and corrosion propertied of steam generator tube materials

Evaluation on mechanical and corrosion properties of steam generator tubing materials

Energy Technology Data Exchange (ETDEWEB)

Kim, In Sup; Lee, Byong Whi; Lee, Sang Kyu; Lee, Young Ho; Kim, Jun Whan; Lee, Ju Seok; Kwon, Hyuk Sang; Kim, Su Jung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1998-06-15

Steam generator is one of the major components of nuclear reactor pressure boundary. It's main function os transferring heat which generated in the reactor to turbine generator through steam generator tube. In these days, steam generator tubing materials of operating plant are used Inconel 600 alloys. But according to the operation time, there are many degradation phenomena which included mechanical damage due to flow induced vibration and corrosion damage due to PWSCC, IGA/SCC and pitting etc. Recently Inconel 690 alloys are selected as new and replacement steam generator tubes for domestic nuclear power plant. But there are few study about mechanical and corrosion properties of Inconel 600 and 690. The objectives of this study is to evaluate and compare mechanical and corrosion propertied of steam generator tube materials.

Steam generation at Rihand STPP Stage 1

Energy Technology Data Exchange (ETDEWEB)

1987-12-01

The steam generation plant at Rihand in India has two 500 MW boilers. The boilers are of the balanced draught, single cell, radiant furnace type, and are controlled automatically. Cochran Thermax shell type auxillary steam boilers are used for preheating air to the main boilers and for heating fuel oil during storage and pumping. Electrostatic precipitators and ash handling plants are provided to keep dust and ash within limits. 2 figs.

Steam generator operating experience: Update for 1984-1986

International Nuclear Information System (INIS)

Frank, L.; Stokley, J.

1988-06-01

This report summarizes operational events and degradation mechanisms affecting pressurized water reactor steam generator integrity, provides updated inspection results reported in 1984, 1985, and 1986, and highlights both prevalent problem areas and advances in improved equipment test practices, preventive measures, repair techniques, and replacement procedures. It describes equipment design features of the three major suppliers and discusses 68 plants in detail. Steam generator degradation mechanisms include intergranular stress corrosion cracking, primary water stress corrosion cracking, pitting, intergranular attack, and vibration wear that effects tube integrity and causes leakage. Plugging, sleeving heat treatment, peening, chemical cleaning, and steam generator replacements are described and regulatory instruments and inspection guidelines for nondestructive evaluations and girth weld cracking are discusses. The report concludes that although degradation mechanisms are generally understood, the elimination of unscheduled plant shutdowns and costly repairs resulting from leaking tubes has not been achieved. Highlights of steam generator research and unresolved safety issues are discussed. 21 refs., 8 tabs

Pressurizer and steam-generator behavior under PWR transient conditions

International Nuclear Information System (INIS)

Wahba, A.B.; Berta, V.T.; Pointner, W.

1983-01-01

Experiments have been conducted in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR), at the Idaho National Engineering Laboratory, in which transient phenomena arising from accident events with and without reactor scram were studied. The main purpose of the LOFT facility is to provide data for the development of computer codes for PWR transient analyses. Significant thermal-hydraulic differences have been observed between the measured and calculated results for those transients in which the pressurizer and steam generator strongly influence the dominant transient phenomena. Pressurizer and steam generator phenomena that occurred during four specific PWR transients in the LOFT facility are discussed. Two transients were accompanied by pressurizer inflow and a reduction of the heat transfer in the steam generator to a very small value. The other two transients were accompanied by pressurizer outflow while the steam generator behavior was controlled

RELAP5 analysis of reflux condensation behavior in heat transfer tube bundle of a steam generator

International Nuclear Information System (INIS)

Minami, Noritoshi; Chikusa, Toshiaki; Nagae, Takashi; Murase, Michio

2007-01-01

In case of loss of the residual heat removal system and other alternative cooling methods under mid-loop operation during shutdown of the pressurized water reactor plant, reflux condensation in the steam generator (SG) may be an effective heat removal mechanism. In reflux condensation experiments 7.2c with injection of nitrogen gas using the BETHSY facility in France, which is a scale model of a pressurized water reactor plant, 34 heat transfer tubes were divided into two kinds of flow patterns, which were steam forward flow and nitrogen reverse flow. In this study, we simulated the BETHSY experiments using the transient analysis code RELAP5. Modifying calculation equations for interfacial friction force and wall friction force between the inlet plenum and heat transfer tubes, nitrogen reverse flow was successfully simulated. In calculations with alteration of the flow area ratio to two flow channels for the heat transfer tube bundle, the number of active tubes with the maximum nitrogen recirculation flow rate agreed rather well with the observed number of active tubes. In calculations with three flow channels for the heat transfer tube bundle, the average number of active tubes in several calculations with different flow area ratios of the three flow channels predicted the number of active tubes well. (author)

CFD evaluation on the thermohydraulic characteristics of tube support plates in steam generator

Energy Technology Data Exchange (ETDEWEB)

Zhang, B.; Zhang, H.; Han, B.; Yang, B.W. [Xi' an Jiaotong Univ. (China). School of Nuclear Science and Technology; Mo, S.J.; Ren, H.B.; Qin, J.M.; Zuo, C.P. [China Nuclear Power Design Co. Ltd., ShenZhen (China)

2016-07-15

The integrity and thermal hydraulic characteristics of steam generator are of great concern in the nuclear industry. The tube support plates (TSP), one of the most important components of the steam generator, not only support the heat transfer tubes, but also affect the flow dynamic and thermal hydraulic characteristics of the secondary-side flow inside the steam generator. Different working conditions, ranging from single-phase adiabatic condition to two-phase high-void boiling condition, are simulated and analyzed. Calculated void fraction, under simple geometry, agrees well with the experiment data whilst the simulated heat transfer coefficient is tremendously close to the empirical correlation. Temperature, void fraction, and velocity distributions in different locations show reasonable distribution. The simulation results indicate that TSP can enhance the heat transfer in the secondary side of the steam generator. On the top of TSP, with the increase in cross-section flow area, the back-flow phenomenon occurs, which might lead to the contamination of precipitation.

Steam generator of FBR type reactor

International Nuclear Information System (INIS)

Hashiguchi, Ko.

1992-01-01

Liquid metal (for example, mercury) which is scarcely reactive with metal sodium is contained and cover gases which are scarcely reactive with the liquid metal are filled in a steam generator of an FBR type reactor and it is closed. The heat of primary sodium is transferred to the liquid metal, which is not reactive with sodium, in a primary thermal conduction portion. Since the temperature of the primary thermal conduction portion is high, the density is extremely low. On the other hand, since a second thermal conduction portion is kept at a single phase and the temperature is lower compared with that of the first thermal conduction portion, the density is kept high. since the density difference and gas jetting speed generate a great circulating force to liquid metal passing the opening of a partition plate, heat can be conducted on the side of water without disposing pumps. The steam concentration in the liquid metal is low being in a single phase of steams, corrosion caused from the outside of pipes of the primary thermal conduction pipe is scarcely promoted. Even if sodium leaks should be caused, since the sodium concentration in the liquid metal is extremely low and the reactivity is low, the temperature of the liquid metal is not elevated. (N.H.)

Design and construction of a steam generator with feedback

International Nuclear Information System (INIS)

Camargo, Camila C.; Placco, Guilherme M.; Guimaraes, Lamartine N.F.

2013-01-01

The EARTH project aims to develop technologies to design and build systems that generate electricity in space, using microreactors. One of the activities within the TERRA project aims to build a closed thermal cycle Rankine type in order to test a Tesla turbine type. The objective of this work is to design and build a steam generator with feedback, which should ensure a satisfactory range of steam supply, security system, feedback system and heating system

Overview of the United States steam generator development programs

Energy Technology Data Exchange (ETDEWEB)

Kaspar, P W; Lowe, P A

1975-07-01

The LMFBR steam generator development program of the USA was initiated to support the development of reliable designs and meaningful performance data for these critical components. Since the steam generators include the structural boundary between heated sodium and water, the consequences of small flaws in the materials that form the boundary are significant. Successful development and demonstration of commercial LMFBR power plants requires the consideration of many factors in addition to the design, construction and operation of a particular plant. Additional factors which must be assessed include: economics, reliability, safety, environment, operability, maintainability and conservation of the resources. In terms of the steam generator these items led to the selection of a single wall tube design using a forced recirculating system for the present Clinch River Breeder Reactor. There are strong economic incentives to use a once-through steam generating system in future designs.

Process for superheating the steam generated by a light water nuclear reactor

International Nuclear Information System (INIS)

Vakil, H.B.; Brown, D.H.

1976-01-01

A process is submitted for superheating the pressurised steam generated in a light water nuclear reactor in which the steam is brought to 340 0 C at least. This superheated steam is used to operate a turbo-generator unit. The characteristic of the process is that an exothermal chemical reaction is used to generate the heat utilised during the superheating stage. The chemical reaction is a mechanisation, oxidation-reduction or hydrogenation reaction [fr

Transient analysis of a U-tube natural circulation steam generator

Energy Technology Data Exchange (ETDEWEB)

Gaikwad, A J; Kumar, Rajesh; Bhadra, Anu; Chakraborty, G; Venkat Raj, V [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai (India)

1994-06-01

A computer code has been developed, for transient thermal-hydraulic analysis of proposed 500 MWe PHWR steam generator. The transient behaviour of a nuclear power plant is very much dependent on the steam generator performance, as it provides a thermal linkage between the primary and secondary systems. Study of dynamics of steam generator is essential for over all power plant dynamics as well as design of control systems for steam generator. A mathematical model has been developed for the simulation of thermal-hydraulic phenomena in a U tube natural circulation steam generator. Fluid model is based on one dimensional, nonlinear, single fluid conservation equations of mass, momentum, energy and equation of state. This model includes coupled two phase flow heat transfer and natural circulation. The model accounts for both compressibility and thermal expansion effects. The process simulation and results obtained for transients such as step change in load and total loss of feed water are presented. (author). 5 refs., 7 figs.

Planning of the steam generators for nuclear applications using optimization techniques

International Nuclear Information System (INIS)

Sakai, M.; Silvares, O.M.

1978-01-01

Procedure for the maximization of the net power of a nuclear power plant through the application of the optimal control theory of dynamic systems is presented. The problem is formulated in the steam generator which links the primary and the secondary cycle. The solution of the steam generator, optimization problem is obtained simultaneously with the heat balance in both primary and secondary cycle, through an iterative process. By this way the optimal parameters are obtained for the steam generator, the vapor and the cooling gas cycle [pt

Flow-induced vibration and fretting-wear specifications to ensure steam-generator and heat exchanger lifetime performance

International Nuclear Information System (INIS)

Janzen, V.P.; Han, Y.; Pettigrew, M.J.

2008-01-01

The current interest in refurbishment, life extension and new-build activity has meant a renewed emphasis on technical specifications that will ensure improved reliability and longer life. Preventing vibration and fretting-wear problems in steam generators and heat exchangers requires design specifications that bring together specific guidelines, analysis methods, requirements and appropriate performance criteria. The specifications must be firmly based on experimental data and field inspections. In addition, the specifications must be supported by theoretical analyses and fundamental scaling correlations, to cover conditions and geometries over the wide range applicable to existing components and probable future designs. The specifications are expected to evolve to meet changing industry requirements. This paper outlines the steps required to generate and support design specifications, and relates them to typical steam-generator design features and computer modeling capabilities. It also describes current issues that are driving changes to flow-induced vibration and fretting-wear specifications that can be applied to the design process for component refurbishment, replacement or new designs. These issues include recent experimental or field evidence for new excitation mechanisms, e.g., the possibility of in-plane fluidelastic instability of U-tubes, the demand for longer reactor and component lifetimes, the need for better predictions of dynamic properties and vibration response, e.g., two-phase random-turbulence excitation, and requirements to consider system 'excursions' or abnormal scenarios, e.g., a main steam line break in the case of steam generators. The paper describes steps being taken to resolve these issues. (author)

A reflux capsule steam generator for sodium cooled reactors

International Nuclear Information System (INIS)

Lantz, E.

Pressurized water reactor plants at numerous sites have sustained significant leakage through their steam generators. The consequent shutdowns for repairs and replacements have damaged their economics. This experience suggests that if steam generators for liquid metal fast breeder reactors (LMFBR's) continue to be built as presently designed some of them will have similar problems. Because of their larger capital investment, the consequent damage to the economics of LMFBR's could be more serious. Reflux capsules provide a way to separate sodium from water and to reduce thermal stresses in steam generators for sodium cooled reactors. Their use would also eliminate the need for a primary heat exchanger and a secondary sodium loop pump. (author)

Life extension and replacement management for RAPS type steam generators

International Nuclear Information System (INIS)

Arya, R.C.; Rastogi, A.K.

1996-01-01

The steam generating equipment in first four units of Indian PHWRs Rajasthan Atomic Power Station (RAPS) 1-2 and Madras Atomic Power Station (MAPS) 1-2 are hairpin type and comprise of eight boiler assemblies. Each assembly consists of identical, single pass, inverted and vertical hairpin heat exchangers (10 for RAPS and 11 for MAPS) containing 195 monel-400 U tubes of 12.7 mm dia x 1.242 mm thick. The hot heavy water flows through these tubes and imparts heat to feed, light demineralized water entering the shell at the bottom of preheat leg. The heat is generated on the outer surface of the tubes. Details of studies carried out for life extension and replacement management for RAPS type steam generators are given. 1 fig., 5 tabs

Optimum thermal sizing and operating conditions for once through steam generator

Energy Technology Data Exchange (ETDEWEB)

Yi, Kunwoo; Ju, Kyongin; Im, Inyoung; Kim, Eunkee [KEPCO Engineering and Construction Company., Inc., Daejeon (Korea, Republic of)

2014-10-15

The steam generator is designed to be optimized so as to remove heat and to produce steam vapor. Because of its importance, theoretical and experimental researches have been performed on forced convection boiling heat transfer. The purpose of this study is to predict the thermal behavior and to perform optimum thermal sizing of once through steam generator. To estimate the tube thermal sizing and operating conditions of the steam generator, the analytical modeling is employed on the basis of the empirical correlation equations and theory. The optimized algorithm model, Non-dominated Sorting Genetic Algorithm (NSGA)-II, uses for this analysis. This research is focused on the design of in-vessel steam generator. An one dimensional analysis code is developed to evaluate previous researches and to optimize steam generator design parameters. The results of one-dimensional analysis need to be verified with experimental data. Goals of multi-objective optimization are to minimize tube length, pressure drop and tube number. Feedwater flow rate up to 115.425kg/s is selected so as to have margin of feedwater temperature 20 ..deg. C. For the design of 200MWth once through steam generator, it is evaluated that the tube length shall be over 12.0m for the number of tubes, 2500ea, and the length of the tube shall be over 8.0m for the number of tubes, 4500ea. The parallel coordinates chart can be provided to determine the optimal combination of number of tube, pressure drop, tube diameter and length.

Mitigation of steam generator tube rupture in a pressurized water reactor with passive safety systems

Science.gov (United States)

McDermott, Daniel J.; Schrader, Kenneth J.; Schulz, Terry L.

1994-01-01

The effects of steam generator tube ruptures in a pressurized water reactor are mitigated by reducing the pressure in the primary loop by diverting reactor coolant through the heat exchanger of a passive heat removal system immersed in the in containment refueling water storage tank in response to a high feed water level in the steam generator. Reactor coolant inventory is maintained by also in response to high steam generator level introducing coolant into the primary loop from core make-up tanks at the pressure in the reactor coolant system pressurizer. The high steam generator level is also used to isolate the start-up feed water system and the chemical and volume control system to prevent flooding into the steam header. 2 figures.

Steam generator for nuclear reactors

International Nuclear Information System (INIS)

Byerley, W.M.; Bennett, R.R.

1978-01-01

In the steam generator, the primary medium is led through a U-shaped tube bundle heating up a secondary medium (feedwater) which flows around the tube bundle via a preheating chamber. In order to optimize heat transfer inside the preheating chamber, the feedwater is separated into a counterflow and a parallel flow with regard to the primary medium by means of partitioning walls and deflectors. The ratio is 70/30%. This way, boiling in the preheater is avoided, i.e. the high LMTD (logaritmic mean temperature difference) is fully utilized. (DG) [de

Numerical analysis of APR1400 Steam Generator by CUPID/MARS heat structure coupling

Energy Technology Data Exchange (ETDEWEB)

Lee, Jae Ryong Lee; Lee, Seung Jun; Pakr, Ik Kyu; Yoon, Han Young [KAERI, Daejeon (Korea, Republic of); Cho, Hyoung Kyu [Seoul National University, Seoul (Korea, Republic of)

2015-05-15

To design and analyze steam generators, many computer codes have been developed and used around the world. In this study, the coupled CUPID and MARS code was used for the simulation of boiler side of the PWR steam generator. This paper presents the description of the coupling method, validation for porous media approach against the rod bundle experiment and the preliminary simulation results of PWR steam generator using the coupled code. In the present study, the multi-scale thermal-hydraulic analysis method using the coupled CUPID/MARS code was applied for the simulation of the steam generator. The primary side of the steam generator and other RCS was simulated by MARS and the secondary side was calculated by CUPID with porous media approach. For coupled simulation, the porous medium was applied in order to take into account the effect of the U-tube bundle and other supporting structure which play a role to be a flow resistance. More realistic physical model such as moisture separator slug behavior should be developed for the near future. The application of the coupled simulation should be extended to the accident scenario.

Analysis of density wave instability in counter-flow steam generators using STEAMFREQ-X

International Nuclear Information System (INIS)

Chan, K.C.; Yadigaroglu, G.

1986-01-01

The STEAMFREQ-X computer model was developed to provide a more comprehensive modeling of the different phenomena that are important to stability analysis of counter-flow steam generators. It uses a frequency-domain analysis and considers heat-flux/flow coupling between the primary and secondary fluids in space and time. Predictions by STEAMFREQ-X were compared with data from both a multi-channel liquid-sodium heated steam generator and a set of single pipe test data. Predicted outlet steam qualities at instability thresholds were within 15% of experimental data for all test points. (orig.)

Steam generator life management

International Nuclear Information System (INIS)

Tapping, R.L.; Nickerson, J.; Spekkens, P.; Maruska, C.

1998-01-01

Steam generators are a critical component of a nuclear power reactor, and can contribute significantly to station unavailability, as has been amply demonstrated in Pressurized Water Reactors (PWRs). CANDU steam generators are not immune to steam generator degradation, and the variety of CANDU steam generator designs and tube materials has led to some unexpected challenges. However, aggressive remedial actions, and careful proactive maintenance activities, have led to a decrease in steam generator-related station unavailability of Canadian CANDUs. AECL and the CANDU utilities have defined programs that will enable existing or new steam generators to operate effectively for 40 years. Research and development work covers corrosion and mechanical degradation of tube bundles and internals, chemistry, thermal hydraulics, fouling, inspection and cleaning, as well as provision for specially tool development for specific problem solving. A major driving force is development of CANDU-specific fitness-for-service guidelines, including appropriate inspection and monitoring technology to measure steam generator condition. Longer-range work focuses on development of intelligent on-line monitoring for the feedwater system and steam generator. New designs have reduced risk of corrosion and fouling, are more easily inspected and cleaned, and are less susceptible to mechanical damage. The Canadian CANDU utilities have developed programs for remedial actions to combat degradation of performance (Gentilly-2, Point Lepreau, Bruce A/B, Pickering A/B), and have developed strategic plans to ensure that good future operation is ensured. This report shows how recent advances in cleaning technology are integrated into a life management strategy, discusses downcomer flow measurement as a means of monitoring steam generator condition, and describes recent advances in hideout return as a life management tool. The research and development program, as well as operating experience, has identified

Gas-steam turbine plant for cogenerative process at 'Toplifikacija' - Skopje (Joint-Stock Co. for district heating - Macedonia)

International Nuclear Information System (INIS)

Cvetkovski, Andrijan

2003-01-01

The gas-steam power plant for combined heat and electric power production at A.D. 'Toplifikacija' Skopje - TO 'Zapad' is analyzed and determined. The analyzed plant is consisted of gas turbine, heat recovery steam generator (HRSG) and condensate steam turbine with controlled steam extraction. It operates on natural gas as a main fuel source. The heating of the water for the district heating is dine in the heat exchanger, with // heat of controlled extraction from condensate turbine. The advantages of the both binary plant and centralized co generative production compared with the individual are analyzed. The natural gas consumption of for both specific heating and electrical capacity in join production as well as fuel savings compared to the separate production of the same quantity of energy is also analyzed. (Original)

Experimental studies of PWR vertical steam generator performance

International Nuclear Information System (INIS)

Ding Xunshen

1998-06-01

The characteristics of heat transfer, natural circulation and moisture separation equipment with a vertical steam generator model are tested on a high-temperature and high-pressure test facility. The primary loop pipe is made of low alloy steel. the primary water at 10.13 MPa is driven by two centrifugal pumps. Boiler is designed to elevate the primary water temperature by 60 degree C at a flow rate of 65 t/h. The test model can produce a maximum steam flow of 8 t/h; 240 stainless steel U tubes are invert connected with the tubesheet, the tube dimensions are φ15 mm x 1.5 mm. The model has a height of 5.789 m. Heat transfer characteristics experiment is obtained: The steam generator thermal design can neglect the existence of preheating region, considering the saturated boiling takes place over all the tube surface. This does not exactly reflect the actual heat transfer coefficient and temperature difference. But, the product of two parameters is lower for design than for experiment so that heat transfer coefficient and temperature difference combine to result in a larger heat transfer area for design than for experiment. Moisture separator is a swirl vane separator. Dryer is a single circle of vertical chevron plate separator. Test results indicate the separation efficiency of the moisture separator and dryer is very satisfactory and the exit moisture is much better than 0.25% (the required content)

Integrated steam generation process and system for enhanced oil recovery

Energy Technology Data Exchange (ETDEWEB)

Betzer-Zilevitch, M. [Ex-Tar Technologies Inc., Calgary, AB (Canada)

2010-07-01

A method of producing steam for the extraction of heavy bitumens was presented. The direct contact steam generation (DCSG) method is used for the direct heat transfer between combustion gas and contaminated liquid phase water to generate steam. This paper presented details of experimental and field studies conducted to demonstrate the DCSG. Results of the study demonstrated that pressure and temperature are positively correlated. As pressure increases, the flow rate of the discharged mass decreases and the steam ratio decreases. As pressure increases, the condensate and distillate flow rates increases while water vapor losses in the non-condensable gases decrease. The study indicated that for a 10 bar pressurized system producing 9.6 mt per hour of 10,000 kpa steam and 9.6 mt per hour of distillate BFW, 70 percent of the combustion energy should be recovered to generate 10,000 kpa pressure steam for EOR. Combustion energy requirements were found to decrease when pressure decreases. 11 refs., 5 tabs., 8 figs.

