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.)

Highly Flexible and Efficient Solar Steam Generation Device.

Science.gov (United States)

Chen, Chaoji; Li, Yiju; Song, Jianwei; Yang, Zhi; Kuang, Yudi; Hitz, Emily; Jia, Chao; Gong, Amy; Jiang, Feng; Zhu, J Y; Yang, Bao; Xie, Jia; Hu, Liangbing

2017-08-01

Solar steam generation with subsequent steam recondensation has been regarded as one of the most promising techniques to utilize the abundant solar energy and sea water or other unpurified water through water purification, desalination, and distillation. Although tremendous efforts have been dedicated to developing high-efficiency solar steam generation devices, challenges remain in terms of the relatively low efficiency, complicated fabrications, high cost, and inability to scale up. Here, inspired by the water transpiration behavior of trees, the use of carbon nanotube (CNT)-modified flexible wood membrane (F-Wood/CNTs) is demonstrated as a flexible, portable, recyclable, and efficient solar steam generation device for low-cost and scalable solar steam generation applications. Benefitting from the unique structural merits of the F-Wood/CNTs membrane-a black CNT-coated hair-like surface with excellent light absorbability, wood matrix with low thermal conductivity, hierarchical micro- and nanochannels for water pumping and escaping, solar steam generation device based on the F-Wood/CNTs membrane demonstrates a high efficiency of 81% at 10 kW cm -2 , representing one of the highest values ever-reported. The nature-inspired design concept in this study is straightforward and easily scalable, representing one of the most promising solutions for renewable and portable solar energy generation and other related phase-change applications. © 2017 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

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

RELAP5/MOD2 code modifications to obtain better predictions for the once-through steam generator

International Nuclear Information System (INIS)

Blanchat, T.; Hassan, Y.

1989-01-01

The steam generator is a major component in pressurized water reactors. Predicting the response of a steam generator during both steady-state and transient conditions is essential in studying the thermal-hydraulic behavior of a nuclear reactor coolant system. Therefore, many analytical and experimental efforts have been performed to investigate the thermal-hydraulic behavior of the steam generators during operational and accident transients. The objective of this study is to predict the behavior of the secondary side of the once-through steam generator (OTSG) using the RELAP5/MOD2 computer code. Steady-state conditions were predicted with the current version of the RELAP5/MOD2 code and compared with experimental plant data. The code predictions consistently underpredict the degree of superheat. A new interface friction model has been implemented in a modified version of RELAP5/MOD2. This modification, along with changes to the flow regime transition criteria and the heat transfer correlations, correctly predicts the degree of superheat and matches plant data

Mo0.5W0.5S2 for Q-switched pulse generation in ytterbium-doped fiber laser

Science.gov (United States)

Wang, Junli; Chen, Lei; Dou, Chenxi; Yan, Haiting; Meng, Lingjie; Wei, Zhiyi

2018-06-01

In this work, we fabricate the Mo0.5W0.5S2 by microwave-assisted solvothermal method, and report the Q-switched Yb-doped fiber lasers (YDFL) using Mo0.5W0.5S2 polymer film and tapered fiber as the saturable absorbers (SAs). The modulation depth and saturable intensity of the film SA are 5.63% and 6.82 MW cm‑2. The shortest pulse duration and the maximum single pulse energy are 1.22 μs and 148.8 nJ for the film SA, 1.46 μs and 339 nJ for the fiber-taper SA. To the best of our knowledge, this is the first report on the Q-switched YDFL using Mo0.5W0.5S2 SAs.

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

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.

Procedure for generating steam and steam generator for operating said procedure

International Nuclear Information System (INIS)

Chlique, Bernard.

1975-01-01

This invention concerns the generation of steam by bringing the water to be vaporised into indirect thermal exchange relation with the heating steam which condenses when passing in series, along alternate routes, through bundles of tubes immersed in a vaporising chamber. A number of steam generators working on this principle already exist. The purpose of the invention is to modify the operating method of these steam generators by means of a special disposition making it possible to build a compact unit including an additional bundle of tubes heated by the condensates collected at the outlet of each bundle through which the heating steam passes [fr

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)

Vibration Analysis for Steam Dryer of APR1400 Steam Generator

Energy Technology Data Exchange (ETDEWEB)

Han, Sung-heum; Ko, Doyoung [KHNP CRI, Daejeon (Korea, Republic of); Cho, Minki [Doosan Heavy Industry, Changwon (Korea, Republic of)

2016-10-15

This paper is related to comprehensive vibration assessment program for APR1400 steam generator internals. According to U.S. Nuclear Regulatory Commission, Regulatory Guide 1.20 (Rev.3, March 2007), we conducted vibration analysis for a steam dryer as the second steam separator of steam generator internals. The vibration analysis was performed at the 100 % power operating condition as the normal operation condition. The random hydraulic loads were calculated by the computational fluid dynamics and the structural responses were predicted by power spectral density analysis for the probabilistic method. In order to meet the recently revised U.S. NRC RG 1.20 Rev.3, the CVAP against the potential adverse flow effects in APR1400 SG internals should be performed. This study conducted the vibration response analysis for the SG steam dryer as the second moisture separator at the 100% power condition, and evaluated the structural integrity. The predicted alternating stress intensities were evaluated to have more than 17.78 times fatigue margin compared to the endurance limit.

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.)

Steam generator operating experience update, 1982-1983

International Nuclear Information System (INIS)

Frank, L.

1984-06-01

This report is a continuation of earlier reports by the staff addressing pressurized water reactor steam generator operating experience. NUREG-0886, Steam Generator Tube Experience, published in February 1982 summarized experience in domestic and foreign plants through December 1981. This report summarizes steam generator operating experience in domestic plants for the years 1982 and 1983. Included are new problems encountered with secondary-side loose parts, sulfur-induced stress-assisted corrosion cracking, and flow-induced vibrational wear in the new preheater design steam generators. The status of Unresolved Safety Issues A3, A4, and A5 is also discussed

Parametric study for horizontal steam generator modelling

Energy Technology Data Exchange (ETDEWEB)

Ovtcharova, I. [Energoproekt, Sofia (Bulgaria)

1995-12-31

In the presentation some of the calculated results of horizontal steam generator PGV - 440 modelling with RELAP5/Mod3 are described. Two nodalization schemes have been used with different components in the steam dome. A study of parameters variation on the steam generator work and calculated results is made in cases with separator and branch.