Heat and electricity generating methods

International Nuclear Information System (INIS)

Buter, J.

1977-01-01

A short synopsis on the actual methods of heating of lodgings and of industrial heat generation is given. Electricity can be generated in steam cycles heated by burning of fossil fuels or by nuclear energy. A valuable contribution to the electricity economy is produced in the hydroelectric power plants. Besides these classical methods, also the different procedures of direct electricity generation are treated: thermoelectric, thermionic, magnetohydrodynamic power sources, solar and fuel cells. (orig.) [de

Completion of a simulation model for calculation of static and dynamic behaviour of steam generators and its use for a pilot steam generator

Energy Technology Data Exchange (ETDEWEB)

Seybold, M

1979-07-20

In the present paper, first of all the model concept reported in detail by Schittke is applied to the steady-state predictive calculation of an arbitrarily nested steam generator based on the design data and empirical values on heat discharge and heat transfer. The dynamics of the system are precalculated on the basis of the known steady-state behaviour and with the construction data, that is without experimental identification, according to the Schittke concept. The object for the calculation is a steam generator with once-through forced flow (VDE) which was put at the disposal of the Institute for Process Engineering and Steamboiler Studies. Data for the computer programs were taken only from the construction documents (geometries, nesting etc.) or from design data (thermic parameters). Parallel with the program preparation and the calculation measurements on the experimental plant were planned, adaptations and tests were carried out. The correctness of precalculating a steam generator could then be checked with the experimental plant and by comparison with the results of the measurements.

Thermal-Hydraulic Design of the Modular Once Through Helical Steam Generator

International Nuclear Information System (INIS)

Mazufri, C.M

2000-01-01

The steam generator system of the CAREM reactor consists of twelve individual modules located in the annular place between the pressure vessel and barrel walls. Each steam generator module consists of a tube system, an upper header, an external shroud, a collector and a lower seal.The tube system is an arrangement of several multi-start cylindrical coils.In the present work the computation of the necessary heat transfer area to fulfill the heat removal requirements from the primary circuit, and the pressure drop in the primary and secondary side of the helical design of a modular steam generator is presented. Additionally, a first order estimation of the restriction to be used in the secondary side to assure the thermal-hydraulic stability is also made.It is concluded that an array of 6 concentric cylindrical coils fulfills the necessary design requirements

Improvements to the COBRA-TF (EPRI) computer code for steam generator analysis. Final report

International Nuclear Information System (INIS)

Stewart, C.W.; Barnhart, J.S.; Koontz, A.S.

1980-09-01

The COBRA-TF (EPRI) code has been improved and extended for pressurized water reactor steam generator analysis. New features and models have been added in the areas of subcooled boiling and heat transfer, turbulence, numerics, and global steam generator modeling. The code's new capabilities are qualified against selected experimental data and demonstrated for typical global and microscale steam generator analysis

Steam generator tube failures

International Nuclear Information System (INIS)

MacDonald, P.E.; Shah, V.N.; Ward, L.W.; Ellison, P.G.

1996-04-01

A review and summary of the available information on steam generator tubing failures and the impact of these failures on plant safety is presented. The following topics are covered: pressurized water reactor (PWR), Canadian deuterium uranium (CANDU) reactor, and Russian water moderated, water cooled energy reactor (VVER) steam generator degradation, PWR steam generator tube ruptures, the thermal-hydraulic response of a PWR plant with a faulted steam generator, the risk significance of steam generator tube rupture accidents, tubing inspection requirements and fitness-for-service criteria in various countries, and defect detection reliability and sizing accuracy. A significant number of steam generator tubes are defective and are removed from service or repaired each year. This wide spread damage has been caused by many diverse degradation mechanisms, some of which are difficult to detect and predict. In addition, spontaneous tube ruptures have occurred at the rate of about one every 2 years over the last 20 years, and incipient tube ruptures (tube failures usually identified with leak detection monitors just before rupture) have been occurring at the rate of about one per year. These ruptures have caused complex plant transients which have not always been easy for the reactor operators to control. Our analysis shows that if more than 15 tubes rupture during a main steam line break, the system response could lead to core melting. Although spontaneous and induced steam generator tube ruptures are small contributors to the total core damage frequency calculated in probabilistic risk assessments, they are risk significant because the radionuclides are likely to bypass the reactor containment building. The frequency of steam generator tube ruptures can be significantly reduced through appropriate and timely inspections and repairs or removal from service

Roll-expanded plugs for steam generator heating tubes verification of leak tightness over the component lifetime

International Nuclear Information System (INIS)

Beck, J.; Ziegler, R.; Schönheit, N.

2013-01-01

Highlights: • Design description of roll-expanded plugs. • Experimental simulation of 40 years lifetime of plugged steam generator tubes. • Destructive testing for off-design loads. • Evaluation of release pressure and tightness before and after the tests. -- Abstract: Steam generator heating tubes are the boundary between the irradiated primary cycle and the conventional secondary cycle in a pressurized water reactor. Despite their operational task to transfer the heat from the primary to the secondary cycle, these tubes have a crucial safety function: the retention of irradiated primary coolant inside the circuit in all operating, emergency and off-design conditions. The heating tubes are subject to various degradation mechanisms during operation. To verify the integrity of each single tube, nuclear power plants carry out frequent in-service inspections. In case of a tube wall degradation beyond the permissible limit, the tube needs to be taken out of service in order to maintain the overall component integrity. The most common method to do so is to plug a damaged tube by a roll-expanded plug. After plugging, the roll-expanded plug acts as pressure boundary between the primary and the secondary cycle instead of the damaged heating tube. The plug must be able to maintain this function, previously provided by the heating tube, in all operational, emergency and off-design conditions. This article describes the approach to this verification by launching several comprehensive process qualification programmes consisting of mechanical analyses as well as static and dynamic testing programmes. The result was a qualified roll-expanded plug which remains leak-tight even during off-design conditions

Horizontal steam generator thermal-hydraulics

Energy Technology Data Exchange (ETDEWEB)

Ubra, O. [SKODA Praha Company, Prague (Czechoslovakia); Doubek, M. [Czech Technical Univ., Prague (Czechoslovakia)

1995-09-01

Horizontal steam generators are typical components of nuclear power plants with pressure water reactor type VVER. Thermal-hydraulic behavior of horizontal steam generators is very different from the vertical U-tube steam generator, which has been extensively studied for several years. To contribute to the understanding of the horizontal steam generator thermal-hydraulics a computer program for 3-D steady state analysis of the PGV-1000 steam generator has been developed. By means of this computer program, a detailed thermal-hydraulic and thermodynamic study of the horizontal steam generator PGV-1000 has been carried out and a set of important steam generator characteristics has been obtained. The 3-D distribution of the void fraction and 3-D level profile as functions of load and secondary side pressure have been investigated and secondary side volumes and masses as functions of load and pressure have been evaluated. Some of the interesting results of calculations are presented in the paper.

Steam generators - problems and prognosis

International Nuclear Information System (INIS)

Tapping, R.L.

1997-05-01

Steam-generator problems, largely a consequence of corrosion and fouling, have resulted in increased inspection requirements and more regulatory attention to steam-generator integrity. In addition, utilities have had to develop steam-generator life-management strategies, including cleaning and replacement, to achieve design life. This paper summarizes the pertinent data to 1993/1994, and presents an overview of current steam-generator management practices. (author)

EPRI steam generator programs

International Nuclear Information System (INIS)

Martel, L.J.; Passell, T.O.; Bryant, P.E.C.; Rentler, R.M.

1977-01-01

The paper describes the current overall EPRI steam generator program plan and some of the ongoing projects. Because of the recent occurrence of a corrosion phenomenon called ''denting,'' which has affected a number of operating utilities, an expanded program plan is being developed which addresses the broad and urgent needs required to achieve improved steam generator reliability. The goal of improved steam generator reliability will require advances in various technologies and also a management philosophy that encourages conscientious efforts to apply the improved technologies to the design, procurement, and operation of plant systems and components that affect the full life reliability of steam generators

Steam generator tube integrity requirements and operating experience in the United States

International Nuclear Information System (INIS)

Karwoski, K.J.

2009-01-01

Steam generator tube integrity is important to the safe operation of pressurized-water reactors. For ensuring tube integrity, the U.S. Nuclear Regulatory Commission uses a regulatory framework that is largely performance based. This performance-based framework is supplemented with some prescriptive requirements. The framework recognizes that there are three combinations of tube materials and heat treatments currently used in the United States and that the operating experience depends, in part, on the type of material used. This paper summarizes the regulatory framework for ensuring steam generator tube integrity, it highlights the current status of steam generators, and it highlights some of the steam generator issues and challenges that exist in the United States. (author)

Design of SMART steam generator cassette

International Nuclear Information System (INIS)

Kim, Y. W.; Kim, J. I.; Jang, M. H.

2001-01-01

Basic design development for the steam generator to be installed in the integral reactor SMART has been performed. Optimization of the steam generator shape, determination of the basic dimension and confirmation of the structural strength have been carried out. Individual steam generator cassette can be replaced in the optimized design concept of steam generator. Shape design of the steam generator cassette has been done on the computer based on 3-D CAE strategy. The structural integrity of the developed steam generator was investigated by performing the dynamic analysis for the steam generator cassette, flow induced vibration analysis for the tube bundle, and the thermo-mechanical analysis for the module header and tube. As for the manufacturing of steam generator, the numerical and the experimental simulation have been carried to control the amount of spring back and to eliminate residual stress. SMART steam generator cassette was developed by a sequential research of the aforementioned activities

Effect of a ballast zone on the hydraulic stability of a single-pass steam generator

International Nuclear Information System (INIS)

Belyakov, I.I.; Kvetnyj, M.A.; Loginov, D.A.

1985-01-01

A new mechanism of hydraulic instability of boiling channels with convection heating which reveals in the presence of a developed ballast zone at decreased loads of a counterflan steam generator operation is considered. It is shown that for the certain combinations of thermal and technical parameters pulsation regimes caused by the ballast zone displacement over the heating surface are possible. The parameter relation at which the ballast zone position becomes unstable is obtained. The effect of the ballast zone on the statis steam generator stability is established. A mechanism of whole-circuit pulsations revealed when developing start regimes of single-pass steam generator heated with liquid sodium is explained from the positions of the instability

Cold start-up condition model for heat recovery steam generators

International Nuclear Information System (INIS)

Sindareh-Esfahani, Peyman; Habibi-Siyahposh, Ehsan; Saffar-Avval, Majid; Ghaffari, Ali; Bakhtiari-Nejad, Firooz

2014-01-01

A dynamic modeling of Heat Recovery Steam Generator (HRSG) during cold start-up operation in Combined Cycle Power Plant (CCPP) is introduced. In order to characterize the essential dynamic behavior of the HRSG during cold start-up; Dynamic equations of all HRSG's components are developed based on energy and mass balances. To describe precisely the operation of HRSG; a method based on nonlinear estimated functions for thermodynamic properties is applied to estimate the model parameters. Model parameters are evaluated by a designed algorithm based on Genetic Algorithm (GA). A wide set of experimental data is used to validate HRSG model during cold start-up operation. The simulation results show the reliability and validity of the developed model for cold start-up operation. - Highlights: •Presenting a mathematical model for HRSGs cold start-up based on energy and mass balances. •A designed parameter identification algorithm based on GA is presented. •Application of experimental data in order to model and validate simulation results

Operating experience of steam generator test facility

International Nuclear Information System (INIS)

Sureshkumar, V.A.; Madhusoodhanan, G.; Noushad, I.B.; Ellappan, T.R.; Nashine, B.K.; Sylvia, J.I.; Rajan, K.K.; Kalyanasundaram, P.; Vaidyanathan, G.

2006-01-01

Steam Generator (SG) is the vital component of a Fast Reactor. It houses both water at high pressure and sodium at low pressure separated by a tube wall. Any damage to this barrier initiates sodium water reaction that could badly affect the plant availability. Steam Generator Test Facility (SGTF) has been set up in Indira Gandhi Centre for Atomic Research (IGCAR) to test sodium heated once through steam generator of 19 tubes similar to the PFBR SG dimension and operating conditions. The facility is also planned as a test bed to assess improved designs of the auxiliary equipments used in Fast Breeder Reactors (FBR). The maximum power of the facility is 5.7 MWt. This rating is arrived at based on techno economic consideration. This paper covers the performance of various equipments in the system such as Electro magnetic pumps, Centrifugal sodium pump, in-sodium hydrogen meters, immersion heaters, and instrumentation and control systems. Experience in the system operation, minor modifications, overall safety performance, and highlights of the experiments carried out etc. are also brought out. (author)

Optimization of the steam generator project of a gas cooled nuclear reactor

International Nuclear Information System (INIS)

Sakai, Massao

1978-01-01

The present work is concerned with the modeling of the primary and secondary circuits of a gas cooled nuclear reactor in order to obtain the relation between the parameters of the two cycles and the steam generator performance. The procedure allows the optimization of the steam generator, through the maximization of the plant net power, and the application of the optimal control theory of dynamic systems. The heat balances for the primary and secondary circuits are carried out simultaneously with the optimized - design parameters of the steam generator, obtained using an iterative technique. (author)

High efficiency, quasi-instantaneous steam expansion device utilizing fossil or nuclear fuel as the heat source

International Nuclear Information System (INIS)

Claudio Filippone

1999-01-01

Thermal-hydraulic analysis of a specially designed steam expansion device (heat cavity) was performed to prove the feasibility of steam expansions at elevated rates for power generation with higher efficiency. The steam expansion process inside the heat cavity greatly depends on the gap within which the steam expands and accelerates. This system can be seen as a miniaturized boiler integrated inside the expander where steam (or the proper fluid) is generated almost instantaneously prior to its expansion in the work-producing unit. Relatively cold water is pulsed inside the heat cavity, where the heat transferred causes the water to flash to steam, thereby increasing its specific volume by a large factor. The gap inside the heat cavity forms a special nozzle-shaped system in which the fluid expands rapidly, accelerating toward the system outlet. The expansion phenomenon is the cause of ever-increasing fluid speed inside the cavity system, eliminating the need for moving parts (pumps, valves, etc.). In fact, the subsequent velocity induced by the sudden fluid expansion causes turbulent conditions, forcing accelerating Reynolds and Nusselt numbers which, in turn, increase the convective heat transfer coefficient. When the combustion of fossil fuels constitutes the heat source, the heat cavity concept can be applied directly inside the stator of conventional turbines, thereby greatly increasing the overall system efficiency

Steam generator secondary side chemical cleaning at Gentilly-2

International Nuclear Information System (INIS)

Plante, S.

2006-01-01

After more than 20 years of operation, the secondary side of the four steam generators at Gentilly-2 were chemically cleaned during the 2005 annual outage. The FRAMATOME ANP high temperature cleaning process used to remove magnetite loading involved stepwise injection of solvent with PHT temperature in the range 160 o C to 175 o C. The heat required to maintain the PHT temperature was provided by the operation of the main PHT pumps and the reactor core residual heat. The temperature control was accomplished by the shutdown cooling system heat exchangers. A total of 1280 kg of magnetite was removed from the four steam generators. A copper-cleaning step was applied after the iron step. The PHT has been cooled down and the steam generators drained to temporary tanks and dried in preparation of the copper step. The process has been applied at room temperature, two boilers at a time. The solvent removed a total of 116 kg of copper. During the iron step, steam flow to the feedwater tank chemically contaminate the Balance Of Plant (BOP) systems. The isolation of this path should have been part of the G2 procedures. Around 700 m3 of water had to be drained to interim storage tanks for subsequent resin treatment before disposal. Visual inspection of BO1 tubesheet and first support plate showed clean surfaces without measurable sludge pile. Upper support plates visual inspection of BO4 revealed that broach holes blockage reported in 2000 is still present in peripheral area. Following the plant restart, the medium range level measurement instability observed since several years for BO3 was no more present. As anticipated, it also has been observed that the medium and wide range level measurements have shifted down as a result of downcomer flow increase after the cleaning. The cleaning objectives were achieved regarding the fouling reduction on the steam generators secondary side but broach holes blockage of the upper support plate is still present in periphery. (author)

Fast reactor fuel failures and steam generator leaks: Transient and accident analysis approaches

International Nuclear Information System (INIS)

1996-10-01

This report consists of a survey of activities on transient and accident analysis for the LMFR. It is focused on the following subjects: Fuel transient tests and analyses in hypothetical incident/accident situations; sodium-water interaction in steam generators, and sodium fires: test and analyses. There are also sections dealing with the experimental and analytical studies of: fuel subassembly failures; sodium boiling, molten fuel-coolant interaction; molten material movement and relocation in fuel bundles; heat removal after an accident or incident; sodium-water reaction in steam generator; steam generator protection systems; sodium-water contact in steam generator building; fire-fighting methods and systems to deal with sodium fires. Refs, figs, tabs

Equipment for secondary water distribution in heat exchanger, especially saturated steam generator for nuclear power plants and heat plants

Energy Technology Data Exchange (ETDEWEB)

Riman, J; Manek, O; Rybnicek, J

1979-09-15

A special structure consisting of a system of channels and a distribution plate with ports in-built above the tube-plate of a vertical-type steam generator prevents secondary water vaporization in the space above the tube-plate and thus also salt and sludge sedimentation which causes increased corrosion of heat transfer tubes. The size of the distribution plate ports is variable in the radial direction. The distribution plate is divided by means of the system of channels into at least two parts. The middle section of each part is of the through-flow type.

Calculation of wastage by small water leaks in sodium heated steam generators

International Nuclear Information System (INIS)

Tregonning, K.

1976-01-01

On the basis of mechanistic arguments it is suggested that the temperature of the wasting surface would provide a single physically meaningful parameter with which to correlate wastage data. A lumped parameter model is developed which predicts reaction temperature as a function of the major variables in the small water leak situation (Leak rate, tube spacing, sodium temperature). The calculated temperatures explain much of the observed behaviour of wastage rate with these variables and compare well with the limited temperature data available. Wastage rates are correlated with predicted temperature on a total activation energy basis. The results are encouraging and a first conservative method for the calculation of wastage by small water leaks in sodium-heated steam generators is produced

Evaluation of EDTA based chemical formulations for the cleaning of monel-400 tubed steam generators

International Nuclear Information System (INIS)

Velmurugan, S.; Rufus, A.L.; Sathyaseelan, V.S.; Kumar, P.S.; Veena, S.N.; Srinivasan, M.P.; Narasimhan, S.V.

1998-01-01

The Steam Generator (SG) is an important component in any nuclear power plant which contributes significantly for the over all performance of the reactor. The failure of SG tubes occurs mainly by corrosion under accelerated conditions caused by fouling. There is continuous ingress of the corrosion products and ionic impurities from the condenser and feed train of the secondary heat transfer system. The corrosion products accumulate in the stagnant areas near the tube sheet, over the tube support plates and in the tube to tube support plate crevices. These accumulated deposits help to concentrate the aggressive impurities and induce a variety of corrosion processes affecting the structural materials and finally leading to failure of the SG tube. Scale forming impurities can deposit over the tube surfaces and result in reduction of heat transfer efficiency and over heating of the surfaces. Every effort is being made to control the transport of impurities to the steam generator. Increased blow down, installation of condensate polishers and use of all volatile amines have helped to reduce the corrosion product and ionic impurities input into the steam generators of PHWRs. Despite these efforts, failures of SG tubes in PHWRs have been reported. Hence, attempts are being made to develop chemical formulations to clean the deposits accumulated in the steam generators. The EPRI-SGOG chemical cleaning process has been tried with good success in steam generators of different designs including the steam generators of PHWRs. This paper discusses the work on the evaluation of EDTA based chemical cleaning formulations for monel-400 tubed steam generators of PHWRs. (author)

Set-up for steam generator tube bundle washing after explosion expanding the tubes

International Nuclear Information System (INIS)

Osipov, S.I.; Kal'nin, A.Ya.; Mazanenko, M.F.

1985-01-01

Set-up for steam generator tube bundle washing after the explosion expanding of tubes is described. Washing is accomplished by distillate. Steam is added to distillate for heating, and compersed air for preventing hydraulic shock. The set-up is equiped by control equipment. Set-up performances are presented. Time for one steam generator washing constitutes 8-12 h. High economic efficiency is realized due to the set-up introduction

Numerical methods on flow instabilities in steam generator

International Nuclear Information System (INIS)

Yoshikawa, Ryuji; Hamada, Hirotsugu; Ohshima, Hiroyuki; Yanagisawa, Hideki

2008-06-01

The phenomenon of two-phase flow instability is important for the design and operation of many industrial systems and equipment, such as steam generators. The designer's job is to predict the threshold of flow instability in order to design around it or compensate for it. So it is essential to understand the physical phenomena governing such instability and to develop computational tools to model the dynamics of boiling systems. In Japan Atomic Energy Agency, investigations on heat transfer characteristics of steam generator are being performed for the development of Sodium-cooled Fast Breeder Reactor. As one part of the research work, the evaluations of two-phase flow instability in the steam generator are being carried out experimentally and numerically. In this report, the numerical methods were studied for two-phase flow instability analysis in steam generator. For numerical simulation purpose, the special algorithm to calculate inlet flow rate iteratively with inlet pressure and outlet pressure as boundary conditions for the density-wave instability analysis was established. There was no need to solve property derivatives and large matrices, so the spurious numerical instabilities caused by discontinuous property derivatives at boiling boundaries were avoided. Large time-step was possible. The flow instability in single heat transfer tube was successfully simulated with homogeneous equilibrium model by using the present algorithm. Then the drift-flux model including the effects of subcooled boiling and two phase slip was adopted to improve the accuracy. The computer code was developed after selecting the correlations of drift velocity and distribution parameter. The capability of drift flux model together with the present algorithm for simulating density-wave instability in single tube was confirmed. (author)

WWER steam generator tube structural and leakage integrity

International Nuclear Information System (INIS)

Splichal, K.; Krhounek, Vl.; Otruba, J.; Ruscak, M.

1998-01-01

The integrity of heat exchange tubes may influence the lifetime of WWER steam generators and appears to be an important criterion for the evaluation of their safety and operational reliability. The basic requirements are to assure very low probability of radioactive water leakage, preventing unstable crack growth and sudden tube rupture. These requirements led to development of permissible limits for primary to secondary leak evaluation and heat exchange tubes plugging. The stress corrosion cracking and pitting are the main corrosion damages of WWER heat exchange tubes and are initiated from the outer surface. Both the initiation and crack growth cause thinning of the tube wall and lead to part thickness cracks and through wall cracks, oriented preferentially in the axial direction. The paper presents the leakage and plugging limits for WWER steam generators, which have been determined from leak tests and burst tests. The tubes with axial part-through and through-wall defects have been used. The permissible value of primary to secondary leak rate was evaluated with respect to permissible axial through-wall defect size of WWER 440 and 1000 steam generator tubes. Blocking of the tube cracks by corrosion product particles and other compounds reduces the primary to secondary leak rate. The plugging limits involve the following factors: permissible tube wall thickness which determine further operation of the tubes with defects and assures their integrity under operating conditions and permissible size of a through-wall crack which is sufficiently stable under normal and accident conditions in relation to the critical crack length. For the evaluation of burst test of heat exchange tubes with longitudinal through-wall defects the instability criterion has been used and the dependence of the normalised burst pressure on the normalised length of an axial through-wall defect has been determined. The validity of the criterion of instability for WWER tubes with through

Hybrid preheat/recirculating steam generator

International Nuclear Information System (INIS)

Lilly, G.P.

1985-01-01

The patent describes a hybrid preheat/recirculating steam generator for nuclear power plants. The steam generator utilizes recirculated liquid to preheat incoming liquid. In addition, the steam generator incorporates a divider so as to limit the amount of recirculating water mixed with the feedwater. (U.K.)

On possibility of application of the parallel-mixed type coolant flow scheme to NPP steam generators linked with superheaters

International Nuclear Information System (INIS)

Malkis, V.A.; Lokshin, V.A.

1983-01-01

Optimum distribution of the coolant straight-through flow between the superheater, evaporator and economizer is determined and the parallel-mixed type flow scheme is compared with other schemes. The calculations are performed for the 250 MW(e) steam generator for the WWER-1000 reactor unit the inlet and outlet primary coolant temperature of which is 324 and 290 deg C, respectively, while the feed water and saturation temperatures are 220 and 278.5 deg C, respectively. The rated superheating temperature is 300 deg C. The comparison of different schemes has been performed according to the average temperature head value at the steam-generator under the condition of equality as well as essential difference in the heat transfer coefficients in certain steam-generator sections. The calculations have shown that the use of parallel-mixed type flow permits to essentially increase the temperature head of the steam generator. At a constant heat transfer coefficient in all steam generator sections the highest temperature head is reached. At relative flow rates in the steam generator, economizer and evaporator equal to 6, 8 and 86%, respectively. The superheated steam generator temperature head in this case by 12% exceeds the temperature head of the WWER-1000 reactor unit wet steam generator. In case of heat transfer coefficient reduction in the superheater by a factor of three, the choice of the primary coolant, optimum distribution permits to maintain the steam generator temperature head at the level of the WWER-1000 reactor unit wet-steam steam generator. The use of the parallel-mixed type flow scheme permits to design a steam generator of slightly superheated steam for the parameters of the WWER-1000 unit

Heysham II/Torness AGR steam generator

International Nuclear Information System (INIS)

Charcharos, A.N.; Wood, M.B.; Glasgow, J.R.

1988-01-01

The AGR Steam Generators for Heysham II and Torness Power Stations have been installed at site and are being operated in the initial low temperature commissioning plant engineering tests. In this paper a description of the high pressure once-through steam generators together with layout arrangements, materials employed, operating parameters, plant operating conditions and constraints is given. An outline of the development of the design through thermo-hydraulic considerations, mechanical design, instrumentation to component testing is presented. Special features of the design directed to accommodate such requirements as seismic loadings, waterside static and dynamic stability, gas flow induced vibration, thermal expansions are described in detail. The fabrication facilities employed and techniques selected and developed for the manufacture and assembly of the heating surfaces are presented. These include welding processes, tube manipulation and heat treatment with details of the automation applied to the processes. Operating experience in the early commissioning plant engineering tests at Site is described with an emphasis on those tests which provide the final confirmation of the design prior to operation at full load. The paper concludes with a description of the outstanding commissioning activities up to raise power. (author)

Heysham II/Torness AGR steam generator

Energy Technology Data Exchange (ETDEWEB)

Charcharos, A N [National Nuclear Corporation Ltd., Knutsford (United Kingdom); Wood, M B; Glasgow, J R [NEI Power Projects Ltd., Gateshead (United Kingdom)

1988-07-01

The AGR Steam Generators for Heysham II and Torness Power Stations have been installed at site and are being operated in the initial low temperature commissioning plant engineering tests. In this paper a description of the high pressure once-through steam generators together with layout arrangements, materials employed, operating parameters, plant operating conditions and constraints is given. An outline of the development of the design through thermo-hydraulic considerations, mechanical design, instrumentation to component testing is presented. Special features of the design directed to accommodate such requirements as seismic loadings, waterside static and dynamic stability, gas flow induced vibration, thermal expansions are described in detail. The fabrication facilities employed and techniques selected and developed for the manufacture and assembly of the heating surfaces are presented. These include welding processes, tube manipulation and heat treatment with details of the automation applied to the processes. Operating experience in the early commissioning plant engineering tests at Site is described with an emphasis on those tests which provide the final confirmation of the design prior to operation at full load. The paper concludes with a description of the outstanding commissioning activities up to raise power. (author)

Disposal of Steam Generators from Decommissioning of PWR Nuclear Power Plants

International Nuclear Information System (INIS)

Walberg, Mirko; Viermann, Joerg; Beverungen, Martin; Kemp, Lutz; Lindstroem, Anders

2008-01-01

Amongst other materials remarkable amounts of radioactively contaminated or activated scrap are generated from the dismantling of Nuclear Power Plants. These scrap materials include contaminated pipework, fittings, pumps, the reactor pressure vessel and other large components, most of them are heat exchangers. Taking into account all commercial and technical aspects an external processing and subsequent recycling of the material might be an advantageous option for many of these components. The disposal of steam generators makes up an especially challenging task because of their measures, their weight and compared to other heat exchangers high radioactive inventory. Based on its experiences from many years of disposal of smaller components of NPP still in operation or under decommissioning GNS and Studsvik Nuclear developed a concept for disposal of steam generators, also involving experiences made in Sweden. The concept comprises transport preparations and necessary supporting documents, the complete logistics chain, steam generator treatment and the processing of arising residues and materials not suitable for recycling. The first components to be prepared, shipped and treated according to this concept were four steam generators from the decommissioning of the German NPP Stade which were removed from the plant and shipped to the processing facility during the third quarter of 2007. Although the plant had undergone a full system decontamination, due to the remaining contamination in a number of plugged tubes the steam generators had to be qualified as industrial packages, type 2 (IP-2 packages), and according to a special requirement of the German Federal Office for Radiation Protection a license for a shipment under special arrangement had to be applied for. The presentation gives an overview of the calculations and evidences required within the course of the IP-2 qualification, additional requirements of the competent authorities during the licensing procedure as

Forced circulation type steam generator simulation code: HT4

International Nuclear Information System (INIS)

Okamoto, Masaharu; Tadokoro, Yoshihiro

1982-08-01

The purpose of this code is a understanding of dynamic characteristics of the steam generator, which is a component of High-temperature Heat Transfer Components Test Unit. This unit is a number 4th test section of Helium Engineering Demonstration Loop (HENDEL). Features of this report are as follows, modeling of the steam generator, a basic relationship for the continuity equation, numerical analysis techniques of a non-linear simultaneous equation and computer graphics output techniques. Forced circulation type steam generator with strait tubes and horizontal cut baffles, applied in this code, have be designed at the Over All System Design of the VHTRex. The code is for use with JAERI's digital computer FACOM M200. About 1.5 sec required for each time step reiteration, then about 40 sec cpu time required for a standard problem. (author)

Qualification of Alloy 800 for sodium heated steam generators

International Nuclear Information System (INIS)

Duke, J.M.; Sessions, C.E.; Ray, W.E.

1976-01-01

A reference specification of Alloy 800 for use in LMFBR steam generators is defined considering waterside corrosion, weldability and mechanical properties. Additional mechanical test data are being generated to support ASME Code acceptance. Candidate weld filler metals were assessed for use in subsequent weld process development. Progress on resolving technical concerns related to the role of tertiary creep identified the impact of test conditions and gamma prime strengthening in determining the creep behavior of Alloy 800

Steam generator blowdown heat exchangers degradations operational experience on EDF French NPP fleet

International Nuclear Information System (INIS)

Praud, M.; Doyen, F.; Wintergest, M.; Jourdain, W.; Roussillon, M.; Zidane, A.; Mayos, M.

2015-01-01

The main function of the Steam Generator Blowdown System (SGBS) is to purify the secondary fluid from all kinds of pollutions: corrosion products from the secondary system, consequences of raw water pollutions through condenser's leakage, potential radiochemical pollutions resulting from Primary-to-Secondary leaks. The topic of this paper is to present the main SGBS dysfunctions linked to the degradation of the tubular heat exchangers, which sometimes can lead to integrity failure, through corrosion phenomenon. The degradation mechanisms have been characterized by various visual inspections and destructive examinations performed on pulled tubes and bundles. It appears that SGBS tubes suffer two main forms of corrosion. First, for the non-regenerative heat exchangers, where external surface of tubes is exposed to intermediate fluid, alkaline corrosion under tube sheet or shell-side baffles may occur. Caustic attack results from Na 3 PO 4 decomposition by thermal or chemical process. Secondly, mainly for regenerative heat exchangers, pitting and under-deposits corrosion linked to lay-up conditions during outages. This kind of attack is the root cause of a potential 'domino effect': a steam jet from the leaking tube can induce mechanical and/or erosion on many tubes located in its vicinity. Concerning the external degradation by caustic corrosion, only design modifications and strong monitoring of the raw water inlet may able to limit the occurrence of tube perforation. The lay-up guidelines should be carefully followed to mitigate internal corrosion: a controlled atmosphere (limited humidity) and cleanliness of the tube (avoiding deposits formation on the bottom line) seem to be the main parameters

STEAM GENERATOR TUBE INTEGRITY ANALYSIS OF A TOTAL LOSS OF ALL HEAT SINKS ACCIDENT FOR WOLSONG NPP UNIT 1

Directory of Open Access Journals (Sweden)

HEOK-SOON LIM

2014-02-01

Full Text Available A total loss of all heat sinks is considered a severe accident with a low probability of occurrence. Following a total loss of all heat sinks, the degasser/condenser relief valves (DCRV become the sole means available for the depressurization of the primary heat transport system. If a nuclear power plant has a total loss of heat sinks accident, high-temperature steam and differential pressure between the primary heat transport system (PHTS and the steam generator (SG secondary side can cause a SG tube creep rupture. To protect the PHTS during a total loss of all heat sinks accident, a sufficient depressurization capability of the degasser/condenser relief valve and the SG tube integrity is very important. Therefore, an accurate estimation of the discharge through these valves is necessary to assess the impact of the PHTS overprotection and the SG tube integrity of the primary circuit. This paper describes the analysis of DCRV discharge capacity and the SG tube integrity under a total loss of all heat sink using the CATHENA code. It was found that the DCRV's discharge capacity is enough to protect the overpressure in the PHTS, and the SG tube integrity is maintained in a total loss of all heat accident.

Steam Generator Tube Integrity Analysis of A Total Loss of all Heat Sinks Accident for Wolsong NPP Unit 1

Energy Technology Data Exchange (ETDEWEB)

Lim, Heoksoon; Song, Taeyoung; Chi, Moongoo [Korea Htydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of); Kim, Seoungrae [Nuclear Engineering Service and Solution, Daejeon (Korea, Republic of)

2014-02-15

A total loss of all heat sinks is considered a severe accident with a low probability of occurrence. Following a total loss of all heat sinks, the degasser/condenser relief valves (DCRV) become the sole means available for the depressurization of the primary heat transport system. If a nuclear power plant has a total loss of heat sinks accident, high-temperature steam and differential pressure between the primary heat transport system (PHTS) and the steam generator (SG) secondary side can cause a SG tube creep rupture. To protect the PHTS during a total loss of all heat sinks accident, a sufficient depressurization capability of the degasser/condenser relief valve and the SG tube integrity is very important. Therefore, an accurate estimation of the discharge through these valves is necessary to assess the impact of the PHTS overprotection and the SG tube integrity of the primary circuit. This paper describes the analysis of DCRV discharge capacity and the SG tube integrity under a total loss of all heat sink using the CATHENA code. It was found that the DCRV's discharge capacity is enough to protect the overpressure in the PHTS, and the SG tube integrity is maintained in a total loss of all heat accident.

Steam Generator Tube Integrity Analysis of A Total Loss of all Heat Sinks Accident for Wolsong NPP Unit 1

International Nuclear Information System (INIS)

Lim, Heoksoon; Song, Taeyoung; Chi, Moongoo; Kim, Seoungrae

2014-01-01

A total loss of all heat sinks is considered a severe accident with a low probability of occurrence. Following a total loss of all heat sinks, the degasser/condenser relief valves (DCRV) become the sole means available for the depressurization of the primary heat transport system. If a nuclear power plant has a total loss of heat sinks accident, high-temperature steam and differential pressure between the primary heat transport system (PHTS) and the steam generator (SG) secondary side can cause a SG tube creep rupture. To protect the PHTS during a total loss of all heat sinks accident, a sufficient depressurization capability of the degasser/condenser relief valve and the SG tube integrity is very important. Therefore, an accurate estimation of the discharge through these valves is necessary to assess the impact of the PHTS overprotection and the SG tube integrity of the primary circuit. This paper describes the analysis of DCRV discharge capacity and the SG tube integrity under a total loss of all heat sink using the CATHENA code. It was found that the DCRV's discharge capacity is enough to protect the overpressure in the PHTS, and the SG tube integrity is maintained in a total loss of all heat accident

Numerical Study on the Helium Flow Characteristics for Steam Generator Subsystem of HTR

International Nuclear Information System (INIS)

Ha, Jung Hoon; Ham, Jin Ki; Ki, Min-Hwan; Lee, Won Jae

2014-01-01

The High Temperature Reactor (HTR), one of the 4th generation reactors, utilizes helium as the primary coolant. A Steam Generator Subsystem (SGS) is installed to transfer heat from the primary coolant to feed water and subsequently produce steam so that it supplies electricity as well as process heat over a wide range. The SGS is composed of a helical heat exchanger, shrouds directing the flow of the shell side helium and support systems, which are located within the steam generator vessel. In this study, helium flow characteristics in the SGS were investigated at various operating conditions using Computational Fluid Dynamics (CFD). A full-scale 3-D model of the SGS was developed and the reynolds stress model with standard wall treatment was used as a turbulence model. The CFD result was compared to that of the concept design of the steam cycle modular helium reactor for the design verification of the SGS. From the CFD analysis, it was found that the primary coolant flow had non-uniform distribution while it passed the inlet in the helical heat exchanger. In order to make the uniform primary coolant flow uniform, a special type of screen was suggested in front of the helical heat exchanger. As a result, the overall design adequacy of the SGS has been evaluated. (author)

CANDU steam generator life management

International Nuclear Information System (INIS)

Tapping, R.L.; Nickerson, J.; Spekkens, P.; Maruska, C.

1998-01-01

Steam generators are a critical component of a nuclear power reactor, and can contribute significantly to station unavailability, as has been amply demonstrated in Pressurized Water Reactors (PWRs). CANDU steam generators are not immune to steam generator degradation, and the variety of CANDU steam generator designs and tube materials has led to some unexpected challenges. However, aggressive remedial actions, and careful proactive maintenance activities, have led to a decrease in steam generator-related station unavailability of Canadian CANDUs. AECL and the CANDU utilities have defined programs that will enable existing or new steam generators to operate effectively for 40 years. Research and development work covers corrosion and mechanical degradation of tube bundles and internals, chemistry, thermalhydraulics, fouling, inspection and cleaning, as well as provision for specially tool development for specific problem solving. A major driving force is development of CANDU-specific fitness-for-service guidelines, including appropriate inspection and monitoring technology to measure steam generator condition. Longer-range work focuses on development of intelligent on-line monitoring for the feedwater system and steam generator. New designs have reduced risk of corrosion and fouling, are more easily inspected and cleaned, and are less susceptible to mechanical damage. The Canadian CANDU utilities have developed programs for remedial actions to combat degradation of performance (Gentilly-2, Point Lepreau, Bruce A/B, Pickering A/B), and have developed strategic plans to ensure that good future operation is ensured. The research and development program, as well as operating experience, has identified where improvements in operating practices and/or designs can be made in order to ensure steam generator design life at an acceptable capacity factory. (author)

Heat Recovery From Tail Gas Incineration To Generate Power

Energy Technology Data Exchange (ETDEWEB)

Tawfik, Tarek

2010-09-15

Many industrial processes result in tail gas wastes that must be flared or incinerated to abide with environmental guidelines. Tail gas incineration occurs in several chemical processes resulting in high-temperature exhaust gas that simply go to the stack, thus wasting all that valuable heat! This paper discusses useful heat recovery and electric power generation utilizing available heat in exhaust gas from tail gas incinerators. This heat will be recovered in a waste-heat recovery boiler that will produce superheated steam to expand in a steam turbine to generate power. A detailed cost estimate is presented.

Development of Technologies on Innovative-Simplified Nuclear Power Plant Using High-Efficiency Steam Injectors (12) Evaluations of Spatial Distributions of Flow and Heat Transfer in Steam Injector

International Nuclear Information System (INIS)

Yutaka Abe; Yujiro Kawamoto; Chikako Iwaki; Tadashi Narabayashi; Michitsugu Mori; Shuichi Ohmori

2006-01-01

Next-generation nuclear reactor systems have been under development aiming at simplified system and improvement of safety and credibility. One of the innovative technologies is the supersonic steam injector, which has been investigated as one of the most important component of the next-generation nuclear reactor. The steam injector has functions of a passive pump without large motor or turbo-machinery and a high efficiency heat exchanger. The performances of the supersonic steam injector as a pump and a heat exchanger are dependent on direct contact condensation phenomena between a supersonic steam and a sub-cooled water jet. In previous studies of the steam injector, there are studies about the operating characteristics of steam injector and about the direct contact condensation between static water pool and steam in atmosphere. However, there is a little study about the turbulent heat transfer and flow behavior under the great shear stress. In order to examine the heat transfer and flow behavior in supersonic steam injector, it is necessary to measure the spatial temperature distribution and velocity in detail. The present study, visible transparent supersonic steam injector is used to obtain the axial pressure distributions in the supersonic steam injector, as well as high speed visual observation of water jet and steam interface. The experiments are conducted with and without non-condensable gas. The experimental results of the interfacial flow behavior between steam and water jet are obtained. It is experimentally clarified that an entrainment exists on the water jet surface. It is also clarified that discharge pressure is depended on the steam supply pressure, the inlet water flow rate, the throat diameter and non-condensable flow rate. Finally a heat flux is estimated about 19 MW/m 2 without non-condensable gas condition in steam. (authors)

Pulsed high-pressure (PHP) drain-down of steam generating system

International Nuclear Information System (INIS)

Petrusek, R.A.