Parametric study for horizontal steam generator modelling

Energy Technology Data Exchange (ETDEWEB)

Ovtcharova, I [Energoproekt, Sofia (Bulgaria)

1996-12-31

In the presentation some of the calculated results of horizontal steam generator PGV - 440 modelling with RELAP5/Mod3 are described. Two nodalization schemes have been used with different components in the steam dome. A study of parameters variation on the steam generator work and calculated results is made in cases with separator and branch.

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.

Future development LMFBR-steam generators SNR2

International Nuclear Information System (INIS)

Essebaggers, J.; Pors, J.G.

1975-01-01

The development work for steam generators for large LMFBR plants by Neratoom will be reviewed consisting of: 1. Development engineering information. 2. Concept select studies followed by conceptual designs of selected models. 3. Development manufacturing techniques. 4. Detail design of a prototype unit. 5. Testing of sub-constructions for prototype steam generators. In this presentation item 1 and 2 above will be high lighted, identifying the development work for the SNR-2 steam generators on short term basis. (author)

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

Electrosleeve process for in-situ nuclear steam generator repair

International Nuclear Information System (INIS)

Renaud, E.; Brennenstuhl, A.M.; Stewart, D.R.; Gonzalez, F.

2000-01-01

Degradation of steam generator tubing by localized corrosion is a widespread problem in the nuclear industry that can lead to costly forced outages, unit derating, steam generator replacement or even the permanent shutdown of a reactor. In response to the onset of steam generator degradation at Ontario Power Generation's Pickering Nuclear Generating Station (PNGS) Unit 5, and the determined unsuitability of conventional repair methods (mechanically expanded or welded sleeves) for Alloy 400, an alternative repair technology was developed. Electrosleeve is a non-intrusive, low-temperature process that involves the electrodeposition of a nanocrystalline nickel microalloy forming a continuously bonded, structural layer over the internal diameter of the degraded region. This technology is designed to provide a long-term pressure boundary repair, fully restoring the structural integrity of the damaged region to its original state. This paper describes the Electrosleeve process for steam generator tubing repair and the unique properties of the advanced sleeve material. The successful installation of fourteen Electrosleeves that have been in service for more than six years in Alloy 400 tubing at the Pickering-S CANDU unit, and the more recent (Nov. 99) extension of the technology to Alloy 600 by the installation of 57 sleeves in a U.S. pressurized water reactor (PWR) at Callaway, is presented. The Electrosleeve process has been granted a conditional license by the U.S. Nuclear Regulatory Commission (NRC). In Canada, the process of licensing Electrosleeve with the CNSC / TSSA has begun. (author)

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

Digital simulation for nuclear once-through steam generators

International Nuclear Information System (INIS)

Chen, A.T.

1976-01-01

Mathematical models for calculating the dynamic response of the Oconee type once through steam generator (OTSG) and the integral economizer once through steam generator (IEOTSG) was developed and presented in this dissertation. Linear and nonlinear models of both steam generator types were formulated using the state variable, lumped parameter approach. Transient and frequency responses of system parameters were calculated for various perturbations from both the primary coolant side and the secondary side. Transients of key parameters, including primary outlet temperature, superheated steam outlet temperature, boiling length/subcooled length and steam pressure, were generated, compared and discussed for both steam generator types. Frequency responses of delta P/sub s//deltaT/sub pin/ of the linear OTSG model were validated by using the dynamic testing results obtained at the Oconee I nuclear power station. A sensitivity analysis in both the time and the frequency domains was performed. It was concluded that the mathematical and computer models developed in this dissertation for both the OTSG and the IEOTSG are suitable for overall plant performance evaluation and steam generator related component/system design analysis for nuclear plants using either type of steam generator

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

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.

RELAP/MOD1.5 analysis of steam line break transients for a 3-loop and a 4-loop Westinghouse nuclear steam supply system

International Nuclear Information System (INIS)

Peeler, G.B.; McDonald, T.A.; Kennedy, M.F.

1984-01-01

RELAP/MOD1.5 (Cycle 31 and 34) calculations were made to assess the assumptions used by Westinghouse (W) to analyze mainsteam line break transients. Models of a W 3-loop and 4-loop nuclear steam supply system were used. Sensitivity studies were performed to determine the effect of the availability of offsite power, break size and initial core power. Comparison with W results indicated that if the assumptions used by W are replicated within the RELAP5 framework, then the W methodology for prediction of the Nuclear Steam Supply System (NSSS) response is conservative for steam line break transients

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

Lightweight, Mesoporous, and Highly Absorptive All-Nanofiber Aerogel for Efficient Solar Steam Generation.

Science.gov (United States)

Jiang, Feng; Liu, He; Li, Yiju; Kuang, Yudi; Xu, Xu; Chen, Chaoji; Huang, Hao; Jia, Chao; Zhao, Xinpeng; Hitz, Emily; Zhou, Yubing; Yang, Ronggui; Cui, Lifeng; Hu, Liangbing

2018-01-10

The global fresh water shortage has driven enormous endeavors in seawater desalination and wastewater purification; among these, solar steam generation is effective in extracting fresh water by efficient utilization of naturally abundant solar energy. For solar steam generation, the primary focus is to design new materials that are biodegradable, sustainable, of low cost, and have high solar steam generation efficiency. Here, we designed a bilayer aerogel structure employing naturally abundant cellulose nanofibrils (CNFs) as basic building blocks to achieve sustainability and biodegradability as well as employing a carbon nanotube (CNT) layer for efficient solar utilization with over 97.5% of light absorbance from 300 to 1200 nm wavelength. The ultralow density (0.0096 g/cm 3 ) of the aerogel ensures that minimal material is required, reducing the production cost while at the same time satisfying the water transport and thermal-insulation requirements due to its highly porous structure (99.4% porosity). Owing to its rationally designed structure and thermal-regulation performance, the bilayer CNF-CNT aerogel exhibits a high solar-energy conversion efficiency of 76.3% and 1.11 kg m -2 h -1 at 1 kW m -2 (1 Sun) solar irradiation, comparable or even higher than most of the reported solar steam generation devices. Therefore, the all-nanofiber aerogel presents a new route for designing biodegradable, sustainable, and scalable solar steam generation devices with superb performance.

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

Robust and Low-Cost Flame-Treated Wood for High-Performance Solar Steam Generation.