1991-01-01

This patent describes an improved method of draining down contained reactor-coolant water from the inverted vertical U-tubes of at least one vertical-type steam generator in which the upper inverted U-shaped ends of the tubes are closed and the lower ends thereof are open, the steam generator having a channel head at its lower end including a vertical dividing wall defining a primary water inlet side and a primary water outlet side of the generator, the steam generator having chemical volume control system means and residual heat removal system means, and the steam generator being part of a nuclear-powered steam generating system wherein the reactor-coolant water is normally circulated from and back into the reactor via a loop comprising the steam generator and inlet and outlet conduits connected to the lower end of the steam generator, and the reactor being in communication with pressurizer means and comprising the steps of introducing a gas which is inert to the system and which is under pressure above atmospheric pressure into at least one of the downwardly facing open ends of each of the U-tubes from below the tubesheet in which the open ends of the U-tubes are mounted adjacent the lower end of the steam generator while permitting the water to flow out from the open ends of the U-tubes, the improvement in combination therewith for substantially increasing the effectiveness and efficiency of such water removal from the tubes. It includes determining the parameters effecting a first average volumetric rate of removal for a predetermined period of time, infra, of the reactor-coolant water from the inverted vertical U-tubes, the specific unit for the first average volumetric rate expressing properties identical with the properties expressed in a second average volumetric rate maintained in a later mentioned step

Steam generator for use in nuclear power plants

International Nuclear Information System (INIS)

Cella, A.

1980-01-01

An improved steam generator is described for use in a nuclear power plant of the pressurized water type in which a turbine generator is driven by the steam output of the steam generator to provide electrical power therefrom. The improvement comprises providing a vertically movable grid structure vertically extending within the interior of the lower housing portion of the steam generator through which individual tubes comprising a vertically extending tube bundle extend. The tube bundle has a tube sheet at one end thereof supporting the tube bundle for the tubes extending through the tube sheet in flow through communication with a heat exchange fluid inlet. The grid structure defines grid apertures therein through which the individual tubes extend with each of the grid apertures being in surrounding relationship with a portion of an associated one of the tubes. The grid structure is movable for a predetermined vertical extent, such as by hydraulic means, such as a piston, along the tubes for vertically displacing the means defining the grid apertures by a sufficient amount for removing the previously surrounded portion of each of the tubes from the associated grid apertures whereby an enhanced reading of the condition of the tubes at the previously surrounded portion is enabled. The steam generator may comprise vertically assemblable modules which are removably mounted together in sealing relationship, with the modules comprising a base module, a tube bundle module removably mountable on the base module in sealing relationship therewith and an uppermost drier module removably mountable on the tube bundle module in sealing relationship therewith whereby ready access to removal of the tube bundle module in situ from the nuclear power plant steam generator is facilitated

Dew point measurements of flue gases in steam generators with brown coal combustion

Energy Technology Data Exchange (ETDEWEB)

Schinkel, W.

1980-01-01

This paper examines empirical data on sulfuric acid condensation and resulting internal corrosion in brown coal fired steam generators. Due to the high sulfur content in brown coal (0.5% to 5.0%) and relative short duration of the gases in the combustion chamber the concentrations of sulfur trioxide present in the flue gases can condense at the heat exchange surfaces of the steam generators. A number of diagrams show sulfuric acid dew point temperatures depending on brown coal sulfur content, the influence of combustion air supply on the dew point, and condensing speed and the rate of corrosion in relation to different heat exchange surface temperatures. The conclusion is made that a five-fold increase in corrosion can be caused by a 10 K higher flue gas dew point, a 5 K cooling of heating surfaces can also cause heavy corrosion at a certain dew point. Maximum corrosion results at 20 to 50 K differences between flue gas dew point and heat exchange surfaces. Optimum operation of steam generators with minimal internal corrosion requires the consideration of flue gas and heating surface temperatures as well as flue gas sulfur acid dew points. (10 refs.) (In German)

Hideout in steam generator tube deposits

International Nuclear Information System (INIS)

Balakrishnan, P.V.; Franklin, K.J.; Turner, C.W.

1998-05-01

Hideout in deposits on steam generator tubes was studied using tubes coated with magnetite. Hideout from sodium chloride solutions at 279 degrees C was followed using an on-line high-temperature conductivity probe, as well as by chemical analysis of solution samples from the autoclave in which the studies were done. Significant hideout was observed only at a heat flux greater than 200 kW/m 2 , corresponding to a temperature drop greater than 2 degrees C across the deposits. The concentration factor resulting from the hideout increased highly non-linearly with the heat flux (varying as high as the fourth power of the heat flux). The decrease in the apparent concentration factor with increasing deposit thickness suggested that the pores in the deposit were occupied by a mixture of steam and water, which is consistent with the conclusion from the thermal conductivity measurements on deposits in a separate study. Analyses of the deposits after the hideout tests showed no evidence of any hidden-out solute species, probably due to the concentrations being very near the detection limits and to their escape from the deposit as the tests were being ended. This study showed that hideout in deposits may concentrate solutes in the steam generator bulk water by a factor as high as 2 x 10 3 . Corrosion was evident under the deposit in some tests, with some chromium enrichment on the surface of the tube. Chromium enrichment usually indicates an acidic environment, but the mobility required of chromium to become incorporated into the thick magnetite deposit may indicate corrosion under an alkaline environment. An alkaline environment could result from preferential accumulation of sodium in the solution in the deposit during the hideout process. (author)

Evolution of design of steam generator for sodium cooled reactors

International Nuclear Information System (INIS)

Chetal, S.C.; Vaidyanathan

1997-01-01

The first sodium cooled reactor was the experimental breeder reactor (EBR-I) in usa which was commissioned in 1951 and was incidentally the first nuclear reactor to generate electrical energy. This was followed by fast breeder reactors in USSR, UK, france, USA, japan, germany and India. The use of sodium as a coolant is due to its low moderation which helps in breeding fissile fuel from fertile materials and also its high heat transfer coefficient at comparatively low velocities. The good heat transfer properties introduce thermal stresses when there are rapid changes in the sodium temperatures. Also sodium has a chemical affinity with air and water. The steam generators for sodium cooled reactors have to allow for these novel conditions and in addition, unlike other components. Choices have to be made whether it is a recirculation type as in most fossil plants or an once through unit, the power rating, shape of the tube (straight, helical, U-tube), materials (Ferritic or austenitic), with free level of sodium or not, sodium on tube side or shell side and so on. With higher pressures and steam temperatures reheating steam after partial expansion in the turbine becomes essential as in conventional turbines. For this purpose the choice of reheating fluid viz sodium or live main steam has to be made. This paper traces the evolution of steam generator designs in the different sodium cooled reactors (chronologically) and the operation experience. 16 figs., 1 tab

Cogeneration steam turbine plant for district heating of Berovo (Macedonia)

International Nuclear Information System (INIS)

Armenski, Slave; Dimitrov, Konstantin

2000-01-01

A plant for combined heat and electric power production, for central heating of the town Berovo (Macedonia) is proposed. The common reason to use a co-generation unit is the energy efficiency and a significant reduction of environmental pollution. A coal dust fraction from B rik' - Berovo coal mine is the main energy resource for cogeneration steam turbine plant. The heat consumption of town Berovo is analyzed and determined. Based on the energy consumption of a whole power plant, e. i. the plant for combined and simultaneous production of power is proposed. All necessary facilities of cogeneration plant is examined and determined. For proposed cogeneration steam turbine power plant for combined heat and electric production it is determined: heat and electric capacity of the plant, annually heat and electrical quantity production and annually coal consumption, the total investment of the plant, the price of both heat and electric energy as well as the pay back period. (Authors)

Corrosion allowances for sodium heated steam generators: evaluation of effects and extrapolation to component life time

Energy Technology Data Exchange (ETDEWEB)

Grosser, E E; Menken, G

1975-07-01

Steam generator tubes are subjected to two categories of corrosion; metal/sodium reactions and metal/water-steam interactions. Referring to these environmental conditions the relevant parameters are discussed. The influences of these parameters on the sodium corrosion and water/steam-reactions are evaluated. Extrapolations of corrosion values to steam generator design conditions are performed and discussed in detail. (author)

Corrosion allowances for sodium heated steam generators: evaluation of effects and extrapolation to component life time

International Nuclear Information System (INIS)

Grosser, E.E.; Menken, G.

1975-01-01

Steam generator tubes are subjected to two categories of corrosion; metal/sodium reactions and metal/water-steam interactions. Referring to these environmental conditions the relevant parameters are discussed. The influences of these parameters on the sodium corrosion and water/steam-reactions are evaluated. Extrapolations of corrosion values to steam generator design conditions are performed and discussed in detail. (author)

Tube tightness survey during Phenix steam generator operation

International Nuclear Information System (INIS)

Cambillard, E.

1976-01-01

Phenix steam generators are once-through vessels with single-wall heat-exchange tubes. This design means that any leakage of water into the sodium must be detected as quickly as possible so that the installation can be shut down before extensive damage occurs. The detection of water leaks in Phenix steam generators is based on measurement of the concentration in the sodium, of hydrogen produced by the sodium-water reaction. Since the various modules--evaporators, superheaters, and reheaters--have no free sodium surfaces, detection of hydrogen in argon is not used in Phenix steam generators. The measurement systems employ a probe made of nickel tubes 0.3 mm thick. Hydrogen in the sodium diffuses into a chamber kept under vacuum by an ion pump. The hydrogen pressure in the chamber is measured by a quadrupole mass spectrometer. The nine measurement systems (three per steam generator) are calibrated by injecting hydrogen into the sodium of the secondary circuits. The data-processing computer calculates the hydrogen concentration in the sodium from the spectrometer signals and the probe temperatures, which are not regulated in Phenix; it generates instructions that enable the operator to act if a leak appears. So far, no leaks have been detected. These systems also make it possible to determine rates of hydrogen diffusion caused by corrosion of the steel walls on the water side

Direct measurements of secondary water inventory of steam generator PGV-213 in operation

Energy Technology Data Exchange (ETDEWEB)

Tarankov, G.A.; Trunov, N.B.; Dranchenko, B.N.; Kamiagin, W.W. [OKB Gidropress (Russian Federation)

1997-12-31

Results of weight measurement of PGV-213 steam generator during filling in, heating-up and power increase are described. Special measurement system based on stress gauges has been developed. Method of derivation of secondary water inventory is described. Comparison of the data for two steam generators prove accuracy of the measurements. (orig.). 1 refs.

Direct measurements of secondary water inventory of steam generator PGV-213 in operation

Energy Technology Data Exchange (ETDEWEB)

Tarankov, G A; Trunov, N B; Dranchenko, B N; Kamiagin, W W [OKB Gidropress (Russian Federation)

1998-12-31

Results of weight measurement of PGV-213 steam generator during filling in, heating-up and power increase are described. Special measurement system based on stress gauges has been developed. Method of derivation of secondary water inventory is described. Comparison of the data for two steam generators prove accuracy of the measurements. (orig.). 1 refs.

Steam generator replacement in Bruce A Unit 1 and Unit 2

International Nuclear Information System (INIS)

Hart, R.S.

2007-01-01

The Bruce A Generating Station consists of four 900 MW class CANDU units. The reactor and Primary Heat Transport System for each Unit are housed within a reinforced concrete reactor vault. A large duct running below the reactor vaults accommodates the shared fuel handling system, and connects the four reactor vaults to the vacuum building. The reactor vaults, fuelling system duct and the vacuum building constitute the station vacuum containment system. Bruce A Unit 2 was shut down in 1995 and Bruce A Units 1, 3 and 4 were shutdown in 1997. Bruce A Units 3 and 4 were returned to service in late 2003 and are currently operating. Units 1 and 2 remain out of service. Bruce Power is currently undertaking a major rehabilitation of Bruce A Unit 1 and Unit 2 that will extend the in-service tile of these units by at least 25 years. Replacement of the Steam Generators (eight in each unit) is required; this work was awarded to SNC-Lavalin Nuclear (SLN). The existing steam drums (which house the steam separation and drying equipment) will be retained. Unit 2 is scheduled to be synchronized with the grid in 2009, followed by Unit 1 in 2009. Each Bruce A unit has two steam generating assemblies, one located above and to each end of the reactor. Each steam generating assembly consists of a horizontal cylindrical steam drum and four vertical Steam Generators. The vertical Steam Generators connect to individual nozzles that are located on the underside of the Steam Drum (SD). The steam drums are located in concrete shielding structures (steam drum enclosures). The lower sections of the Steam Generators penetrate the top of the reactor vaults: the containment pressure boundary is established by bellows assemblies that connect between the reactor vault roof slab and the Steam Generators. Each Steam Generators is supported from he bottom by a trapeze that is suspended from the reactor vault top structure. The Steam Generator Replacement (SGR) methodology developed by SLN for Unit 1

Steam injection for heavy oil recovery: Modeling of wellbore heat efficiency and analysis of steam injection performance

International Nuclear Information System (INIS)

Gu, Hao; Cheng, Linsong; Huang, Shijun; Li, Bokai; Shen, Fei; Fang, Wenchao; Hu, Changhao

2015-01-01

Highlights: • A comprehensive mathematical model was established to estimate wellbore heat efficiency of steam injection wells. • A simplified approach of predicting steam pressure in wellbores was proposed. • High wellhead injection rate and wellhead steam quality can improve wellbore heat efficiency. • High wellbore heat efficiency does not necessarily mean good performance of heavy oil recovery. • Using excellent insulation materials is a good way to save water and fuels. - Abstract: The aims of this work are to present a comprehensive mathematical model for estimating wellbore heat efficiency and to analyze performance of steam injection for heavy oil recovery. In this paper, we firstly introduce steam injection process briefly. Secondly, a simplified approach of predicting steam pressure in wellbores is presented and a complete expression for steam quality is derived. More importantly, both direct and indirect methods are adopted to determine the wellbore heat efficiency. Then, the mathematical model is solved using an iterative technique. After the model is validated with measured field data, we study the effects of wellhead injection rate and wellhead steam quality on steam injection performance reflected in wellbores. Next, taking cyclic steam stimulation as an example, we analyze steam injection performance reflected in reservoirs with numerical reservoir simulation method. Finally, the significant role of improving wellbore heat efficiency in saving water and fuels is discussed in detail. The results indicate that we can improve the wellbore heat efficiency by enhancing wellhead injection rate or steam quality. However, high wellbore heat efficiency does not necessarily mean satisfactory steam injection performance reflected in reservoirs or good performance of heavy oil recovery. Moreover, the paper shows that using excellent insulation materials is a good way to save water and fuels due to enhancement of wellbore heat efficiency

Steam generator replacement in Bruce A Unit 1 and Unit 2

International Nuclear Information System (INIS)

Hart, R.S.

2006-01-01

The Bruce A Generating Station consists of four 900 MW class CANDU units. The reactor and Primary Heat Transport System for each Unit are housed within a reinforced concrete reactor vault. A large duct running below the reactor vaults accommodates the shared fuel handling system, and connects the four reactor vaults to the vacuum building. The reactor vaults, fuelling system duct and the vacuum building constitute the station vacuum containment system. Bruce A Unit 2 was shut down in 1995 and Bruce A Units 1, 3 and 4 were shutdown in 1997. Bruce A Units 3 and 4 were returned to service in late 2003 and are currently operating. Units 1 and 2 remain out of service. Bruce Power is currently undertaking a major rehabilitation of Bruce A Unit 1 and Units 2 that will extend the in-service life of these units by at least 25 years. Replacement of the Steam Generators (eight in each unit) is required; this work was awarded to SNC-Lavalin Nuclear (SLN). The existing steam drums (which house the steam separation and drying equipment) will be retained. Unit 2 is scheduled to be synchronized with the grid in 2009, followed by Unit 1 in 2009. Each Bruce A unit has two steam generating assemblies, one located above and to each end of the reactor. Each steam generating assembly consists of a horizontal cylindrical steam drum and four vertical Steam Generators. The vertical Steam Generators connect to individual nozzles that are located on the underside of the Steam Drum (SD). The steam drums are located in concrete shielding structures (steam drum enclosures). The lower sections of the Steam Generators penetrate the top of the reactor vaults: the containment pressure boundary is established by bellows assemblies that connect between the reactor vault roof slab and the Steam Generators. Each Steam Generators is supported from the bottom by a trapeze that is suspended from the reactor vault top structure. The Steam Generator Replacement (SGR) methodology developed by SLN for Unit 1

Steam generator water lancing

International Nuclear Information System (INIS)

Kamler, F.; Schneider, W.

1992-01-01

The tubesheet and tube support plate deposits in CANDU steam generators are notable for their hardness. Also notable is the wide variety of steam generator access situations. Because of the sludge hardness and the difficulty of the access, traditional water lancing processes which directed jets from the central tube free lane or from the periphery of the bundle have proven unsuitable. This has led to the need for some very unique inter tube water lancing devices which could direct powerful water jets directly onto the deposits. This type of process was applied to the upper broached plates of the Bruce A steam generators, which had become severely blocked. It has since been applied to various other steam generator situations. This paper describes the flexlance equipment development, qualification, and performance in the various CANDU applications. 4 refs., 2 tabs., 7 figs

French steam generator design developments

International Nuclear Information System (INIS)

Ginier, R.; Campan, J.L.; Pontier, M.; Leridon, A.; Remond, A.; Castello, G.; Holcblat, A.; Paurobally, H.

1986-01-01

From the outset of the French nuclear power program, a significant R and D effort has been invested in improvement of the design and operation of Pressurized Water Reactors including a special committment to improving steam generators. The steam generator enhancement program has spawned a wide variety of specific R and D resources, e.g., low temperature hydraulic models for investigation of areas with single-phase flow, and freon-filled models for simulation of areas of steam generators experiencing two-phase flow (tube bundles and moisture separators). For the moisture separators, a large scale research program using freon-filled models and highly sophisticated instrumentation was used. Tests at reactor sites during startup of both 900 MWe and 1300 MWe have been used to validate the assumptions made on the basis of loop tests. These tests also demonstrated the validity of using freon to simulate two-phase flow conditions. The wealth of knowledge accumulated by the steam generator R and D program has been used to develop a new design of steam generators for the N4 plants. The current R and D effort is aimed at qualifying the N4 steam generator model and developing more comprehensive models. One prong of the R and D effort is the Megeve program. Megeve is a 25 MW steam generator which simulates operating conditions of the N4 model. The other prong is Clotaire, a freon-filled steam generator model which will be used to qualify thermal/hydraulic design codes used for multidimensional calculations for design of tube bundles

Potential ability of zeolite to generate high-temperature vapor using waste heat

Science.gov (United States)

Fukai, Jun; Wijayanta, Agung Tri

2018-02-01

In various material product industries, a large amount of high temperature steam as heat sources are produced from fossil fuel, then thermal energy retained by condensed water at lower than 100°C are wasted. Thermal energies retained by exhaust gases at lower than 200°C are also wasted. Effective utilization of waste heat is believed to be one of important issues to solve global problems of energy and environment. Zeolite/water adsorption systems are introduced to recover such low-temperature waste heats in this study. Firstly, an adsorption steam recovery system was developed to generate high temperature steam from unused hot waste heat. The system used a new principle that adsorption heat of zeolite/water contact was efficiently extracted. A bench-scaled system was constructed, demonstrating contentious generation of saturated steam nearly 150°C from hot water at 80°C. Energy conservation is expected by returning the generated steam to steam lines in the product processes. Secondly, it was demonstrated that superheated steam/vapor at higher than 200°C could be generated from those at nearly 120°C using a laboratory-scaled setup. The maximum temperature and the time variation of output temperature were successfully estimated using macroscopic heat balances. Lastly, the maximum temperatures were estimated whose saturate air at the relative humidity 20-80% were heated by the present system. Theoretically, air at higher than 200°C was generated from saturate air at higher than 70°C. Consequently, zeolite/water adsorption systems have potential ability to regenerate thermal energy of waste water and exhaust gases.

Study group meeting on steam generators for LMFBR's. Summary report

Energy Technology Data Exchange (ETDEWEB)

NONE

1975-07-01

The Meeting organised by IAEA international working group on fast reactors which considered that the subject of sodium heated steam generators was a topic which needed study by the experts of several disciplines. For example: people who design such steam generators, specialists in the field of sodium water reactions, experts in material and water chemistry and members of the utilities who would be the customers for such units. Besides the exchange of large amount of information, it was considered that further special studies were necessary for the following subjects: materials; maintenance and repair; operating procedures and control of steam generators. A separate study of sodium-water reactions was recommended considering the safety aspects related to large water leakage and economic advantage of possible detection and protection against small water leaks.