Science.gov (United States)

Xue, Guobin; Liu, Kang; Chen, Qian; Yang, Peihua; Li, Jia; Ding, Tianpeng; Duan, Jiangjiang; Qi, Bei; Zhou, Jun

2017-05-03

Solar-enabled steam generation has attracted increasing interest in recent years because of its potential applications in power generation, desalination, and wastewater treatment, among others. Recent studies have reported many strategies for promoting the efficiency of steam generation by employing absorbers based on carbon materials or plasmonic metal nanoparticles with well-defined pores. In this work, we report that natural wood can be utilized as an ideal solar absorber after a simple flame treatment. With ultrahigh solar absorbance (∼99%), low thermal conductivity (0.33 W m -1 K -1 ), and good hydrophilicity, the flame-treated wood can localize the solar heating at the evaporation surface and enable a solar-thermal efficiency of ∼72% under a solar intensity of 1 kW m -2 , and it thus represents a renewable, scalable, low-cost, and robust material for solar steam applications.

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)

Identification of leaky steam generators by iodine mapping technique and development of tools for cutting of tubes of steam generators of Indian PHWRS

International Nuclear Information System (INIS)

Subba Rao, D.

2006-01-01

Kakrapar Atomic Power Station (2X220 MWe) located in Mandvi Taluka of Surat District in the state of Gujarat is the fifth Nuclear Power Station of the country. It has got an excellent record in the field of operation, safety, public awareness and emergency preparedness. KAPS Unit -1 achieved first criticality in Sep-1992 and was declared for commercial operation in may-1993. KAPS Unit -2 achieved first criticality in Jan-1995 and was declared for commercial operation in Sep-1995. So far station has generated about 30 billion units.Unit-1 achieved 98.4% and was graded as a world's No.1 in year 2002 amongst all CANDU type reactors. KAPS Unit -1 has made another record of operating continuously for more than 300 days in Indian PHWR s operating history. This paper mainly deals with the Indian PHWRs Steam Generators (SG) tube leaks, leaky steam generator identification by Iodine mapping, and development of special tool for cutting, removal and plugging of leaky tubes. These Steam Generators are designed by M/s Kraft Werke Union (KWU) of Siemens Group, West Germany, and Manufactured by M/s ENSA, SPAIN for Unit- 1 and by M/s MAN-GHH, Germany for Unit- 2. First time in October-2002 one of the Steam Generators of Unit-1 developed tube leak. To identify leaky Steam Generator, KAPS has established a method of Iodine mapping. With that the leaky SG was identified in very short time and corrective actions were taken immediately. Total three tube leaks (two in SG-4 of Unit-1 and one in SG-1 of unit-2) were experienced in both Units'. Following observations were made on SG tubes failure: All failures were in cold leg side; All Failures / deterioration locations were in front of main feed water nozzle; All Failures / deterioration locations were observed to be just above tube support plate (TSP) number 4 or 5; Deterioration ( i.e. wall thinning) observed from OD side and these tubes were adjacent to failed tubes; In all the three incidents, failed / deteriorated tubes were

Response of the steam generator VVER 1000 to a steam line break

International Nuclear Information System (INIS)

Novotny, J.; Novotny, J. Jr.

2003-01-01

Dynamic effects of a steam line break in the weld of the steam pipe and the steam collector on the steam generator system are analyzed. Modelling of a steam line break may concern two cases. The steam line without a restraint and the steam line protected by a whip restraint with viscous elements applied at the postulated break cross-section. The second case is considered. Programme SYSTUS offers a special element the stiffness and viscous damping coefficients of which may be defined as dependent on the relative displacement and velocity of its nodes respectively. A circumferential crack is simulated by a sudden decrease of longitudinal and lateral stiffness coefficients of these special SYSTUS elements to zero. The computation has shown that one can simulate the pipe to behave like completely broken during a time interval of 0,0001 s or less. These elements are used to model the whip restraint with viscous elements and viscous dampers of the GERB type as well. In the case of a whip restraint model the stiffness coefficient-displacement relation and damping coefficient - velocity relation are chosen to fit the given characteristics of the restraint. The special SYSTUS elements are used to constitute Maxwell elements modelling the elasto-plastic and viscous properties of the GERB dampers applied to the steam generator. It has been ascertained that a steam line break at the postulated weld crack between the steam pipe and the steam generator collector cannot endanger the integrity of the system even in a case of the absence of a whip restraint effect. (author)

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.)

Water leak detection in steam generator of Super Phenix

International Nuclear Information System (INIS)

Kong, N.; Brunet, M.; Garnaud, P.; Ghaleb, D.

1990-01-01

With the intent of detecting water leaks inside steam generators, we developed a third system, called acoustic detector, to complement hydrogen detectors and rupture disks (burst disks). The role of the acoustic system is to enable rapid intervention in the event of a leak growing rapidly which could rupture neighbouring tubes. In such a case, the detectable flow rate of the leak varies from a few tens of g/s to a few hundred g/s. At the Super Phenix, three teams work in parallel in complementary frequency bands: EDF (0-20 kHz), CEA/SPCI (20-100 kHz) and CEA/STA (50-300 kHz). The simulation of water leaks in the steam generator by the argon injections performed to date at 50% of the rated power has shown promising results. An anomaly in the evolution of the background noise at more than 50% loading of one of the two instrumented steam generators would make difficult any extrapolation to full power behaviour. 5 refs, 6 figs, 1 tab

Extending service life of steam generators by sleeving tubes

International Nuclear Information System (INIS)

Gutzwiller, J.E.

1982-01-01

Steam generator tubes that are failing due to IGA in the tubesheet crevice can be kept in service by using the basic sealable sleeve design developed by BandW. Variations of the present sleeve design could significantly reduce the number of tubes that must be plugged each year. Sleeving had the potential of keeping 28 percent more tubes in service during 1979. Lowering the overall rate at which tubes are removed from service by plugging will reduce the probability of having to derate the plant or replace the steam generator. Considering tomorrow's replacement power costs, sleeving to keep tubes in service is a practical and sound investment

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)

Mechanical design of a sodium heated steam generator

International Nuclear Information System (INIS)

Chetal, S.C.

1975-01-01

FBTR steam generator is a once through type unit consisting of four 12.5 MW thermal modules generating a total of 74 tons per hour of steam at 125 bar and 480 0 C. This paper outlines the mechanical design of such type of steam generator with emphasis on special design problems associated with this type of sodium to water steam heat exchanger, namely, thermal cycling of transition zone where nucleate boiling changes over to film boiling, application of pressure vessel design criteria for transient pressures, thermal stress evaluation resulting from differential expansion between shell and tube in this typical configuration, sodium headers support design, thermal sleeve, design, thermal shock analysis in thick tubes, thermal stress resulting from stratification and stability of expansion bends against vibration. Some of the possible design changes for the future large size steam generator are outlined. (author)

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

Babcock and Wilcox Canada steam generators past, present and future

Energy Technology Data Exchange (ETDEWEB)

Smith, J.C. [Babcock and Wilcox Canada, Cambridge, Ontario (Canada)

1998-07-01

The steam generators in all of the domestic CANDU Plants, and most of the foreign CANDU plants, were supplied by Babcock and Wilcox Canada, either on their own or in co-operation with local manufacturers. More than 200 steam generators have been supplied. In addition, Babcock and Wilcox Canada has taken the technology which evolved out of the CANDU steam generators and has adapted the technology to supply of replacement steam generators for PWR's. There is enough history and operating experience, plus laboratory experience, to point to the future directions which will be taken in steam generator design. This paper documents the steam generators which have been supplied, the experience in operation and maintenance, what has worked and not worked, and how the design, materials, and operating and maintenance philosophy have evolved. The paper also looks at future requirements in the market, and the continuing research and product development going on at Babcock and Wilcox to address the future steam generator requirements. (author)

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

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

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)

Analysis of steam generator loss-of-feedwater experiments with APROS and RELAP5/MOD3.1 computer codes

Energy Technology Data Exchange (ETDEWEB)

Virtanen, E.; Haapalehto, T. [Lappeenranta Univ. of Technology, Lappeenranta (Finland); Kouhia, J. [VTT Energy, Nuclear Energy, Lappeenranta (Finland)

1995-09-01

Three experiments were conducted to study the behavior of the new horizontal steam generator construction of the PACTEL test facility. In the experiments the secondary side coolant level was reduced stepwise. The experiments were calculated with two computer codes RELAP5/MOD3.1 and APROS version 2.11. A similar nodalization scheme was used for both codes to that the results may be compared. Only the steam generator was modelled and the rest of the facility was given as a boundary condition. The results show that both codes calculate well the behaviour of the primary side of the steam generator. On the secondary side both codes calculate lower steam temperatures in the upper part of the heat exchange tube bundle than was measured in the experiments.

Steam generator materials performance in high temperature gas-cooled reactors

International Nuclear Information System (INIS)

Chafey, J.E.; Roberts, D.I.

1980-11-01

This paper reviews the materials technology aspects of steam generators for HTGRs which feature a graphite-moderated, uranium-thorium, all-ceramic core and utilizes high-pressure helium as the primary coolant. The steam generators are exposed to gas-side temperatures approaching 760 0 C and produce superheated steam at 538 0 C and 16.5 MPa (2400 psi). The prototype Peach Bottom I 40-MW(e) HTGR was operated for 1349 EFPD over 7 years. Examination after decommissioning of the U-tube steam generators and other components showed the steam generators to be in very satisfactory condition. The 330-MW(e) Fort St. Vrain HTGR, now in the final stages of startup, has achieved 70% power and generated more than 1.5 x 10 6 MWh of electricity. The steam generators in this reactor are once-through units of helical configuration, requiring a number of new materials factors including creep-fatigue and water chemistry control. Current designs of larger HTGRs also feature steam generators of helical once-through design. Materials issues that are important in these designs include detailed consideration of time-dependent behavior of both base metals and welds, as required by current American Society of Mechanical Engineers (ASME) Code rules, evaluation of bimetallic weld behavior, evaluation of the properties of large forgings, etc

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.)

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

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

Analysis of steam generator loss-of-feedwater experiments with APROS and RELAP5/MOD3.1 computer codes

International Nuclear Information System (INIS)

Virtanen, E.; Haapalehto, T.; Kouhia, J.

1997-01-01

Three experiments were conducted to study the behaviour of the new horizontal steam generator construction of the PACTEL test facility. In the experiments the secondary side coolant level was reduced stepwise. The experiments were calculated with two computer codes RELAP5/MOD3.1 and APROS version 2.11. A similar nodalization scheme was used for both codes so that the results may be compared. Only the steam generator was modeled and the rest of the facility was given as a boundary condition. The results show that both codes calculate well the behaviour of the primary side of the steam generator. On the secondary side both codes calculate lower steam temperatures in the upper part of the heat exchange tube bundle than was measured in the experiments. (orig.)

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...

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)

Analysis of steam generator behaviour in nuclear power plant with computer program RELAP5; Analiza delovanja uparjalnika jedrske elektrarne s programom RELAP5

Energy Technology Data Exchange (ETDEWEB)

Zeljko, M; Gregoric, M; Peterlin, G [Institut ' Jozef Stefan' , Ljubljana (Yugoslavia)

1983-07-01

Analyses of nuclear power plant behaviour are made with large computer programs. We used RELAP5/MOD1/CYCLE001, which was developed in Idaho National Engineering laboratory. Input model was prepared to analyze transients in steam generator of NPP Krsko. We found out that this version had a lot of faults so we intend to implement a new cycle. First experience shows the whole complexity of such analysis from technical and economical viewpoints. (author)

Coal fired steam generation for heavy oil recovery

International Nuclear Information System (INIS)

Firmin, K.

1992-01-01

In Alberta, some 21,000 m 3 /d of heavy oil and bitumen are produced by in-situ recovery methods involving steam injection. The steam generation requirement is met by standardized natural-gas-fired steam generators. While gas is in plentiful supply in Alberta and therefore competitively priced, significant gas price increases could occur in the future. A 1985 study investigating the alternatives to natural gas as a fuel for steam generation concluded that coal was the most economic alternative, as reserves of subbituminous coal are not only abundant in Alberta but also located relatively close to heavy oil and bitumen production areas. The environmental performance of coal is critical to its acceptance as an alternate fuel to natural gas, and proposed steam generator designs which could burn Alberta coal and control emissions satisfactorily are assessed. Considerations for ash removal, sulfur dioxide sorption, nitrogen oxides control, and particulate emission capture are also presented. A multi-stage slagging type of coal-fired combustor has been developed which is suitable for application with oilfield steam generators and is being commissioned for a demonstration project at the Cold Lake deposit. An economic study showed that the use of coal for steam generation in heavy oil in-situ projects in the Peace River and Cold Lake areas would be economic, compared to natural gas, at fuel price projections and design/cost premises for a project timing in the mid-1990s. 7 figs., 3 tabs

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

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.