Study group meeting on steam generators for LMFBR's. Summary report

International Nuclear Information System (INIS)

1975-07-01

The Meeting organised by IAEA international working group on fast reactors which considered that the subject of sodium heated steam generators was a topic which needed study by the experts of several disciplines. For example: people who design such steam generators, specialists in the field of sodium water reactions, experts in material and water chemistry and members of the utilities who would be the customers for such units. Besides the exchange of large amount of information, it was considered that further special studies were necessary for the following subjects: materials; maintenance and repair; operating procedures and control of steam generators. A separate study of sodium-water reactions was recommended considering the safety aspects related to large water leakage and economic advantage of possible detection and protection against small water leaks

Steam generator tube integrity program

International Nuclear Information System (INIS)

Dierks, D.R.; Shack, W.J.; Muscara, J.

1996-01-01

A new research program on steam generator tubing degradation is being sponsored by the U.S. Nuclear Regulatory Commission (NRC) at Argonne National Laboratory. This program is intended to support a performance-based steam generator tube integrity rule. Critical areas addressed by the program include evaluation of the processes used for the in-service inspection of steam generator tubes and recommendations for improving the reliability and accuracy of inspections; validation and improvement of correlations for evaluating integrity and leakage of degraded steam generator tubes, and validation and improvement of correlations and models for predicting degradation in steam generator tubes as aging occurs. The studies will focus on mill-annealed Alloy 600 tubing, however, tests will also be performed on replacement materials such as thermally-treated Alloy 600 or 690. An overview of the technical work planned for the program is given

Steam generator tube extraction

International Nuclear Information System (INIS)

Delorme, H.

1985-05-01

To enable tube examination on steam generators in service, Framatome has now developed a process for removing sections of steam generator tubes. Tube sections can be removed without being damaged for treating the tube section expanded in the tube sheet

Cheaper power generation from surplus steam generating capacities

International Nuclear Information System (INIS)

Gupta, K.

1996-01-01

Prior to independence most industries had their own captive power generation. Steam was generated in own medium/low pressure boilers and passed through extraction condensing turbines for power generation. Extraction steam was used for process. With cheaper power made available in Nehru era by undertaking large hydro power schemes, captive power generation in industries was almost abandoned except in sugar and large paper factories, which were high consumers of steam. (author)

Development of a Program for Predicting Flow Instability in a Once-through Sodium-Heated Steam Generator (III)

Energy Technology Data Exchange (ETDEWEB)

Kim, Eui Kwang; Yoon, Jung; Kim, Jong Bum; Jeong, Jiyoung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Two-phase flow systems can be subjected to several types of instability problems. Density-wave oscillation is the most common and important type of instability in boiling channels. Such instability gives difficulties in predictions of system performance and system control, and component failure due to thermal fatigue. A computer program developed for predicting two-phase flow instability in a steam generator heated by liquid sodium was presented in the previous works. Limit cycle was predicted even in a fixed node system. The amplitude of inlet flow rate is larger than that of outlet flow rate. The amplitude of phase change location oscillation within boiling-to-vapor boundary node is larger than that of liquid-to-boiling boundary node. Sodium and steam temperature are invariant at tube exit.

Development of a nuclear steam generator system for gas-cooled reactors for application in oil sands extraction

International Nuclear Information System (INIS)

Smith, J.; Hart, R.; Lazic, L.

2009-01-01

Canada has vast energy reserves in the Oil Sands regions of Alberta and Saskatchewan. Present extraction technologies, such as strip mining, where oil deposits are close to the surface, and Steam Assisted Gravity Drainage (SAGD) technologies for deeper deposits consume significant amounts of energy to produce the bitumen and upgraded synthetic crude oil. Studies have been performed to assess the feasibility of using nuclear reactors as primary energy sources to produce, in particular the steam required for the SAGD deeper deposit extraction process. Presently available reactors fall short of meeting the requirements, in two areas: the steam produced in a 'standard' reactor is too low in pressure and temperature for the SAGD process. Requirements can be for steam as high as 12MPa pressure with superheat; and, 'standard' reactors are too large in total output. Ideally, reactors of output in the range of 400 to 500 MWth, in modules are better suited to Oil Sands applications. The above two requirements can be met using gas-cooled reactors. Generally, newer generation gas-cooled reactors have been designed for power generation, using Brayton Cycle gas turbines run directly from the heated reactor coolant (helium). Where secondary steam is required, heat recovery steam generators have been used. In this paper, a steam generating system is described which uses the high temperature helium from the reactor directly for steam generation purposes, with sufficient quantities of steam produced to allow for SAGD steam injection, power generation using a steam turbine-generator, and with potential secondary energy supply for other purposes such as hydrogen production for upgrading, and environmental remediation processes. It is assumed that the reactors will be in one central location, run by a utility type organization, providing process steam and electricity to surrounding Oil Sands projects, so steam produced is at very high pressure (12 MPa), with superheat, in order to

Decontamination of Steam Generator tube using Abrasive Blasting Technology

International Nuclear Information System (INIS)

Min, B. Y.; Kim, G. N.; Choi, W. K.; Lee, K. W.; Kim, D. H.; Kim, K. H.; Kim, B. T.

2010-01-01

As a part of a technology development of volume reduction and self disposal for large metal waste project, We at KAERI and our Sunkwang Atomic Energy Safety (KAES) subcontractor colleagues are demonstrating radioactively contaminated steam generator tube by abrasive blasting technology at Kori-1 NPP. A steam generator is a crucial component in a PWR (pressurized Water Reactor). It is the crossing between the primary, contaminated, circuit and the secondary waste-steam circuit. The heat from the primary reactor coolant loop is transferred to the secondary side in thousands of small tubes. Due to several problems in the material of those tube, like SCC (Stress Corrosion Cracking), insufficient control in water chemistry, which can be cause of tube leakage, more and more steam generators are replaced today. Only in Korea, already 2 of them are replaced and will be replaced in the near future. The retired 300 ton heavy Steam generator was stored at the storage waste building of Kori NPP site. The steam generator waste has a large volume, so that it is necessary to reduce its volume by decontamination. A waste reduction effect can be obtained through decontamination of the inner surface of a steam generator. Therefore, it is necessary to develop an optimum method for decontamination of the inner surface of bundle tubes. The dry abrasive blasting is a very interesting technology for the realization of three-dimensional microstructures in brittle materials like glass or silicon. Dry abrasive blasting is applicable to most surface materials except those that might be shattered by the abrasive. It is most effective on flat surface and because the abrasive is sprayed and can also applicable on 'hard to reach' areas such as inner tube ceilings or behind equipment. Abrasive decontamination techniques have been applied in several countries, including Belgium, the CIS, France, Germany, Japan, the UK and the USA

Commercially Available Activated Carbon Fiber Felt Enables Efficient Solar Steam Generation.

Science.gov (United States)

Li, Haoran; He, Yurong; Hu, Yanwei; Wang, Xinzhi

2018-03-21

Sun-driven steam generation is now possible and has the potential to help meet future energy needs. Current technologies often use solar condensers to increase solar irradiance. More recently, a technology for solar steam generation that uses heated surface water and low optical concentration is reported. In this work, a commercially available activated carbon fiber felt is used to generate steam efficiently under one sun illumination. The evaporation rate and solar conversion efficiency reach 1.22 kg m -2 h -1 and 79.4%, respectively. The local temperature of the evaporator with a floating activated carbon fiber felt reaches 48 °C. Apart from the high absorptivity (about 94%) of the material, the evaporation performance is enhanced thanks to the well-developed pores for improved water supply and steam escape and the low thermal conductivity, which enables reduced bulk water temperature increase. This study helps to find a promising material for solar steam generation using a water evaporator that can be produced economically (∼6 $/m 2 ) with long-term stability.

Optimizing the Heat Exchanger Network of a Steam Reforming System

DEFF Research Database (Denmark)

Nielsen, Mads Pagh; Korsgaard, Anders Risum; Kær, Søren Knudsen

2004-01-01

Proton Exchange Membrane (PEM) based combined heat and power production systems are highly integrated energy systems. They may include a hydrogen production system and fuel cell stacks along with post combustion units optionally coupled with gas turbines. The considered system is based on a natural...... stationary numerical system model was used and process integration techniques for optimizing the heat exchanger network for the reforming unit are proposed. Objective is to minimize the system cost. Keywords: Fuel cells; Steam Reforming; Heat Exchanger Network (HEN) Synthesis; MINLP....... gas steam reformer along with gas purification reactors to generate clean hydrogen suited for a PEM stack. The temperatures in the various reactors in the fuel processing system vary from around 1000°C to the stack temperature at 80°C. Furthermore, external heating must be supplied to the endothermic...

Apparatus for inspecting and repairing a pressurized-water reactor's steam generator heat exchanger tubes

International Nuclear Information System (INIS)

Mueller, O.; Roettger, H.; Kasti, H.; Hagen, H.G.

1976-01-01

Described is an apparatus provided for use with a pressurized-water reactor' steam generator having a manifold chamber enclosing the bottom side of a horizontal tube sheet having holes therethrough in which are mounted the tubes of a heat exchanger tube bundle. The manifold chamber has a manhole giving access to the tube's bottom side to permit internal inspection or repair of the tubes by registration of an end of a flexible guide conduit with the tube sheet holes and through which a flexible carrier can be guided for insertion via these holes in the tube sheet and through the tubes extending from the tube sheet's other side

Expandable antivibration bar for heat transfer tubes of a pressurized water reactor steam generator

International Nuclear Information System (INIS)

Appleman, R.H.

1985-01-01

An expandable antivibration bar for use in stabilizing the U-bend portion of heat transfer tubes in a pressurized water reactor steam generator comprises two adjustable rods connected together by an arcuate connector. The two adjustable rods preferably comprise two mating rod sections having complementary angular sliding surfaces thereon, with means provided to move the rod sections relative to each other along the sliding surfaces so as to expand the rods from a first mated cross-sectional width to a second larger cross-sectional width. The ends of the rod sections have means for aligning the two rod sections and maintaining them in alignment during expansion. (author)

Development of Design Criteria for Fluid Induced Structural Vibrations in Steam Generators and Heat Exchangers

International Nuclear Information System (INIS)

Catton, Ivan; Dhir, Vijay K.; Alquaddoomi, O.S.; Mitra, Deepanjan; Adinolfi, Pierangelo

2004-01-01

OAK-B135 Flow-induced vibration in heat exchangers has been a major cause of concern in the nuclear industry for several decades. Many incidents of failure of heat exchangers due to apparent flow-induced vibration have been reported through the USNRC incident reporting system. Almost all heat exchangers have to deal with this problem during their operation. The phenomenon has been studied since the 1970s and the database of experimental studies on flow-induced vibration is constantly updated with new findings and improved design criteria for heat exchangers. In the nuclear industry, steam generators are often affected by this problem. However, flow-induced vibration is not limited to nuclear power plants, but to any type of heat exchanger used in many industrial applications such as chemical processing, refrigeration and air conditioning. Specifically, shell and tube type heat exchangers experience flow-induced vibration due to the high velocity flow over the tube banks. Flow-induced vibration in these heat exchangers leads to equipment breakdown and hence expensive repair and process shutdown. The goal of this research is to provide accurate measurements that can help modelers to validate their models using the measured experimental parameters and thereby develop better design criteria for avoiding fluid-elastic instability in heat exchangers. The research is divided between two primary experimental efforts, the first conducted using water alone (single phase) and the second using a mixture of air or steam and water as the working fluid (two phase). The outline of this report is as follows: After the introduction to fluid-elastic instability, the experimental apparatus constructed to conduct the experiments is described in Chapter 2 along with the measurement procedures. Chapter 3 presents results obtained on the tube array and the flow loop, as well as techniques used in data processing. The project performance is described and evaluated in Chapter 4 followed by

Thermal circuit and supercritical steam generator of the BGR-300 nuclear power plant

International Nuclear Information System (INIS)

Afanas'ev, B.P.; Godik, I.B.; Komarov, N.F.; Kurochnkin, Yu.P.

1979-01-01

Secondary coolant circuit and a steam generator for supercritical steam parameters of the BGR-300 reactor plant are described. The BGR-300 plant with a 300 MW(e) high-temperature gas-cooled fast reactor is developed as a pilot commercial plant. It is shown that the use of a supercritical pressure steam increases the thermal efficiency of the plant and descreases thermal releases to the environment, permits to use home-made commercial turbine plants of large unit power. The proposed supercritical pressure steam generator has considerable advantages from the viewpoint of heat transfer and hydrodynamical processes

Evaluation of steam generator U-tube integrity during PWR station blackout with secondary system depressurization

International Nuclear Information System (INIS)

Hidaka, Akihide; Asaka, Hideaki; Sugimoto, Jun; Ueno, Shingo; Yoshino, Takehito

1999-12-01

In PWR severe accidents such as station blackout, the integrity of steam generator U-tube would be threatened early at the transient among the pipes of primary system. This is due to the hot leg countercurrent natural circulation (CCNC) flow which delivers the decay heat of the core to the structures of primary system if the core temperature increases after the secondary system depressurization. From a view point of accident mitigation, this steam generator tube rupture (SGTR) is not preferable because it results in the direct release of primary coolant including fission products (FP) to the environment. Recent SCDAP/RELAP5 analyses by USNRC showed that the creep failure of pressurizer surge line which results in release of the coolant into containment would occur earlier than SGTR during the secondary system depressurization. However, the analyses did not consider the decay heat from deposited FP on the steam generator U-tube surface. In order to investigate the effect of decay heat on the steam generator U-tube integrity, the hot leg CCNC flow model used in the USNRC's calculation was, at first, validated through the analysis for JAERI's LSTF experiment. The CCNC model reproduced well the thermohydraulics observed in the LSTF experiment and thus the model is mostly reliable. An analytical study was then performed with SCDAP/RELAP5 for TMLB' sequence of Surry plant with and without secondary system depressurization. The decay heat from deposited FP was calculated by JAERI's FP aerosol behavior analysis code, ART. The ART analysis showed that relatively large amount of FPs may deposit on steam generator U-tube inlet mainly by thermophoresis. The SCDAP/RELAP5 analyses considering the FP decay heat predicted small safety margin for steam generator U-tube integrity during secondary system depressurization. Considering associated uncertainties in the analyses, the potential for SGTR cannot be ignored. Accordingly, this should be considered in the evaluation of merits

Steam generator with perfected dryers

International Nuclear Information System (INIS)

Fenet, J.C.

1987-01-01

This steam generator has vertically superposed array of steam dryers. These dryers return the steam flow of 180 0 . The return of the water is made by draining channels to the steam production zone [fr

Optimum design of dual pressure heat recovery steam generator using non-dimensional parameters based on thermodynamic and thermoeconomic approaches

International Nuclear Information System (INIS)

Naemi, Sanaz; Saffar-Avval, Majid; Behboodi Kalhori, Sahand; Mansoori, Zohreh

2013-01-01

The thermodynamic and thermoeconomic analyses are investigated to achieve the optimum operating parameters of a dual pressure heat recovery steam generator (HRSG), coupled with a heavy duty gas turbine. In this regard, the thermodynamic objective function including the exergy waste and the exergy destruction, is defined in such a way to find the optimum pinch point, and consequently to minimize the objective function by using non-dimensional operating parameters. The results indicated that, the optimum pinch point from thermodynamic viewpoint is 2.5 °C and 2.1 °C for HRSGs with live steam at 75 bar and 90 bar respectively. Since thermodynamic analysis is not able to consider economic factors, another objective function including annualized installation cost and annual cost of irreversibilities is proposed. To find the irreversibility cost, electricity price and also fuel price are considered independently. The optimum pinch point from thermoeconomic viewpoint on basis of electricity price is 20.6 °C (75 bar) and 19.2 °C (90 bar), whereas according to the fuel price it is 25.4 °C and 23.7 °C. Finally, an extensive sensitivity analysis is performed to compare optimum pinch point for different electricity and fuel prices. -- Highlights: ► Presenting thermodynamic and thermoeconomic optimization of a heat recovery steam generator. ► Defining an objective function consists of exergy waste and exergy destruction. ► Defining an objective function including capital cost and cost of irreversibilities. ► Obtaining the optimized operating parameters of a dual pressure heat recovery boiler. ► Computing the optimum pinch point using non-dimensional operating parameters

Thermal-Hydraulic Analysis of a Once-Through Steam Generator Considering Performance Degradation

International Nuclear Information System (INIS)

Han, Hun Sik; Kang, Han Ok; Yoon, Ju Hyeon; Kim, Young In; Song, Jae Seung; Kim, Keung Koo

2016-01-01

Several countries have entered into a global race for the commercialization of SMRs, and considerable research and development have been implemented. Among the various reactor designs, many SMRs have adopted an integral type pressurized water reactor (PWR) to enhance the nuclear safety and system reliability. In the integral reactor design, a single reactor pressure vessel contains primary system components such as fuel and core, steam generators, pumps, and a pressurizer. For the component integration into a reactor vessel, it is important to design each component as small as possible. Thus, it is a common practice to employ a once-through steam generator in the integral reactor design due to its advantages in compactness. In general, gradual degradation in thermal-hydraulic performance of the steam generator occurs with time, and it changes slowly the operating point of the steam generator during plant lifetime. Numerical solutions are acquired to evaluate the thermal-hydraulic performance of the steam generator at various AUFs. The design results obtained show that the average tube length of the steam generator is augmented with the increase of design margin to compensate for the design uncertainties and heat transfer area reduction by plugging, fouling, etc. A helically coiled tube once-through steam generator with 30% design margin is considered for comparison of thermal-hydraulic performances according to the degradation rate

Thermal-Hydraulic Analysis of a Once-Through Steam Generator Considering Performance Degradation

Energy Technology Data Exchange (ETDEWEB)

Han, Hun Sik; Kang, Han Ok; Yoon, Ju Hyeon; Kim, Young In; Song, Jae Seung; Kim, Keung Koo [KAERI, Daejeon (Korea, Republic of)

2016-05-15

Several countries have entered into a global race for the commercialization of SMRs, and considerable research and development have been implemented. Among the various reactor designs, many SMRs have adopted an integral type pressurized water reactor (PWR) to enhance the nuclear safety and system reliability. In the integral reactor design, a single reactor pressure vessel contains primary system components such as fuel and core, steam generators, pumps, and a pressurizer. For the component integration into a reactor vessel, it is important to design each component as small as possible. Thus, it is a common practice to employ a once-through steam generator in the integral reactor design due to its advantages in compactness. In general, gradual degradation in thermal-hydraulic performance of the steam generator occurs with time, and it changes slowly the operating point of the steam generator during plant lifetime. Numerical solutions are acquired to evaluate the thermal-hydraulic performance of the steam generator at various AUFs. The design results obtained show that the average tube length of the steam generator is augmented with the increase of design margin to compensate for the design uncertainties and heat transfer area reduction by plugging, fouling, etc. A helically coiled tube once-through steam generator with 30% design margin is considered for comparison of thermal-hydraulic performances according to the degradation rate.

Studies and solutions of steam turbines for nuclear heating power stations

International Nuclear Information System (INIS)

Drahy, J.

1979-01-01

The possibilities of combined generation of heat and electric power and special features of the corresponding equipment for WWER type reactors are considered. Condensing steam turbines with bled steam points and the constructional solution of bled points are presented for heating the network water to 110 0 C, 120 0 C, and 160 0 C, respectively. The dimensions of the low pressure final stage of the turbine are given. Problems concerning condensing and bleeding turbines and combination types of back-pressure and condensing turbines as well as solutions to the design of 250 MW and 500 MW turbines are discussed

Steam generator reliability improvement project

International Nuclear Information System (INIS)

Blomgren, J.C.; Green, S.J.

1987-01-01

Upon successful completion of its research and development technology transfer program, the Electric Power Research Institute's Steam Generator Owners Group (SGOG II) will disband in December 1986 and be replaced in January 1987 by a successor project, the Steam Generator Reliability Project (SGRP). The new project, funded in the EPRI base program, will continue the emphasis on reliability and life extension that was carried forward by SGOG II. The objectives of SGOG II have been met. Causes and remedies have been identified for tubing corrosion problems, such as stress corrosion cracking and pitting, and steam generator technology has been improved in areas such as tube wear prediction and nondestructive evaluation (NDE). These actions have led to improved reliability of steam generators. Now the owners want to continue with a centrally managed program that builds on what has been learned. The goal is to continue to improve steam generator reliability and solve small problems before they become large problems

Steam generator reliability improvement project

International Nuclear Information System (INIS)

Blomgren, J.C.; Green, S.J.

1987-01-01

Upon successful completion of its research and development technology transfer program, the Electric Power Research Institute's (EPRI's) Steam Generator Owners Group (SGOG II) will disband in December 1986, and be replaced in January 1987, by a successor project, the Steam Generator Reliability Project (SGRP). The new project, funded in the EPRI base program, will continue to emphasize reliability and life extension, which were carried forward by SGOG II. The objectives of SGOG II have been met. Causes and remedies have been identified for tubing corrosion problems such as stress corrosion cracking and pitting, and steam generator technology has been improved in areas such as tube wear prediction and nondestructive evaluation. These actions have led to improved reliability of steam generators. Now the owners want to continue with a centrally managed program that builds on what has been learned. The goal is to continue to improve steam generator reliability and to solve small problems before they become large problems

Materials engineering issues, LMFBR steam generators

International Nuclear Information System (INIS)

Spalaris, C.N.; Challenger, K.D.; Day, R.A.; Dutina, D.; Ring, P.J.

1976-01-01

Selection of 2-1/4 Cr-1 Mo as the reference construction material for LMFBR steam generators assumed a balance between its known intrinsic properties and our ability to accommodate certain of its deficiencies through design allowance. A comprehensive development program was undertaken to define base data needed, confirm assumptions made relative to desired performance, minimize defects by optimization of melting, fabrication and heat treatment processes, and prepare specifications for purchasing reactor components

Device indicating start of steam or water reaction with sodium and damage of steam generator heat exchange tube wall

International Nuclear Information System (INIS)

Jung, J.; Sobotka, J.

1984-01-01

Eddy currents induced by the alternating current of an exciting coil in the vicinity of steam or water leakage are used for indication. The coil is supplied from a power amplifier whose input is connected to an exciting generator by two measuring coils connected across each other. Their voltage is applied to a differential amplifier with an indicator. The equipment may be used for steam generators of nuclear power plants with sodium cooled reactors. (E.F.)

49 CFR 229.105 - Steam generator number.

Science.gov (United States)

2010-10-01

... 49 Transportation 4 2010-10-01 2010-10-01 false Steam generator number. 229.105 Section 229.105..., DEPARTMENT OF TRANSPORTATION RAILROAD LOCOMOTIVE SAFETY STANDARDS Safety Requirements Steam Generators § 229.105 Steam generator number. An identification number shall be marked on the steam generator's...

Small leak detection by measuring surface oscillation during sodium-water reaction in steam generator

International Nuclear Information System (INIS)

Nei, Hiromichi; Hori, Masao

1977-01-01

Small leak sodium-water reaction tests were conducted to develop various kinds of leak detectors for the sodium-heated steam generator in FBR. The super-heated steam was injected into sodium in a reaction vessel having a sodium free surface, simulating the steam generator. The level gauge in the reaction vessel generated the most reliable signal among detectors, as long as the leak rates were relatively high. The level gauge signal was estimated to be the sodium surface oscillation caused by hydrogen bubbles produced in sodium-water reaction. Experimental correlation was derived, predicting the amplitude as a function of leak rate, hydrogen dissolution ratio, bubble rise velocity and other parameters concerned, assuming that the surface oscillation is in proportion to the gas hold-up. The noise amplitude under normal operation without water leak was increased with sodium flow rate and found to be well correlated with Froud number. These two correlations predict that a water leak in a ''MONJU'' class (300 MWe) steam generator could possibly be detected by level gauges at a leak rate above 2 g/sec. (auth.)

Forced convection heat transfer of steam in a square ribbed channel

Energy Technology Data Exchange (ETDEWEB)

Liu, Jiazeng; Gao, Jianmin; Gao, Tieyu [Xi' an Jiaotong University, Shaanxi (China)

2012-04-15

An experimental study of heat transfer characteristics of steam in a square channel (simulating a gas turbine blade cooling passage) with two opposite surfaces roughened by 60 deg parallel ribs was performed. The ranges of key governing parameters were: Reynolds numbers (Re) based on the channel hydraulic diameter (30000-140000), entry gauge pressure (0.2Mpa-0.5Mpa), heat flux of heat transfer surface area (5kWm{sup -2}-20kWm{sup -2}), and steam superheat (13 .deg. C-51 .deg. C). The test channel length was 1000mm, while the rib spacing (p/e) was 10, and the ratio of rib height (e) to hydraulic diameter (D) was 0.048. The test channel was heated by passing current through stainless steel walls instrumented with thermocouples. The local heat transfer coefficients on the ribbed wall from the channel entrance to the fully developed regions were measured. The semi-empirical correlation was fitted out by using the average Nusselt numbers in the fully developed region to cover the range of Reynolds number. The correlation can be used in the design of new generation of gas turbine blade cooled by steam.

Power generation from a 7700C heat source by means of a main steam cycle, a topping closed gas cycle and a ammonia bottoming cycle

International Nuclear Information System (INIS)

Tilliette, Z.P.

1981-03-01

For power generation, steam cycles make an efficient use of medium temperature heat sources. They can be adapted to dry cooling, higher power ratings and output increase in winter by addition of an ammonia bottoming cycle. Active development is carried out in this field by 'Electricite de France'. As far as heat sources at higher temperatures are concerned, particularly related to coal-fired or nuclear power plants, a more efficient way of converting energy is at first to expand a hot working fluid through a gas turbine. It is shown in this paper that a satisfactory result, for heat sources of about 770 0 C, is obtained with a topping closed gas cycle of moderate power rating, rejecting its waste heat into the main steam cycle. Attention has to be paid to this gas cycle waste heat recovery and to the coupling of the gas and steam cycles. This concept drastically reduces the importance of new technology components. The use and the significance of an ammonia bottoming cycle in this case are investigated

WWER Steam Generators Tubing Performance and Aging Management

International Nuclear Information System (INIS)

Trunov, Nikolay B.; Davidenko, Stanislav E.; Grigoriev, Vladimir A.; Popadchuk, Valery S.; Brykov, Sergery I.; Karzov, Georgy P.

2006-01-01

At WWER NPPs the horizontal steam generators (SGs), are used that differ in design concept from vertical SGs mostly used at western NPPs. Reliable operation of SG heat-exchanging tubes is the crucial worldwide problem for NPP of various types. According to the operation feedback the water chemistry is the governing factor affecting operability of SG tubing. The secondary side corrosion is considered to be the main mechanism of SG heat-exchanging tubes damage at WWER plants. To make the assessment of the tubing integrity the combination of pressure tests and eddy-current tests is used. Assessment of the tubing performance is an important part of SG life extension practice. The given paper deals with the description of the tube testing strategy and the approach to tube integrity assessment based on deterministic and probabilistic methods of fracture mechanics. Requirements for eddy-current test are given as well. Practice of condition monitoring and implementing the database on steam generators operation are presented. The approach to tubes plugging criteria is described. The research activities on corrosion mechanism studies and residual lifetime evaluation are mentioned. (authors)

Steam generator operation and maintenance

International Nuclear Information System (INIS)

Lee, C.K.

2004-01-01

Corrosion of steam generator tube has resulted in the need for extensive repair and replacement of steam generators. Over the past two decades, steam generator problems in the United States were viewed to be one of the most significant contributor to lost generation in operating PWR plants. When the SGOG-I (Steam Generator Owners Groups) was formed in early 1977, denting was responsible for almost 90% of the tube plugging. By the end of 1982, this figure was reduced to less than 2%. During the existence of SGOG-II (from 1982 to 1986), IGA/SCC (lntergranular Attack/Stress Corrosion Cracking) in the tube sheet, primary side SCC, pitting, and fretting surfaced as the primary causes of tube degradation. Although significant process has been made with wastage and denting, the utilities experience shows that the percentage of reactors plugging tubes and the percentage of tubes being plugged each year has remained relatively constant. The diversity of the damage mechanisms means that no one solution is likely to resolve all problems. The task of maintaining steam generator integrity continues to be formidable and challenging. As the older problems were brought under control, many new problems emerged. SGOG-II (Steam Generator Owners Group program from 1982 to 1986) has focused on these problem areas such as tube stress corrosion cracking (SCC) and intergranular attack (IGA) in the open tube sheet crevice, primary side tube cracking, pitting in the lower span, and tube fretting in preheated section and anti-vibration bar (AVB) locations. Primary Water Stress Corrosion Cracking (PWSCC) in the tube to tubesheet roll transition has been a wide spread problem in the Recirculation Steam Generators (RSG) during this period. Although significant progress has been made in resolving this problem, considerable work still remains. One typical problem in the Once Through Steam Generator (OTSG) was the tube support plate broached hole fouling which affects the OTSG steam generating

MEDEA, Steady-State Pressure and Temperature Distribution in He H2O Steam Generator

International Nuclear Information System (INIS)

Hansen, Ulf

1976-01-01

1 - Nature of physical problem solved: MEDEA calculates the time-independent pressure and temperature distribution in a helium-water steam generator. The changing material properties of the fluids with pressure and temperature are treated exactly. The steam generator may consist of economizer, evaporator, superheater and reheater in variable flow patterns. In case of reheating the high-pressure turbine is taken into account. The main control circuits influencing the behaviour of the system are simulated. These are water spraying of the hot steam, load-dependent control of steam pressure at the HP-turbine inlet and valves before the LP-turbine to ensure constant pressure in the reheater section. Investigations of hydrodynamic flow stability in single tubes can be performed. 2 - Method of solution: The steam generator is calculated as a 1-dimensional model, (i.e. all parallel tubes working under equal conditions) and is divided into small heat exchanger elements with helium and water in ideal parallel or counter flow. The material and thermodynamic properties are kept constant within one element. The calculations start at the cold end of the steam generator and proceed stepwise along the water flow pattern to produce pressure and temperature distributions of helium and water. The gas outlet temperature is changed until convergence is reached with a continuous temperature profile on the gas side. MEDEA chooses the iteration scheme according to flow pattern and other special arrangements in the steam generator. The hydrodynamic stability is calculated for a single tube assuming that all tubes are exposed to the same gas temperature profile and changing the water flow in a single tube will not influence the conditions on the gas side. Varying the water flow by keeping gas temperature constant and repeating the steam generator calculations yield pressure drop and steam temperature as a function of flow rate. 3 - Restrictions on the complexity of the problem: Maximum

Options for Steam Generator Decommissioning

International Nuclear Information System (INIS)

Krause, Gregor; Amcoff, Bjoern; Robinson, Joe

2016-01-01

Selecting the best option for decommissioning steam generators is a key consideration in preparing for decommissioning PWR nuclear power plants. Steam Generators represent a discrete waste stream of large, complex items that can lend themselves to a variety of options for handling, treatment, recycling and disposal. Studsvik has significant experience in processing full size Steam Generators at its metal recycling facility in Sweden, and this paper will introduce the Studsvik steam generator treatment concept and the results achieved to date across a number of projects. The paper will outline the important parameters needed at an early stage to assess options and to help consider the balance between off-site and on-site treatment solutions, and the role of prior decontamination techniques. The paper also outlines the use of feasibility studies and demonstration projects that have been used to help customers prepare for decommissioning. The paper discusses physical, radiological and operational history data, Pro and Contra factors for on- and off-site treatment, the role of chemical decontamination prior to treatment, planning for off-site shipments as well as Studsvik experience This paper has an original focus upon the coming challenges of steam generator decommissioning and potential external treatment capacity constraints in the medium term. It also focuses on the potential during operations or initial shut-down to develop robust plans for steam generator management. (authors)

Design of the steam generator in an energy conversion system based on the aluminum combustion with water

International Nuclear Information System (INIS)

Mercati, Stefano; Milani, Massimo; Montorsi, Luca; Paltrinieri, Fabrizio

2012-01-01

Highlights: ► Development of a numerical approach for the analysis of a co-generation system based on the aluminum water reaction. ► Construction of system operating maps for estimating the system behavior. ► Comparison of two different designs of the steam generator for the system. ► Definition of the operating range where each configuration provides the best performance. -- Abstract: The paper shows the preliminary design of the superheated steam generator to be used in a novel hydrogen production and energy conversion system based on the combustion of aluminum particles with water. The system is aimed at producing hydrogen and pressurized superheated steam, using the heat released by the Al–H 2 O reaction. The interest on this type of technology arises because of the possibility of obtaining hydrogen with very low pollutant and greenhouse gas emissions, compared to the traditional hydrogen production systems, such as the steam reforming from methane. The analysis of the combustion chamber and the heat recovery system is carried out by means of a lumped and distributed parameter numerical approach. The multi phase and gas mixture theoretical principles are used both to characterize the mass flow rate and the heat release in the combustion chamber and within the heat exchangers in order to relate the steam generator performance to the system operating parameters. Finally, the influence of the steam generator performance on the whole energy conversion system behavior is addressed, with particular care to the evaluation of the total power and efficiency variation with the combustion parameters.

Investigation of separation and hydrodynamic characteristics of steam generators used at the NPPs running on PWR-1000 reactors

International Nuclear Information System (INIS)

Ageev, A.G.; Korolkov, B.M.; Nigmatulin, B.I.; Vasileva, R.V.; Nekrasov, A.V.; Titiv, V.F.; Tarankov, G.A.

1997-01-01

The tests were accomplished at the steam generator of unit 5 of the Novovoronezh nuclear power plant. The outbursts of the steam-water mixture from the gap between the steam generator housing and the submerged perforated screen rim at the side of the inlet coolant manifold were investigated. Tests of the steam generator with a modified steam separation system were carried out on the Balakovo nuclear power plant. The gilled separator of the steam generator was replaced with a steam collecting perforated screen, while the gap between the steam generator housing and the heat exchange bundle rim was closed with additional perforated screens at the side of the inlet manifold. This new solution of moisture separation is better. (M.D.)

Scale Thickness Measurement of Steam Generator Tubing Using Eddy Current Signal of Bobbin Coil

International Nuclear Information System (INIS)

Kim, Chang Soo; Um, Ki Soo; Kim, Jae Dong

2012-01-01

Steam generator is one of the major components of nuclear power plant and steam generator tubes are the pressure boundary between primary and secondary side, which makes them critical for nuclear safety. As the operating time of nuclear power plant increases, not only damage mechanisms but also scaled deposits on steam generator tubes are known to be problematic causing tube support flow hole blockage and heat fouling. The ability to assess the extent and location of scaled deposits on tubes became essential for management and maintenance of steam generator and eddy current bobbin data can be utilized to measure thickness of scale on tubes. In this paper, tube reference standards with various thickness of scaled deposit has been set up to provide information about the overall deposit condition of steam generator tubes, providing essential tool for steam generator management and maintenance to predict and prevent future damages. Also, methodology to automatically measure scale thickness on tubes has been developed and applied to field data to estimate overall scale amount.

Circumferential cracking of steam generator tubes

International Nuclear Information System (INIS)

Karwoski, K.J.

1997-04-01

On April 28, 1995, the U.S. Nuclear Regulatory Commission (NRC) issued Generic Letter (GL) 95-03, open-quote Circumferential Cracking of Steam Generator Tubes.close-quote GL 95-03 was issued to obtain information needed to verify licensee compliance with existing regulatory requirements regarding the integrity of steam generator tubes in domestic pressurized-water reactors (PWRs). This report briefly describes the design and function of domestic steam generators and summarizes the staff's assessment of the responses to GL 95-03. The report concludes with several observations related to steam generator operating experience. This report is intended to be representative of significant operating experience pertaining to circumferential cracking of steam generator tubes from April 1995 through December 1996. Operating experience prior to April 1995 is discussed throughout the report, as necessary, for completeness

Steam generator sludge removal apparatus

International Nuclear Information System (INIS)

Schafer, B.W.; Werner, C.E.; Klahn, F.C.

1992-01-01

The present invention relates to equipment for cleaning steam generators and in particular to a high pressure fluid lance for cleaning sludge off the steam generator tubes away from an open tube lane. 6 figs

Advanced technologies on steam generators

Energy Technology Data Exchange (ETDEWEB)

Sakata, Kaoru; Nakamura, Yuuki [Mitsubishi Heavy Industry Co., Takasago (Japan); Nakamori, Nobuo; Mizutani, Toshiyuki; Uwagawa, Seiichi; Saito, Itaru [Mitsubishi Heavy Industry Co., Kobe (Japan); Matsuoka, Tsuyoshi [Mitsubishi Heavy Industry Co., Yokohama (Japan)

1997-12-31

The thermal-hydraulic tests for a horizontal steam generator of a next-generation PWR (New PWR-21) were performed. The purpose of these tests is to understand the thermal-hydraulic behavior in the secondary side of horizontal steam generator during the plant normal operation. A test was carried out with cross section slice model simulated the straight tube region. In this paper, the results of the test is reported, and the effect of the horizontal steam generator internals on the thermalhydraulic behavior of the secondary side and the circulation characteristics of the secondary side are discussed. (orig.). 3 refs.

Advanced technologies on steam generators

Energy Technology Data Exchange (ETDEWEB)

Sakata, Kaoru; Nakamura, Yuuki [Mitsubishi Heavy Industry Co., Takasago (Japan); Nakamori, Nobuo; Mizutani, Toshiyuki; Uwagawa, Seiichi; Saito, Itaru [Mitsubishi Heavy Industry Co., Kobe (Japan); Matsuoka, Tsuyoshi [Mitsubishi Heavy Industry Co., Yokohama (Japan)

1998-12-31

The thermal-hydraulic tests for a horizontal steam generator of a next-generation PWR (New PWR-21) were performed. The purpose of these tests is to understand the thermal-hydraulic behavior in the secondary side of horizontal steam generator during the plant normal operation. A test was carried out with cross section slice model simulated the straight tube region. In this paper, the results of the test is reported, and the effect of the horizontal steam generator internals on the thermalhydraulic behavior of the secondary side and the circulation characteristics of the secondary side are discussed. (orig.). 3 refs.

The decommissioning of the BR3 steam generator

International Nuclear Information System (INIS)

Denissen, L.

2006-01-01

A steam generator is a crucial component in a PWR (Pressurized Water Reactor). It is the crossing between the primary, contaminated, circuit and the secondary water-steam circuit. The heat from the primary reactor coolant loop is transferred to the secondary side in thousands of small tubes. Due to several problems in the material of those tubes, like SCC (Stress Corrosion Cracking), insufficient control in water chemistry, which can be the cause of tube leakage, more and more steam generators are replaced today. Only in Belgium, already 17 of them are replaced. The old 300 ton heavy SGs are stored at the 2 nuclear power plants of Doel and Tihange . While it was foreseen in the BR3 strategy to dismantle the steam generator (only 30 ton), we took the opportunity to search for a complete package in the decommissioning of a steam generator. The complete management consists of a decontamination of the primary side followed by the complete dismantling. The first step, the decontamination with MEDOC (water box + tube bundle) has already been achieved in 2002. It has led to an important DF (Decontamination Factor) between 100 and 1000 and an important dose rate reduction. This hard chemical decontamination process has been described earlier in the scientific report 2002 (The BR3 steam generator decontamination with the MEDOC process - M. Ponnet). The second step, the complete dismantling of the SG has been executed in 2005. With the BR3 SG, the main goal was to dismantle it in a safe way and to free release a maximum of material. We've used two cutting tools to perform the job: A HPWJC (High Pressure Water Jet Cutting) tool in combination with a hydraulic robot and a water cooled diamond cable. The last technique was done in close collaboration with the external company Husqvarna. It was important to have an idea of the performance, the efficiency of the cable and the quantity and the type of secondary waste

On-Line Monitoring and Diagnostics of the Integrity of Nuclear Plant Steam Generators and Heat Exchangers

International Nuclear Information System (INIS)

Upadhyaya, Belle R.; Wesley Hines, J.

2004-01-01

The overall purpose of this Nuclear Engineering Education Research (NEER) project was to integrate new, innovative, and existing technologies to develop a fault diagnostics and characterization system for nuclear plant steam generators (SG) and heat exchangers (HX). Issues related to system level degradation of SG and HX tubing, including tube fouling, performance under reduced heat transfer area, and the damage caused by stress corrosion cracking, are the important factors that influence overall plant operation, maintenance, and economic viability of nuclear power systems. The research at The University of Tennessee focused on the development of techniques for monitoring process and structural integrity of steam generators and heat exchangers. The objectives of the project were accomplished by the completion of the following tasks. All the objectives were accomplished during the project period. This report summarizes the research and development activities, results, and accomplishments during June 2001-September 2004. (1) Development and testing of a high-fidelity nodal model of a U-tube steam generator (UTSG) to simulate the effects of fouling and to generate a database representing normal and degraded process conditions. Application of the group method of data handling (GMDH) method for process variable prediction. (2) Development of a laboratory test module to simulate particulate fouling of HX tubes and its effect on overall thermal resistance. Application of the GMDH technique to predict HX fluid temperatures, and to compare with the calculated thermal resistance. (3) Development of a hybrid modeling technique for process diagnosis and its evaluation using laboratory heat exchanger test data. (4) Development and testing of a sensor suite using piezo-electric devices for monitoring structural integrity of both flat plates (beams) and tubing. Experiments were performed in air, and in water with and without bubbly flow. (5) Development of advanced signal

On-Line Monitoring and Diagnostics of the Integrity of Nuclear Plant Steam Generators and Heat Exchangers.