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

Chemical-cleaning process evaluation: Westinghouse steam generators. Final report

International Nuclear Information System (INIS)

Cleary, W.F.; Gockley, G.B.

1983-04-01

The Steam Generator Owners Group (SGOG)/Electric Power Research Institute (EPRI) Steam Generator Secondary Side Chemical Cleaning Program, under develpment since 1978, has resulted in a generic process for the removal of accumulated corrosion products and tube deposits in the tube support plate crevices. The SGOG/EPRI Project S150-3 was established to obtain an evaluation of the generic process in regard to its applicability to Westinghouse steam generators. The results of the evaluation form the basis for recommendations for transferring the generic process to a plant specific application and identify chemical cleaning corrosion guidelines for the materials in Westinghouse Steam Generators. The results of the evaluation, recommendations for plant-specific applications and corrosion guidelines for chemical cleaning are presented in this report

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.)

Steam generator materials

International Nuclear Information System (INIS)

Kim, Joung Soo; Han, J. H.; Kim, H. P.; Lim, Y. S.; Lee, D. H.; Suh, J. H.; Hwang, S. S.; Hur, D. H.; Kim, D. J.; Kim, Y. H.

2002-05-01

In order to keep the nuclear power plant(NPP)s safe and increase their operating efficiency, axial stress corrosion cracking(SCC)(IGA/IGSCC, PWSCC, PbSCC) test techniques were developed and SCC property data of the archive steam generator tubing materials having been used in nuclear power plants operating in Korea were produced. The data obtained in this study were data-based, which will be used to clarify the damage mechanisms, to operate the plants safely, and to increase the lifetime of the tubing. In addition, the basic technologies for the improvement of the SCC property of the tubing materials, for new SCC inhibition, for damaged tube repair, and for manufacturing processes of the tubing were developed. In the 1 phase of this long term research, basic SCC test data obtained from the archive steam generator tubing materials used in NPPs operating in Korea were established. These basic technologies developed in the 1 phase will be used in developing process optimization during the 2 phase in order to develop application technologies to the field nuclear power plants

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

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)

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)

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

Fifth CNS international steam generator conference

International Nuclear Information System (INIS)

2006-01-01

The Fifth CNS International Steam Generator Conference was held on November 26-29, 2006 in Toronto, Ontario, Canada. In contrast with other conferences which focus on specific aspects, this conference provided a wide ranging forum on nuclear steam generator technology from life-cycle management to inspection and maintenance, functional and structural performance characteristics to design architecture. The 5th conference has adopted the theme: 'Management of Real-Life Equipment Conditions and Solutions for the Future'. This theme is appropriate at a time of transition in the industry when plants are looking to optimize the performance of existing assets, prevent costly degradation and unavailability, while looking ahead for new steam generator investments in life-extension, replacements and new-build. More than 50 technical papers were presented in sessions that gave an insight to the scope: life management strategies; fouling, cleaning and chemistry; replacement strategies and new build design; materials degradation; condition assessment/fitness for service; inspection advancements and experience; and thermal hydraulic performance

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.

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 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.)

Steam generator secondary pH during a steam generator tube rupture

International Nuclear Information System (INIS)

Adams, J.P.; Peterson, E.S.

1991-12-01

The Nuclear Regulatory Commission requires utilizes to determine the response of a pressurized water reactor to a steam generator tube rupture (SGTR) as part of the safety analysis for the plant. The SGTR analysis includes assumptions regarding the partitioning of iodine between liquid and vapor in steam generator secondary. Experimental studies have determined that the partitioning of iodine in water is very sensitive to the pH. Based on this experimental evidence, the NRC requested the INEL to perform an analytical assessment of secondary coolant system (SCS) pH during an SGTR. Design basis thermal and hydraulic calculations were used together with industry standard chemistry guidelines to determine the SCS chemical concentrations during an SGTR. These were used as input to the Facility for Analysis of Chemical Thermodynamics computer system to calculate the equilibrium pH in the SCS at various discrete time during an SGTR. The results of this analysis indicate that the SCS pH decreases from the initial value of 8.8 to approximately 6.5 by the end of the transient, independent of PWR design

Upgraded Steam Generator Lancing System for Uljin NPP no.2

International Nuclear Information System (INIS)

Kim, Seok Tae; Jeong, Woo Tae; Hong, Sung Yull

2005-01-01

KEPRI(Korea Electric Power Research Institute) has developed various types of steam generator lancing systems since 1998. In this paper, we introduce a new lancing system with new improvements from the previous steam generator lancing system for Uljin NPP #2(nuclear power plant) constructed by KEPRI. The previous lancing system is registered as KALANS R -II and was developed for System-80 type steam generators. The previous lancing system was applied to Uljin unit #3 and it lowered radiation exposure of operators in comparison to manually operated lancing systems. And it effectively removed sludge accumulated around kidney bean zone in the Uljin unit #3 steam generators. But the previous lancing system could only clean partially the steam generators of Uljin unit #4. This was because the rail of the previous lancing system interfered with a part of the steam generator. Therefore we developed a new lancing system that can solve the interference problem. This new lancing system was upgraded from the previous lancing system. Also, a new lancing system for System-80 S/G will be introduced in this paper

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)

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)

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)

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

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.

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

Trip report: United States LMFBR Steam Generator Team. IAEA symposium, Bensberg, Germany, October 14--17, 1974

International Nuclear Information System (INIS)

1974-01-01

Information is presented concerning steam generator design characteristics for the AFR reactor, SNR reactor, PHENIX reactor, SUPER PHENIX reactor, MONJU reactor, and BN-350 reactor; steam generator development programs for West Germany, France, Japan, U. K., and the U. S. S. R.; and the fabrication and inspection of steam generator components. Steam generator performance and maintenance requirements for operating LMFBR reactors are reviewed. (U.S.)

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)

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)

Dependence of steam generator vibrations on feedwater pressure

International Nuclear Information System (INIS)

Sadilek, J.

1989-01-01

Vibration sensors are attached to the bottom of the steam generator jacket between the input and output primary circuit collectors. The effective vibration value is recorded daily. Several times higher vibrations were observed at irregular intervals; their causes were sought, and the relation between the steam generator vibrations measured at the bottom of its vessel and the feedwater pressure was established. The source of the vibrations was found to be in the feedwater tract of the steam generator. The feedwater tract is described and its hydraulic characteristics are given. Vibrations were measured on the S02 valve. It is concluded that vibrations can be eliminated by reducing the water pressure before the control valves and by replacing the control valves with ones with more suitable control characteristics. (E.J.). 3 figs., 1 tab., 3 refs

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)

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

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.

Simulation of a main steam line break with steam generator tube rupture using trace

Energy Technology Data Exchange (ETDEWEB)

Gallardo, S.; Querol, A.; Verdu, G. [Departamento de Ingenieria Quimica Y Nuclear, Universitat Politecnica de Valencia, Camino de Vera s/n, 46022, Valencia (Spain)

2012-07-01

A simulation of the OECD/NEA ROSA-2 Project Test 5 was made with the thermal-hydraulic code TRACE5. Test 5 performed in the Large Scale Test Facility (LSTF) reproduced a Main Steam Line Break (MSLB) with a Steam Generator Tube Rupture (SGTR) in a Pressurized Water Reactor (PWR). The result of these simultaneous breaks is a depressurization in the secondary and primary system in loop B because both systems are connected through the SGTR. Good approximation was obtained between TRACE5 results and experimental data. TRACE5 reproduces qualitatively the phenomena that occur in this transient: primary pressure falls after the break, stagnation of the pressure after the opening of the relief valve of the intact steam generator, the pressure falls after the two openings of the PORV and the recovery of the liquid level in the pressurizer after each closure of the PORV. Furthermore, a sensitivity analysis has been performed to know the effect of varying the High Pressure Injection (HPI) flow rate in both loops on the system pressures evolution. (authors)

Examination of steam generator alloy 800 NG tube from the Almaraz unit 2 NPP

International Nuclear Information System (INIS)

Diego, G. de; Gomez Briceno, D.; Maffiotte, C.; Baladia, M.; Arias, C.J.

2015-01-01

The steam generators of Almaraz Unit 2 were replaced in 1997 by the model 61W/D3 (Siemens) with Alloy 800NG steam generator tubes. Denting indications were firstly detected in 2006 in the SG-3. Crack indications were identified in 2009. At the end of 2011, three tubes were recovered from this steam generator to carry out destructive examination in order to identify the root cause of the tubes degradation. Analysis of deposits point out the existence of multiples elements in the removed OD (Outer Diameter) deposits as well as in the deposits at the free tube under sludge and at the transition zone. Deposits are more abundant at the transition zone than at free tube. About 10% Na concentration has been detected, whereas S and Cl appear in small concentrations. Si appears regularly and Cr, Ni concentrations in the deposits are similar. Multiple intergranular cracks have been detected at 3 mm above the last contact point between the tube and the TS (tube support), in a band of around 5 mm, practically in the whole perimeter of the tube. Fracture surface of crack-B was partially covered by a Si rich layer, whereas fracture surface of crack-A seems to be cleaner. However, no significant differences in composition, except higher amount of S in crack-B, were found in the deposits of both cracks. EDX mapping and Auger profiles point out Ni enrichment with slight Cr enrichment or depletion and Fe depletion. The comparison of Auger profiles with available results for Alloy 800 tested in caustic and acid sulfate environments seems to indicate that the environment inside the cracks detected in the tube R67C48 is neutral or moderately caustic

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

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.

Assessment of PWR Steam Generator modelling in RELAP5/MOD2

International Nuclear Information System (INIS)

Putney, J.M.; Preece, R.J.

1993-06-01

An assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD2 is presented. The assessment is based on a review of code assessment calculations performed in the UK and elsewhere, detailed calculations against a series of commissioning tests carried out on the Wolf Creek PWR and analytical investigations of the phenomena involved in normal and abnormal SG operation. A number of modelling deficiencies are identified and their implications for PWR safety analysis are discussed -- including methods for compensating for the deficiencies through changes to the input deck. Consideration is also given as to whether the deficiencies will still be present in the successor code RELAP5/MOD3

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

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)

Endoscopic laser-induced steam generator: a new method of treatment for early gastric cancer

Science.gov (United States)

Hayashi, Takuya; Arai, Tsunenori; Tajiri, Hisao; Nogami, Yashiroh; Hino, Kunihiko; Kikuchi, Makoto

1996-05-01

The minimum invasive endoscopic treatment for early gastric cancer has been popular in Japan. The endoscopic mucosal resection and laser coagulation by Nd:YAG laser irradiation has been the popular treatment method in this field. However, the submucosal cancer has not been successfully treated by these methods. To treat the submucosal cancer endoscopically, we developed a new coagulation therapy using hot steam generated by Nd:YAG laser. The steam of which temperature was over 10 deg. in Celsius was generated by the laser power of 30 W with 5 ml/min. of saline. The steam was emitted to canine gastric wall under laparotomy or endoscopy for 50 s respectively. Follow up endoscopy was performed on 3, 7, 14, 28 days after the treatment. Histological examination was studied on 7, 28 days, and just after the emission. In the acute observation, the submucosal layer was totally coagulated. On the 7th day, ulceration with white coat was seen. The mucosal defect, submucosal coagulation, and marked edema without muscle degeneration were found by the histological study. On the 14th day, the ulcer advanced in the scar stage. On the 28th day, it completely healed into white scar with mucosal regeneration and mucosal muscle thickening. We could obtain reproducible coagulation up to deep submucosal layer with large area in a short operation time. Moreover there were no degeneration of proper muscle. This treatment effectiveness could be easily controlled by the steam temperature and emission duration. We think that this method can be applied to early gastric cancer including the submucosal cancer, in particular poor risk case for operation. Further study should be done to apply this method to clinical therapy.