Energy Technology Data Exchange (ETDEWEB)

Belle R. Upadhyaya; J. Wesley Hines

2004-09-27

The overall purpose of this Nuclear Engineering Education Research (NEER) project was to integrate new, innovative, and existing technologies to develop a fault diagnostics and characterization system for nuclear plant steam generators (SG) and heat exchangers (HX). Issues related to system level degradation of SG and HX tubing, including tube fouling, performance under reduced heat transfer area, and the damage caused by stress corrosion cracking, are the important factors that influence overall plant operation, maintenance, and economic viability of nuclear power systems. The research at The University of Tennessee focused on the development of techniques for monitoring process and structural integrity of steam generators and heat exchangers. The objectives of the project were accomplished by the completion of the following tasks. All the objectives were accomplished during the project period. This report summarizes the research and development activities, results, and accomplishments during June 2001-September 2004. (1) Development and testing of a high-fidelity nodal model of a U-tube steam generator (UTSG) to simulate the effects of fouling and to generate a database representing normal and degraded process conditions. Application of the group method of data handling (GMDH) method for process variable prediction. (2) Development of a laboratory test module to simulate particulate fouling of HX tubes and its effect on overall thermal resistance. Application of the GMDH technique to predict HX fluid temperatures, and to compare with the calculated thermal resistance. (3) Development of a hybrid modeling technique for process diagnosis and its evaluation using laboratory heat exchanger test data. (4) Development and testing of a sensor suite using piezo-electric devices for monitoring structural integrity of both flat plates (beams) and tubing. Experiments were performed in air, and in water with and without bubbly flow. (5) Development of advanced signal

Thermo-economic optimization of heat recovery steam generator for a range of gas turbine exhaust temperatures

International Nuclear Information System (INIS)

Nadir, Mahmoud; Ghenaiet, Adel; Carcasci, Carlo

2016-01-01

Highlights: • Thermo-economic optimization of HRSG configurations. • The maximum value of the net present value was targeted for the economic optimization. • Three level HRSG is the best option in respect of power output and high priced medium. • Two level HRSG is the best for net benefit in low and intermediate priced mediums. - Abstract: This paper illustrates the effect of selling price on the optimum design parameters of a heat recovery steam generator (HRSG) and the selection of its ideal configuration for an outlet temperature range of 350–650 °C. The Particle Swarm Optimization (PSO) method was used, considering the steam cycle specific work as an objective to be maximized, the net present value as another objective to be maximized for the economic optimization and a combination of both. Three configurations of heat recovery steam generators are considered with one, two and three pressure levels and a reheat. The results show that, the three pressure level system is the best configuration from a thermodynamic point of view, but with respect to the economical aspect the two pressure levels is the best configuration for the low and medium selling prices (0.04 $/kW h, 0.08 $/kW h and 0.2 $/kW h), whereas the three pressure level configuration would only be interesting for a high selling price of 0.3 $/kW h and a temperature range 450–600 °C. For a temperature of 650 °C, the high cost of the three level system leads to a decrease in the net present value. As the selling price increases the optimized design parameters of the three pressure level HRSG based on economic or thermodynamic optimization are similar. The obtained results are used to elaborate a new correlation relating the net present value with the gas turbine outlet temperature, gas mass flow rate, number of levels of HRSG and selling price.

Steam turbine cycle

International Nuclear Information System (INIS)

Okuzumi, Naoaki.

1994-01-01

In a steam turbine cycle, steams exhausted from the turbine are extracted, and they are connected to a steam sucking pipe of a steam injector, and a discharge pipe of the steam injector is connected to an inlet of a water turbine. High pressure discharge water is obtained from low pressure steams by utilizing a pressurizing performance of the steam injector and the water turbine is rotated by the high pressure water to generate electric power. This recover and reutilize discharged heat of the steam turbine effectively, thereby enabling to improve heat efficiency of the steam turbine cycle. (T.M.)

Steam generator corrosion 2007; Dampferzeugerkorrosion 2007

Energy Technology Data Exchange (ETDEWEB)

Born, M. (ed.)

2007-07-01

Between 8th and 9th November, 2007, SAXONIA Standortentwicklungs- und -verwertungsgesellschaft GmbH (Freiberg, Federal Republic of Germany) performed the 3rd Freiberger discussion conference ''Fireside boiler corrosion''. The topics of the lectures are: (a) Steam generator corrosion - an infinite history (Franz W. Alvert); (b) CFD computations for thermal waste treatment plants - a contribution for the damage recognition and remedy (Klaus Goerner, Thomas Klasen); (c) Experiences with the use of corrosion probes (Siegfried R. Horn, Ferdinand Haider, Barbara Waldmann, Ragnar Warnecke); (d) Use of additives for the limitation of the high temperature chlorine corrosion as an option apart from other measures to the corrosion protection (Wolfgang Spiegel); (e) Current research results and aims of research with respect to chlorine corrosion (Ragnar Warnecke); (f) Systematics of the corrosion phenomena - notes for the enterprise and corrosion protection (Thomas Herzog, Wolfgang Spiegel, Werner Schmidl); (g) Corrosion protection by cladding in steam generators of waste incinerators (Joerg Metschke); (h) Corrosion protection and wear protection by means of thermal spraying in steam generators (Dietmar Bendix); (i) Review of thick film nickelized components as an effective protection against high-temperature corrosion (Johann-Wilhelm Ansey); (j) Fireproof materials for waste incinerators - characteristics and profile of requirement (Johannes Imle); (k) Service life-relevant aspects of fireproof linings in the thermal recycling of waste (Till Osthoevener and Wolfgang Kollenberg); (l) Alternatives to the fireproof material in the heating space (Heino Sinn); (m) Cladding: Inconal 625 contra 686 - Fundamentals / applications in boiler construction and plant construction (Wolfgang Hoffmeister); (n) Thin films as efficient corrosion barriers - thermal spray coating in waste incinerators and biomass firing (Ruediger W. Schuelein, Steffen Hoehne, Friedrich

Steam generator tube rupture (SGTR) scenarios

International Nuclear Information System (INIS)

Auvinen, A.; Jokiniemi, J.K.; Laehde, A.; Routamo, T.; Lundstroem, P.; Tuomisto, H.; Dienstbier, J.; Guentay, S.; Suckow, D.; Dehbi, A.; Slootman, M.; Herranz, L.; Peyres, V.; Polo, J.

2005-01-01

The steam generator tube rupture (SGTR) scenarios project was carried out in the EU 5th framework programme in the field of nuclear safety during years 2000-2002. The first objective of the project was to generate a comprehensive database on fission product retention in a steam generator. The second objective was to verify and develop predictive models to support accident management interventions in steam generator tube rupture sequences, which either directly lead to severe accident conditions or are induced by other sequences leading to severe accidents. The models developed for fission product retention were to be included in severe accident codes. In addition, it was shown that existing models for turbulent deposition, which is the dominating deposition mechanism in dry conditions and at high flow rates, contain large uncertainties. The results of the project are applicable to various pressurised water reactors, including vertical steam generators (western PWR) and horizontal steam generators (VVER)

Minimize corrosion degradation of steam generator tube materials

International Nuclear Information System (INIS)

Lu, Y.

2006-01-01

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. Experimental data suggest that all steam generator tube materials are susceptible to corrosion degradation under some specific off-specification conditions. The tolerance to the chemistry upset for each steam generator tube alloy is different. Electrochemical corrosion behaviors of major steam generator tube alloys were studied under the plausible aggressive crevice chemistry conditions. The potential hazardous conditions leading to steam generator tube degradation and the conditions, which can minimize steam generator tube degradation have been determined. Recommended electrochemical corrosion potential/pH zones were defined for all major steam generator tube materials, including Alloys 600, 800, 690 and 400, under CANDU steam generator operating and startup conditions. Stress corrosion cracking tests and accelerated corrosion tests were carried out to verify and revise the recommended electrochemical corrosion potential/pH zones. Based on this information, utilities can prevent steam generator material degradation surprises by appropriate steam generator water chemistry management and increase the reliability of nuclear power generating stations. (author)

High-definition radiography of tube-to-tubesheet welds of steam generator of prototype fast breeder reactor

International Nuclear Information System (INIS)

Venkatraman, B.; Sethi, V.K.; Jayakumar, T.; Raj, B.

1995-01-01

In the steam generator of the Prototype Fast Breeder Reactor (PFBR), steam is generated by the transfer of heat from secondary sodium to water. Due to the inherent dangers of sodium-water reaction, the integrity of weld joints separating sodium and water/steam is of paramount importance. This is particularly true and very important for the tube-to-tubesheet joints. This paper discusses the use of projective magnification technique by microfocal radiography for the quality evaluation and optimisation of the welding parameters of such small tube-to-tubesheet welds of the steam generator of PFBR. (author)

Expandable antivibration bar for a steam generator

International Nuclear Information System (INIS)

Lagally, H.O.

1987-01-01

This patent describes a steam generator for a nuclear power plant comprising a shell, a plurality of tubes having a U-shaped configuration arranged in successive columns within the shell. The tubes are adapted to heat feedwater flowing around the outside of the tubes by the flow of hot reactor coolant within the tubes, and antivibration bars any vibrations of the tubes as a result of steam between the columns of tubes. The improvement described here comprises means for varying the thickness of the antivibration bars to fit substantially the actual space between the columns of tubes comprising first and second bars, with at least one bar being movable, and with at least one mating inclined surface between the first and second bars

The Evaluation of Steam Generator Level Measurement Model for OPR1000 Using RETRAN-3D

International Nuclear Information System (INIS)

Doo Yong Lee; Soon Joon Hong; Byung Chul Lee; Heok Soon Lim

2006-01-01

Steam generator level measurement is important factor for plant transient analyses using best estimate thermal hydraulic computer codes since the value of steam generator level is used for steam generator level control system and plant protection system. Because steam generator is in the saturation condition which includes steam and liquid together and is the place that heat exchange occurs from primary side to secondary side, computer codes are hard to calculate steam generator level realistically without appropriate level measurement model. In this paper, we prepare the steam generator models using RETRAN-3D that include geometry models, full range feedwater control system and five types of steam generator level measurement model. Five types of steam generator level measurement model consist of level measurement model using elevation difference in downcomer, 1D level measurement model using fluid mass, 1D level measurement model using fluid volume, 2D level measurement model using power and fluid mass, and 2D level measurement model using power and fluid volume. And we perform the evaluation of the capability of each steam generator level measurement model by simulating the real plant transient condition, the title is 'Reactor Trip by The Failure of The Deaerator Level Control Card of Ulchin Unit 3'. The comparison results between real plant data and RETRAN-3D analyses for each steam generator level measurement model show that 2D level measurement model using power and fluid mass or fluid volume has more realistic prediction capability compared with other level measurement models. (authors)

Economic analysis of process steam and electricity generation by a 200 MW NHR

International Nuclear Information System (INIS)

Tian Li; Wang Yongqing

2000-01-01

New applications for low temperature nuclear heating reactors should be developed using economic analysis. This paper compares and analyzes the economics of the generation 1.5 MPa process steam and electricity by a 200 MW nuclear heating reactor (NHR-200) for industrial development. The project is very attractive economically with an internal rate of return of 19.61%, a net present worth (discount rate 10%) of 765 million yuan RMB and a capital recovery or payback period of about 5 years after construction is completed. Compared with only using the NHR-200 for in winter heating, the economic of process steam and electricity generation by NHR-200 are much better. In addition, the NHR-200 will significantly improve environmental pollution in cities and reduce the transport of coal from north to south in China

Steam generators secondary side chemical cleaning at Point Lepreau using the Siemens high temperature process

International Nuclear Information System (INIS)

Verma, K.; MacNeil, C.; Odar, S.; Kuhnke, K.

1997-01-01

This paper describes the chemical cleaning of the four steam generators at the Point Lepreau facility, which was accomplished as a part of a normal service outage. The steam generators had been in service for twelve years. Sludge samples showed the main elements were Fe, P and Na, with minor amounts of Ca, Mg, Mn, Cr, Zn, Cl, Cu, Ni, Ti, Si, and Pb, 90% in the form of Magnetite, substantial phosphate, and trace amounts of silicates. The steam generators were experiencing partial blockage of broached holes in the TSPs, and corrosion on tube ODs in the form of pitting and wastage. In addition heat transfer was clearly deteriorating. More than 1000 kg of magnetite and 124 kg of salts were removed from the four steam generators

Technical specifications for the provision of heat and steam sources for INPP and Visaginas. Final report

International Nuclear Information System (INIS)

2003-01-01

In October 1999, the National Energy Strategy was approved by the Lithuanian Parliament. The National Energy Strategy included the decision to close Unit-1 of INPP before 2005. Later is has been decided to close Unit 2 before the end of 2009 as well. The closure and decommissioning will have heavy impact on the heat supply for the city of Visaginas. Unit 1 and Unit 2 of INPP supplies hot water and steam to INPP for process purposes and for space heating of residential and commercial buildings. When Unit 1 is permanently shut down, reliable heat and steam sources independent of the power plants own heat and steam generation facilities are required for safety reasons in the event of shutdown of the remaining unit for maintenance or in an emergency. These steam and heat sources must be operational before single unit operation is envisaged. Provision of a reliable independent heat and steam source is therefore urgent. After both reactors are shut down permanently, a steam source will be needed at the plant for radioactive waste storage and disposal. INPP and DEA has performed a feasibility study for the provision of a reliable heat source for Ignalina Nuclear Power Plant and Visaginas, and the modernisation of Visaginas district heating system. The objective of this project is to prepare technical specifications for the provision of new heat and steam sources for INPP and Visaginas, and for rehabilitation of the heat transmission pipeline between INPP, the back-up boiler station and Visaginas City. The results of the study are presented in detail in the reports and technical specifications: 1. Transient analysis for Visaginas DH system, 2. Non-destructive testing of boiler stations, pump stations and transmission lines, 3. Conceptual design, 4. Technical specifications, Package 1 to 6. The study has suggested: 1. Construction of new steam boiler station, 2. Construction of new heat only boiler station, 3. Renovation of existing back-up heat only boiler station, 4

Steam generator development in France for the Super Phenix project; Generateurs de vapeur developpes en France pour Super Phenix

Energy Technology Data Exchange (ETDEWEB)

Robin, M G

1975-07-01

'Steam Generator Development for Super Phenix Project'. The development program of steam generators studied by Fives-Cail Babcock and Stein Industrie Companies, jointly with CEA end EDF, for the Super Phenix 1200 MWe Fast Breeder Power Plant, is presented. The main characteristics of both sodium heated steam generators are emphasized and experimental studies related to their key features are reported. (author)

Two-Phase Instability Characteristics of Printed Circuit Steam Generator for the Low Pressure Condition

International Nuclear Information System (INIS)

Kang, Han-Ok; Han, Hun Sik; Kim, Young-In; Kim, Keung Koo

2015-01-01

Reduction of installation space for steam generators can lead to much smaller reactor vessel with resultant decrease of overall manufacturing cost for the components. A PCHE(Printed Circuit Heat Exchanger) is one of the compact types of heat exchangers available as an alternative to conventional shell and tube heat exchangers. Its name is derived from the procedure used to manufacture the flat metal plates that form the core of the heat exchanger, which is done by chemical milling. These plates are then stacked and diffusion bonded, converting the plates into a solid metal block containing precisely engineered fluid flow passages. PCSG(Printed Circuit Steam Generator) is a potential candidate to be applied to the integral reactor with its compactness and mechanical robustness. For the introduction of new steam generator, design requirement for the two-phase flow instability should be considered. This paper describes two-phase flow instability characteristics of PCSG for the low pressure condition. PCSG is a potential candidate to be applied to the integral reactor with its compactness and mechanical robustness. Interconnecting flow path was developed to mitigate the two-phase flow instability in the cold side. The flow characteristics of two-phase flow instability at the PCSG is examined experimentally in this study

Effect of liquid waste discharges from steam generating facilities

Energy Technology Data Exchange (ETDEWEB)

McGuire, H.E. Jr.

1977-09-01

This report contains a summary of the effects of liquid waste discharges from steam electric generating facilities on the environment. Also included is a simplified model for use in approximately determining the effects of these discharges. Four basic fuels are used in steam electric power plants: three fossil fuels--coal, natural gas, and oil; and uranium--presently the basic fuel of nuclear power. Coal and uranium are expected to be the major fuels in future years. The following power plant effluents are considered: heat, chlorine, copper, total dissolved solids, suspended solids, pH, oil and grease, iron, zinc, chrome, phosphorus, and trace radionuclides.

Effect of liquid waste discharges from steam generating facilities

International Nuclear Information System (INIS)

McGuire, H.E. Jr.

1977-09-01

This report contains a summary of the effects of liquid waste discharges from steam electric generating facilities on the environment. Also included is a simplified model for use in approximately determining the effects of these discharges. Four basic fuels are used in steam electric power plants: three fossil fuels--coal, natural gas, and oil; and uranium--presently the basic fuel of nuclear power. Coal and uranium are expected to be the major fuels in future years. The following power plant effluents are considered: heat, chlorine, copper, total dissolved solids, suspended solids, pH, oil and grease, iron, zinc, chrome, phosphorus, and trace radionuclides

An expert system for steam generator maintenance

International Nuclear Information System (INIS)

Remond, A.

1988-01-01

The tube bundles in PWR steam generators are, by far, the major source of problems whether they are due to primary and secondary side corrosion mechanisms or to tube vibration-induced wear at tube support locations. Because of differences in SG operating, materials, and fabrication processes, the damage may differ from steam generator to steam generator. MPGV, an expert system for steam generator maintenance uses all steam generator data containing data on materials, fabrication processes, inservice inspection, and water chemistry. It has access to operational data for individual steam generators and contains models of possible degradation mechanisms. The objectives of the system are: · Diagnosing the most probable degradation mechanism or mechanisms by reviewing the entire steam generator history. · Identifying the tubes most exposed to future damage and evaluating the urgency of repair by simulating the probable development of the problem in time. · Establishing the appropriate preventive actions such as repair, inspection or other measures and establishing an action schedule. The system is intended for utilities either for individual plants before each inspection outage or any time an incident occurs or for a set of plants through a central MPGV center. (author)

Main characteristics and design features of steam generators for VG-400 plant

International Nuclear Information System (INIS)

Golovko, V.F.; Grebennik, V.N.; Gol'tsev, A.O.; Ivanov, S.M.; Sergeev, A.I.; Pospelov, V.N.

1988-01-01

The description of a steam generator for the VG-400 plant performed in two variants depending on a heat-exchange surface arrangement (one-bundle coil and module-cassette construction) is given. (author)

Measurement of condensation heat transfer coefficients in a steam chamber using a variable conductance heat pipe

International Nuclear Information System (INIS)

Robinson, J.A.; Windebank, S.R.

1988-01-01

Condensation heat transfer coefficients have been measured in a pressurised chamber containing a mixture of saturated steam and air. They were determined as a function of the air-steam ratio in nominally stagnant conditions. The effect of pressure is assessed and preliminary measurements with a forced convective component of velocity are presented. A novel measurement technique was adopted, namely to use a vertical heat pipe whose conductance could easily be varied. It transported heat from an evaporator located inside the chamber to a condenser section outside, at which the heat flow was measured. Heat flux at the evaporator could then be determined and a condensation heat transfer coefficient derived. The range of coefficients covered was from 150 W/m 2 0 K at high air-steam ratios to 20,000 W/m 2 0 K in pure steam. Results show that increasing either total pressure or velocity enhances condensation heat transfer over the range of air/steam ratios considered. (author)

Steam generators: critical components in nuclear steam supply systems

Energy Technology Data Exchange (ETDEWEB)

Stevens-Guille, P D

1974-02-28

Steam generators are critical components in power reactors. Even small internal leaks result in costly shutdowns for repair. Surveys show that leaks have affected one half of all water-cooled reactors in the world with steam generators. CANDU reactors have demonstrated the highest reliability. However, AECL is actively evolving new technology in design, manufacture, inspection and operation to maintain reliability. (auth)

Analysis of Steam Heating of a Two-Layer TBP/N-Paraffin/Nitric Acid Mixtures

International Nuclear Information System (INIS)

Laurinat, J.E.; Hassan, N.M.; Rudisill, T.S.; Askew, N.M.

1998-01-01

This report presents an analysis of steam heating of a two-layer tri-n-butyl phosphate (TBP)/n-paraffin-nitric acid mixture.The purpose of this study is to determine if the degree of mixing provided by the steam jet or by bubbles generated by the TBP/nitric acid reaction is sufficient to prevent a runaway reaction

Moisture separator for steam generator level measurement system

International Nuclear Information System (INIS)

Cantineau, B.J.

1987-01-01

A steam generator level measurement system having a reference leg which is kept full of water by a condensation pot, has a liquid/steam separator in the connecting line between the condensation pot and the steam phase in the steam generator to remove excess liquid from the steam externally of the steam generator. This ensures that the connecting line does not become blocked. The separator pot has an expansion chamber which slows down the velocity of the steam/liquid mixture to aid in separation, and a baffle, to avoid liquid flow into the line connected to the condensate pot. Liquid separated is returned to the steam generator below the water level through a drain line. (author)

Macroporous Double-Network Hydrogel for High-Efficiency Solar Steam Generation Under 1 sun Illumination.

Science.gov (United States)

Yin, Xiangyu; Zhang, Yue; Guo, Qiuquan; Cai, Xiaobing; Xiao, Junfeng; Ding, Zhifeng; Yang, Jun

2018-04-04

Solar steam generation is one of the most promising solar-energy-harvesting technologies to address the issue of water shortage. Despite intensive efforts to develop high-efficiency solar steam generation devices, challenges remain in terms of the relatively low solar thermal efficiency, complicated fabrications, high cost, and difficulty in scaling up. Herein, a double-network hydrogel with a porous structure (p-PEGDA-PANi) is demonstrated for the first time as a flexible, recyclable, and efficient photothermal platform for low-cost and scalable solar steam generation. As a novel photothermal platform, the p-PEGDA-PANi involves all necessary properties of efficient broadband solar absorption, exceptional hydrophilicity, low heat conductivity, and porous structure for high-efficiency solar steam generation. As a result, the hydrogel-based solar steam generator exhibits a maximum solar thermal efficiency of 91.5% with an evaporation rate of 1.40 kg m -2 h -1 under 1 sun illumination, which is comparable to state-of-the-art solar steam generation devices. Furthermore, the good durability and environmental stability of the p-PEGDA-PANi hydrogel enables a convenient recycling and reusing process toward real-life applications. The present research not only provides a novel photothermal platform for solar energy harvest but also opens a new avenue for the application of the hydrogel materials in solar steam generation.