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

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

Development of data management system for steam generator inspection

International Nuclear Information System (INIS)

Jung, Yong Moo; Im, Chang Jae; Lee, Yoon Sang; Kang, Soon Joo; An, Jong Kwan

1994-06-01

The data communications environment for transferring Nuclear Power Plant Steam Generator Eddy Current testing data was investigated and after connecting LAN to Hinet-F network, the remote data transfer with the speed of 56 kbps was tested successfully. Data management system for Steam Generator Eddy current testing was also developed by using HP-UX, RMB (Rock Mountain Basic) 21 figs, 13 tabs, 5 refs. (Author)

Development of data management system for steam generator inspection

Energy Technology Data Exchange (ETDEWEB)

Jung, Yong Moo; Im, Chang Jae; Lee, Yoon Sang; Kang, Soon Joo; An, Jong Kwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1994-06-01

The data communications environment for transferring Nuclear Power Plant Steam Generator Eddy Current testing data was investigated and after connecting LAN to Hinet-F network, the remote data transfer with the speed of 56 kbps was tested successfully. Data management system for Steam Generator Eddy current testing was also developed by using HP-UX, RMB (Rock Mountain Basic) 21 figs, 13 tabs, 5 refs. (Author).

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)

Materials choices for the advanced LWR steam generators

International Nuclear Information System (INIS)

Paine, J.P.N.; Shoemaker, C.E.; McIlree, A.R.

1987-01-01

Current light water reactor (LWR) steam generators have been affected by a variety of corrosion and mechanical damage degradation mechanisms. Included are wear caused by tube vibration, intergranular corrosion, pitting, and thinning or wastage of the steam generator tubing and accelerated corrosion of carbon steel supports (denting). The Electric Power Research Institute (EPRI) and the Steam Generator Owners Groups (I, II) have sponsored laboratory and field studies to provide ameliorative actions for the majority of the damage forms experienced to date. Some of the current corrosion mechanisms are aggravated or caused by unique materials choices or materials interactions. New materials have been proposed and at least partially qualified for use in replacement model steam generators, including an advanced LWR design. In so far as possible, the materials choices for the advanced LWR steam generator avoid the corrosion pitfalls seemingly inherent in the current designs. The EPRI Steam Generator Project staff has recommended materials and design choices for a new steam generator. Based on these recommendations we believe that the advanced LWR steam generators will be much less affected by corrosion and mechanical damage mechanisms than are now experienced

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.

CASTOR - Advanced System for VVER Steam Generator Inspection

International Nuclear Information System (INIS)

Mateljak, Petar

2014-01-01

From the safety point of view, steam generator is a very important component of a nuclear power plant. Only a thin tube wall prevents leakage of radioactive material from the primary side into the environment. Therefore, it is very important to perform inspections in order to detect pipe damage and apply appropriate corrective actions during outage. Application of the nondestructive examination (NDE) technique, that can locate degradation and measure its size and orientation, is an integral part of nuclear power plant maintenance. The steam generator inspection system is consisted of remotely controlled manipulator, testing instrument and software for data acquisition and analysis. Recently, the inspection systems have evolved to a much higher level of automation, efficiency and reliability resulting in a lower cost and shorter outage time. Electronic components have become smaller and deal with more complex algorithms. These systems are very fast, precise, reliable and easy to handle. The whole inspection, from the planning, examination, data analysis and final report, is now a highly automated process, which makes inspection much easier and more reliable. This paper presents the new generation of INETEC's VVER steam generator inspection system as ultimate solution for steam generator inspection and repair. (author)

TRAC-PF1/MOD 1 independent assessment: Semiscale MOD-2A feedwater-line break (S-SF-3) and steam-line break (S-SF-5) tests

International Nuclear Information System (INIS)

Dobranich, D.

1985-11-01

The TRAC-PF1/MOD1 independent assessment project at Sandia is part of an overall effort funded by the NRC to determine the ability of various systems codes to predict the detailed thermal/hydraulic response of LWRs during accident and off-normal conditions. As part of this effort, calculations for Semiscale Mod-2A test S-SF-3, a feedwater-line break test, and S-SF-5, a steam-line break test, were performed with TRAC-PF1/MOD1. Most aspects of both the S-SF-3 and S-SF-5 steady-state calculations were found to be in good agreement with data. However, the need for a better steam separator model was identified from the S-SF-3 calculation. Overall, the qualitative behavior of both transients was calculated reasonably well; however, there were some discrepancies in the prediction of the quantitative behavior. The results for the S-SF-3 transient calculation indicate that the primary-to-secondary heat transfer degradation began too early. This was possibly due to overprediction of entrainment in the steam generator boiler, leading to an incorrect calculation of the secondary-side fluid distribution during the steady state. However, there was insufficient data to verify this. Results for the S-SF-5 transient calculation indicate that the primary-side fluid temperature response to a steam-line break was in reasonable agreement with data but the pressure response did not coincide with the data. Uncertainties in the data are sufficient to prevent us from determining the exact cause of this discrepancy, but there is indirect evidence that the calculated rate of phase change in the pressurizer was incorrect. 16 refs., 73 figs., 13 tabs

Experiment data report for semiscale Mod-1 test S-28-2 (steam generator tube rupture test)

International Nuclear Information System (INIS)

Patton, M.L.; Sackett, K.E.

1977-10-01

Recorded test data are presented for Test S-28-2 of the Semiscale Mod-1 steam generator tube rupture test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-28-2 was conducted from initial conditions of 15 936 kPa and 558 K to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. For Test S-28-2, accumulator injection into the intact loop hot leg was provided to simulate simulate the rupture of six steam generator tubes

Strategic maintenance plan for Cernavoda steam generators

International Nuclear Information System (INIS)

Cicerone, T.; Dhar, D.; VandenBerg, J.P.

2002-01-01

Steam generators are among the most important pieces of equipment in a nuclear power plant. They are required full time during the plant operation and obviously no redundancy exists. Past experience has shown that those utilities which implemented comprehensive steam generator inspection and maintenance programs and strict water chemistry controls, have had good steam generator performance that supports good overall plant performance. The purpose of this paper is to discuss a strategic Life Management and Operational-monitoring program for the Cernavoda steam generators. The program is first of all to develop a base of expertise for the management of the steam generator condition; and that is to be supported by a program of actions to be accomplished over time to assess their condition, to take measures to avoid degradation and to provide for inspections, cleaning and modifications as necessary. (author)

Current state and development trends of inspections of steam generator tubes by eddy currents method

International Nuclear Information System (INIS)

Kubis, S.; Herman, M.