Guidelines for random excitation forces due to cross flow in steam generators

Energy Technology Data Exchange (ETDEWEB)

Taylor, C.E.; Pettigrew, M.J. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

1998-07-01

Random excitation forces can cause low-amplitude tube motion that will result in long-term fretting-wear or fatigue. To prevent these tube failures in steam generators and other heat exchangers, designers and trouble-shooters must have guidelines that incorporate random or turbulent fluid forces. Experiments designed to measure fluid forces have been carried out at Chalk River Laboratories and at other labs around the world. The data from these experiments have been studied and collated to determine suitable guidelines for random excitation forces. In this paper, a guideline for random excitation forces in single-phase cross flow is presented in the form of normalised spectra that are applicable to a wide range of flow conditions and tube frequencies. In particular, the experimental results used in this study were carried out over the full range of flow conditions found in a nuclear steam generator. The proposed guidelines are applicable to steam generators, condensers, reheaters and other shell-and-tube heat exchangers. They may be used for flow-induced vibration analysis of new or existing components, as input to vibration analysis computer codes and as specifications in procurement documents. (author)

Guidelines for random excitation forces due to cross flow in steam generators

International Nuclear Information System (INIS)

Taylor, C.E.; Pettigrew, M.J.

1998-01-01

Random excitation forces can cause low-amplitude tube motion that will result in long-term fretting-wear or fatigue. To prevent these tube failures in steam generators and other heat exchangers, designers and trouble-shooters must have guidelines that incorporate random or turbulent fluid forces. Experiments designed to measure fluid forces have been carried out at Chalk River Laboratories and at other labs around the world. The data from these experiments have been studied and collated to determine suitable guidelines for random excitation forces. In this paper, a guideline for random excitation forces in single-phase cross flow is presented in the form of normalised spectra that are applicable to a wide range of flow conditions and tube frequencies. In particular, the experimental results used in this study were carried out over the full range of flow conditions found in a nuclear steam generator. The proposed guidelines are applicable to steam generators, condensers, reheaters and other shell-and-tube heat exchangers. They may be used for flow-induced vibration analysis of new or existing components, as input to vibration analysis computer codes and as specifications in procurement documents. (author)

Design study on steam generator integration into the VVER reactor pressure vessel

International Nuclear Information System (INIS)

Hort, J.; Matal, O.

2004-01-01

The primary circuit of VVER (PWR) units is arranged into loops where the heat generated by the reactor is removed by means of main circulating pumps, loop pipelines and steam generators, all located outside the reactor pressure vessel. If the primary circuit and reactor core were integrated into one pressure vessel, as proposed, e.g., within the IRIS project (WEC), a LOCA situation would be limited by the reactor pressure vessel integrity only. The aim of this design study regarding the integration of the steam generator into the reactor pressure vessel was to identify the feasibility limits and some issues. Fuel elements and the reactor pressure vessel as used in the Temelin NPP were considered for the analysis. From among the variants analyzed, the variant with steam generators located above the core and vertically oriented circulating pumps at the RPV lower bottom seems to be very promising for future applications

Passive system with steam-water injector for emergency supply of NPP steam generators

International Nuclear Information System (INIS)

Il'chenko, A.G.; Strakhov, A.N.; Magnitskij, D.N.

2009-01-01

The calculation results of reliability indicators of emergency power supply system and emergency feed-water supply system of serial WWER-1000 unit are presented. To ensure safe water supply to steam generators during station blackout it was suggested using additional passive emergency feed-water system with a steam-water injector working on steam generators dump steam. Calculated analysis of steam-water injector operating capacity was conducted at variable parameters of steam at the entrance to injector, corresponding to various moments of time from the beginning of steam-and-water damping [ru

Possibilities of the metallurgical base in the manufacture of tubes for nuclear power plant steam generators

International Nuclear Information System (INIS)

Prnka, T.; Walder, V.; Dolenek, J.

Current possibilities are briefly summarized of metallurgy in the manufacture of high-quality tubes for nuclear power plant steam generators, mainly for fast reactor power plants. Discussed are steel making possibilities, semi-finished product and tube forming with special regard to 2.25Cr1MoNiNb steel problems, heat treatment, finishing, and testing. Necessary equipment and technology for the production of steam generator tubes are less common in the existing practice and are demanding on investment; their introduction, however, is inevitable for securing quality production of steam generator tubes. (Kr)

On-Line Monitoring and Diagnostics of the Integrity of Nuclear Plant Steam Generators and Heat Exchangers, Volumes 1, 2

Energy Technology Data Exchange (ETDEWEB)

Upadhyaya, Belle R. [Univ. of Tennessee, Knoxville, TN (United States); Hines, J. Wesley [Univ. of Tennessee, Knoxville, TN (United States); Lu, Baofu [Univ. of Tennessee, Knoxville, TN (United States)

2005-06-03

The overall purpose of this Nuclear Engineering Education Research (NEER) project was to integrate new, innovative, and existing technologies to develop a fault diagnostics and characterization system for nuclear plant steam generators (SG) and heat exchangers (HX). Issues related to system level degradation of SG and HX tubing, including tube fouling, performance under reduced heat transfer area, and the damage caused by stress corrosion cracking, are the important factors that influence overall plant operation, maintenance, and economic viability of nuclear power systems. The research at The University of Tennessee focused on the development of techniques for monitoring process and structural integrity of steam generators and heat exchangers. The objectives of the project were accomplished by the completion of the following tasks. All the objectives were accomplished during the project period. This report summarizes the research and development activities, results, and accomplishments during June 2001 September 2004. Development and testing of a high-fidelity nodal model of a U-tube steam generator (UTSG) to simulate the effects of fouling and to generate a database representing normal and degraded process conditions. Application of the group method of data handling (GMDH) method for process variable prediction. Development of a laboratory test module to simulate particulate fouling of HX tubes and its effect on overall thermal resistance. Application of the GMDH technique to predict HX fluid temperatures, and to compare with the calculated thermal resistance.Development of a hybrid modeling technique for process diagnosis and its evaluation using laboratory heat exchanger test data. Development and testing of a sensor suite using piezo-electric devices for monitoring structural integrity of both flat plates (beams) and tubing. Experiments were performed in air, and in water with and without bubbly flow. Development of advanced signal processing methods using

Fabrication and inspection development for CRBRP steam generators

International Nuclear Information System (INIS)

McClung, R.W.; Slaughter, G.M.; Spalaris, C.N.; Lillie, A.F.

1975-09-01

One of the critical nonnuclear elements of the CRBRP is the steam generator that transfers the heat from the sodium system to the high-pressure steam system but must maintain integrity and separation of the two fluids. The construction material is 2 1 / 4 Cr--1 Mo alloy steel with high-purity (e.g. vacuum arc remelt) material being used for the tubing and tubesheets. For confidence in successful manufacturing of the several evaporator and superheater modules, key development activities are under way (1) for procurement of high-quality components, (2) to assure proper assembly (with emphasis on welding), and (3) to assure that adequate nondestructive testing methods are available to examine the units. (auth)

Polymeric dispersants for control of steam generator fouling

International Nuclear Information System (INIS)

Balakrishnan, P.V.; Klimas, S.J.; Lepine, L.; Turner, C.W.

1999-05-01

Fouling of steam generators by corrosion products from the feedtrain leads to loss of heat-transfer efficiency, disturbances in thermalhydraulics, and potential corrosion problems resulting from the development of sites for localized accumulation of aggressive chemicals. This report summarizes studies of the use of polymeric dispersants for the control of fouling, which were conducted at the Chalk River Laboratories. High-temperature settling studies on magnetite suspensions were performed to screen available generic dispersants, and the dispersants were ranked in terms of their dispersion efficiency; polyacrylic acid (PAA) and the phosphonate, HEDP, were ranked as the most efficient. Polyacrylic acid was considered more suitable than HEDP for nuclear steam generators, and more emphasis was given to the former in these studies. The dispersants had no effect on the particle deposition rates under single-phase forced-convective flow, but did reduce the deposition rates under flow-boiling conditions. The extent to which the deposition rates were reduced increased in proportion to the dispersant concentration. Preliminary corrosion tests indicated that pitting or general corrosion of steam generator tube materials in the presence of PAA was negligible. Corrosion of carbon steel, although higher in a magnetite-packed crevice under heat flux than in bulk water, was lower in the presence of PAA than in its absence. Some impurities (e.g., sulphate, sodium) were observed in commercially available PAA products at small, though significant concentrations, making these products unacceptable for use in nuclear plants. However, the PAA could be purified by ion exchange. Preliminary experiments, to assess the thermal stability of PAA at steam generator operating temperature, showed the polymer to break down in deaerated solutions and under argon cover to give hydrogen and carbon dioxide as the two major products in the gas phase and variable concentrations of acetate and formate

An homogeneous model of steam generator to simulate operational transiento and accidents in PWR nuclear power plants

International Nuclear Information System (INIS)

Souza, A.L. de.

1981-07-01

GEVAP - A digital computer code was developed to simulate the thermodynamic transient behaviour of steam generators. The steam generator is divided in heating sections. In each section, the conservation equations of mass and energy are integrated numerically, using a predictor-corrector method. As good reslts where obtained, as compared to transients simulated using more detainled codes, it is concluded that GEVAP can be included as the steam generator module of a more complete systems simulation code for PWR's. (E.G.) [pt

Real-time dynamic analysis for complete loop of direct steam generation solar trough collector

International Nuclear Information System (INIS)

Guo, Su; Liu, Deyou; Chu, Yinghao; Chen, Xingying; Shen, Bingbing; Xu, Chang; Zhou, Ling; Wang, Pei

2016-01-01

Highlights: • A nonlinear distribution parameter dynamic model has been developed. • Real-time local heat transfer coefficient and friction coefficient are adopted. • The dynamic behavior of the solar trough collector loop are simulated. • High-frequency chattering of outlet fluid flow are analyzed and modeled. • Irradiance disturbance at subcooled water region generates larger influence. - Abstract: Direct steam generation is a potential approach to further reduce the levelized electricity cost of solar trough. Dynamic modeling of the collector loop is essential for operation and control of direct steam generation solar trough. However, the dynamic behavior of fluid based on direct steam generation is complex because of the two-phase flow in the pipeline. In this work, a nonlinear distribution parameter model has been developed to model the dynamic behaviors of direct steam generation parabolic trough collector loops under either full or partial solar irradiance disturbance. Compared with available dynamic model, the proposed model possesses two advantages: (1) real-time local values of heat transfer coefficient and friction resistance coefficient, and (2) considering of the complete loop of collectors, including subcooled water region, two-phase flow region and superheated steam region. The proposed model has shown superior performance, particularly in case of sensitivity study of fluid parameters when the pipe is partially shaded. The proposed model has been validated using experimental data from Solar Thermal Energy Laboratory of University of New South Wales, with an outlet fluid temperature relative error of only 1.91%. The validation results show that: (1) The proposed model successfully outperforms two reference models in predicting the behavior of direct steam generation solar trough. (2) The model theoretically predicts that, during solar irradiance disturbance, the discontinuities of fluid physical property parameters and the moving back and

Detailed partial load investigation of a thermal energy storage concept for solar thermal power plants with direct steam generation

Science.gov (United States)

Seitz, M.; Hübner, S.; Johnson, M.

2016-05-01

Direct steam generation enables the implementation of a higher steam temperature for parabolic trough concentrated solar power plants. This leads to much better cycle efficiencies and lower electricity generating costs. For a flexible and more economic operation of such a power plant, it is necessary to develop thermal energy storage systems for the extension of the production time of the power plant. In the case of steam as the heat transfer fluid, it is important to use a storage material that uses latent heat for the storage process. This leads to a minimum of exergy losses during the storage process. In the case of a concentrating solar power plant, superheated steam is needed during the discharging process. This steam cannot be superheated by the latent heat storage system. Therefore, a sensible molten salt storage system is used for this task. In contrast to the state-of-the-art thermal energy storages within the concentrating solar power area of application, a storage system for a direct steam generation plant consists of a latent and a sensible storage part. Thus far, no partial load behaviors of sensible and latent heat storage systems have been analyzed in detail. In this work, an optimized fin structure was developed in order to minimize the costs of the latent heat storage. A complete system simulation of the power plant process, including the solar field, power block and sensible and latent heat energy storage calculates the interaction between the solar field, the power block and the thermal energy storage system.

Steam generator life cycle management: Ontario Power Generation (OPG) experience

International Nuclear Information System (INIS)

Maruska, C.C.

2002-01-01

A systematic managed process for steam generators has been implemented at Ontario Power Generation (OPG) nuclear stations for the past several years. One of the key requirements of this managed process is to have in place long range Steam Generator Life Cycle Management (SG LCM) plans for each unit. The primary goal of these plans is to maximize the value of the nuclear facility through safe and reliable steam generator operation over the expected life of the units. The SG LCM plans integrate and schedule all steam generator actions such as inspection, operation, maintenance, modifications, repairs, assessments, R and D, performance monitoring and feedback. This paper discusses OPG steam generator life cycle management experience to date, including successes, failures and how lessons learned have been re-applied. The discussion includes relevant examples from each of the operating stations: Pickering B and Darlington. It also includes some of the experience and lessons learned from the activities carried out to refurbish the steam generators at Pickering A after several years in long term lay-up. The paper is structured along the various degradation modes that have been observed to date at these sites, including monitoring and mitigating actions taken and future plans. (author)

Materials performance in operating PWR steam generators

International Nuclear Information System (INIS)

Weeks, J.R.

1975-01-01

The Inconel-600 tubing in operating PWR steam generators has developed leaks due to intergranular stress corrosion cracking or a general wastage attack, originating from the secondary side of the tubing. Corrosion has been limited to those areas of the steam generators where limited coolant circulation and high heat flux have caused impurities to concentrate. Wastage or pitting attack has always been associated with local concentration of sodium hydrogen phosphates, whereas stress corrosion has been associated with local concentration of sodium or potassium hydroxides. The only instance of stress corrosion originating from the primary side occurred on cold-worked tubing when hydrogen was not added to getter oxygen, and LiOH was not added to raise the pH of the primary coolant. All PWR manufacturers are now recommending that the phosphate treatment of the secondary coolant be abandoned in favor of an all-volatile treatment. Experience in operating plants has shown, however, that removal of phosphate-rich sludge deposits is difficult, and that further wastage and/or intergranular stress corrosion may develop; the residual sodium phosphates gradually convert by reaction with corrosion product hydroxides to sodium hydroxide, which remains concentrated in the limited flow areas. Improvements in circulation patterns have been achieved by inserting flow baffles in some PWR steam generators. Inservice monitoring by eddy current techniques is useful for detecting corrosion-induced defects in the tubing, but irreproducibility in field examinations can lead to uncertainties interpreting the results. (U.S.)

Small leak damage and protection systems in steam generators

International Nuclear Information System (INIS)

Greene, D.A.

1976-01-01

A small leak of water into sodium in a liquid metal heated steam generator can cause damage to adjacent tubes, a phenomenon termed wastage. Theories on this phenomenon range from corrosion from sodium water reaction products to erosion by supersonic particles. An alternative approach considers the water injection to form a simple combustion process. Using this approach many aspects of over 250 wastage experiments can be explained both analytically and physically. The U.S. has an extensive technology in the general area of acoustic surveillance. High temperature in-sodium microphones, in-vessel waveguides, and data analysis techniques have been successfully demonstrated in national development programs. This technology has been applied specifically to the development of an acoustic leak detection/location monitor for small leaks in an operating steam generator

Non-polluting steam generators with fluidized-bed furnaces

Energy Technology Data Exchange (ETDEWEB)

Brandes, H [Deutsche Babcock A.G., Oberhausen (Germany, F.R.)

1979-07-01

The author reports on a 35 MW steam generator with hard coal fluidized-bed furnace a planned 35 MW steam generator with flotation-dirt fluidized-bed furnace, and on planned steam generators for fluidized-bed firing of hard coal up to a steam power of about 200 MW.

CAREM-25 Steam Generator Stability Analysis

International Nuclear Information System (INIS)

Rabiti, A.; Delmastro, D.

2003-01-01

In this work the stability of a once-through CAREM-25 steam generator is analyzed.A fix nodes numerical model, that allows the modelling of the liquid, two-phase and superheated steam zones, is implemented.This model was checked against a mobile finite elements model under saturated steam conditions at the channel exit and a good agreement was obtained.Finally the stability of a CAREM steam generator is studied and the range of in let restrictions that a assure the system stability is analyzed

The impact of NPP Krsko steam generator tube plugging on minimum DNBR at nominal conditions

International Nuclear Information System (INIS)

Lajtman, S.

1996-01-01

Typically, steam generator tube plugging (SGTP) both decreases the reactor coolant system (RCS) flow rate and the heat transfer surface area of the steam generator. At a constant thermal power and vessel outlet temperature, as tube plugging increases, the vessel average temperature, vessel inlet temperature and steam generator secondary side steam pressure decrease. This paper presents the analysis of impact of SGTP on Minimum Departure from Nucleate Boiling Ratio (MDNBR) at NPP Krsko (NEK), using the Improved Thermal Design Procedure (ITDP), WRB-1 correlation, and COBRA-III-C computer code. No credit was given to high plugging percentage region power reduction resulting from turbine volumetric flow limitations. MDNBR is found to be decreasing with increasing plugging, but not under the limiting values. (author)

Steam generators

International Nuclear Information System (INIS)

Hayden, R.L.J.

1979-01-01

Steam generators for nuclear reactors are designed so that deposition of solids on the surface of the inlet side of the tubesheet or the inlet header with the consequent danger of corrosion and eventual tube failure is obviated or substantially reduced. (U.K.)

Failure analysis of retired steam generator tubings

Energy Technology Data Exchange (ETDEWEB)

Kim, Hong Pyo; Kim, J. S.; Hwang, S. S. and others

2005-04-15

Degradation of steam generator leads to forced outage and extension of outage, which causes increase in repair cost, cost of purchasing replacement power and radiation exposure of workers. Steam generator tube rupture incident occurred in Uljin 4 in 2002, which made public sensitive to nuclear power plant. To keep nuclear energy as a main energy source, integrity of steam generator should be demonstrated. Quantitative relationship between ECT(eddy current test) signal and crack size is needed in assesment of integrity of steam generator in pressurized water reactor. However, it is not fully established for application in industry. Retired steam generator of Kori 1 has many kinds of crack such as circumferential and axial primary water stress corrosion crack and outer diameter stress corrosion crack(ODSCC). So, it can be used in qualifying and improving ECT technology and in condition monitoring assesment for crack detected in ISI(in service inspection). In addition, examination of pulled tube of Kori 1 retired steam generator will give information about effectiveness of non welded sleeving technology which was employed to repair defect tubes and remedial action which was applied to mitigate ODSCC. In this project, hardware such as semi hot lab. for pulled tube examination and modification transportation cask for pulled tube and software such as procedure of transportation of radioactive steam generator tube and non-destructive and destructive examination of pulled tube were established. Non-destructive and destructive examination of pulled tubes from Kori 1 retired steam generator were performed in semi hot lab. Remedial actions applied to Kori 1 retired steam generator, PWSCC trend and bulk water chemistry and crevice chemistry in Kori 1 were evaluated. Electrochemical decontamination technology for pulled tube was developed to reduce radiation exposure and enhance effectiveness of pulled tube examination. Multiparameter algorithm developed at ANL, USA was

Darlington steam generator life assurance program

International Nuclear Information System (INIS)

Jelinski, E.; Dymarski, M.; Maruska, C.; Cartar, E.

1995-01-01

The Darlington Nuclear Generating Station belonging to Ontario Hydro is one of the most modern and advanced nuclear generating stations in the world. Four reactor units each generate 881 net MW, enough to provide power to a major city, and representing approximately 20% of the Ontario grid. The nuclear generating capacity in Ontario represents approximately 60% of the grid. In order to look after this major asset, many proactive preventative and predictive maintenance programs are being put in place. The steam generators are a major component in any power plant. World wide experience shows that nuclear steam generators require specialized attention to ensure reliable operation over the station life. This paper describes the Darlington steam generator life assurance program in terms of degradation identification, monitoring and management. The requirements for chemistry control, surveillance of process parameters, surveillance of inspection parameters, and the integration of preventative and predictive maintenance programs such as water lancing, chemical cleaning, RIHT monitoring, and other diagnostics to enhance our understanding of life management issues are identified and discussed. We conclude that we have advanced proactive activities to avoid and to minimize many of the problems affecting other steam generators. An effective steam generator maintenance program must expand the knowledge horizon to understand life limiting processes and to analyze and synthesize observations with theory. (author)