1988-01-01

The requirements are presented for steam generator tube testing and the factors are discussed affecting the accuracy of measurement by the eddy currents method. The individual types of the eddy currents method are described, such as the single-frequency method, the multi-frequency method and the pulse eddy currents method. The demands are summed up on manipulator systems whose design is adapted to the steam generator configuration conditions. (E.S.). 5 figs., 16 possu

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

Steam Generator Tube Integrity Program: Surry Steam Generator Project, Hanford site, Richland, Benton County, Washington: Environmental assessment

International Nuclear Information System (INIS)

1980-03-01

The US Nuclear Regulatory Commission (NRC) has placed a Nuclear Regulatory Research Order with the Richland Operations Office of the US Department of Energy (DOE) for expanded investigations at the DOE Pacific Northwest Laboratory (PNL) related to defective pressurized water reactor (PWR) steam generator tubing. This program, the Steam Generator Tube Integrity (SGTI) program, is sponsored by the Metallurgy and Materials Research Branch of the NRC Division of Reactor Safety Research. This research and testing program includes an additional task requiring extensive investigation of a degraded, out-of-service steam generator from a commercial nuclear power plant. This comprehensive testing program on an out-of-service generator will provide NRC with timely and valuable information related to pressurized water reactor primary system integrity and degradation with time. This report presents the environmental assessment of the removal, transport, and testing of the steam generator along with decontamination/decommissioning plans

Steam-treated wood pellets: Environmental and financial implications relative to fossil fuels and conventional pellets for electricity generation

International Nuclear Information System (INIS)

McKechnie, Jon; Saville, Brad; MacLean, Heather L.

2016-01-01

Highlights: • Steam-treated pellets can greatly reduce greenhouse gas emissions relative to coal. • Cost advantage is seen relative to conventional pellets. • Higher pellet cost is more than balanced by reduced retrofit capital requirements. • Low capacity factors further favour steam-treated pellets over conventional pellets. - Abstract: Steam-treated pellets can help to address technical barriers that limit the uptake of pellets as a fuel for electricity generation, but there is limited understanding of the cost and environmental impacts of their production and use. This study investigates life cycle environmental (greenhouse gas (GHG) and air pollutant emissions) and financial implications of electricity generation from steam-treated pellets, including fuel cycle activities (biomass supply, pellet production, and combustion) and retrofit infrastructure to enable 100% pellet firing at a generating station that previously used coal. Models are informed by operating experience of pellet manufacturers and generating stations utilising coal, steam-treated and conventional pellets. Results are compared with conventional pellets and fossil fuels in a case study of electricity generation in northwestern Ontario, Canada. Steam-treated pellet production has similar GHG impacts to conventional pellets as their higher biomass feedstock requirement is balanced by reduced process electricity consumption. GHG reductions of more than 90% relative to coal and ∼85% relative to natural gas (excluding retrofit infrastructure) could be obtained with both pellet options. Pellets can also reduce fuel cycle air pollutant emissions relative to coal by 30% (NOx), 97% (SOx), and 75% (PM 10 ). Lesser retrofit requirements for steam-treated pellets more than compensate for marginally higher pellet production costs, resulting in lower electricity production cost compared to conventional pellets ($0.14/kW h vs. $0.16/kW h). Impacts of retrofit infrastructure become increasingly

The casebook of technical presentation on a steam generator

International Nuclear Information System (INIS)

1986-05-01

This casebook consists of seven presentations, which are measures and experience of maintenance of water quality in PWR generator, corrosion in steam generator, safely evaluation by management and closing in steam generator, testing of eddy current in steam generator, unsettled problems of safety in steam generator and maintenance of water quality in PWR generator.

Review of EPRI's steam generator R and D program

International Nuclear Information System (INIS)

Millett, P.J.; Welty, C.J.

1998-01-01

EPRI has carried out an extensive R and D program on SG technology since the mid 1970's. Very early efforts under the auspices of the Steam Generator Owners Group (SGOG) focused on developing remedial actions for the critical SG corrosion issues of denting, wastage and pitting. Fundamental work was also carried out in the development of thermal hydraulic models for vibration and wear, chemical cleaning and tube repair techniques. In the late 1980's and continuing through today, the program has shifted emphasis towards management of steam generator degradation, primarily stress corrosion cracking of the SG tubes on both the primary and secondary sides. The current Steam Generator Management Program (SGMP) carries out R and D in four areas; materials, chemistry, thermal hydraulics and non-destructive testing. The strategic goals of this program and projects put in place to achieve these goals will be reviewed in detail in this paper. (author)

Water leak detection in steam generator of SUPER PHENIX

International Nuclear Information System (INIS)

Brunet, M.; Garnaud, P.; Ghaleb, D.; Kong, N.

1988-01-01

With the intent of detecting water leaks inside steam generators, we developed a third system, called acoustic detector, to complement hydrogen detectors and rupture disks (burst disks). The role of the acoustic system is to enable rapid intervention in the event of a leak growing rapidly which could rupture neighbouring tubes. In such a case, the detectable flow rate of the leak varies from a few tens of g/s to a few hundred g/s. At the SUPER PHENIX, three teams work in [20-100 kHz] and CEA/STA* [50-300 kHz]. The simulation of water leaks in the steam generator by the argon injections performed to date at 50% of the rated power has shown promising results. An anomaly in the evolution of the background noise at more than 50% loading of one of the two instrumented steam generators would make difficult any extrapolation to full power behaviour. (author)

Characteristics of U-tube assembly design for CANDU 6 type steam generators

International Nuclear Information System (INIS)

Park, Jun Su; Jeong, Seung Ha

1996-06-01

Since the first operation of nuclear steam generator early 1960s, its performance requirements have been met but the steam generator problems have been met but the steam generator problems have been major cause of reducing the operational reliability, plant safety and availability. U-tube assembly of steam generator forms the primary system pressure boundary of the plant and have experienced several types of tube degradation problems. Tube failure and leakage resulting from the degradation will cause radioactive contamination of secondary system by the primary coolant, and this may lead to unplanned plant outages and costly repair operations such as tube plugging or steam generator replacement. For the case of steam generators for heavy water reactors, e.g. Wolsong 2, 3, and 4 NPP, a high cost of heavy water will be imposed additionally. During the plant operation, steam generator tubes can potentially be subject to adverse environmental conditions which will cause damages to U-tube assembly. Types of the damage depend upon the combined effects of design factors, materials and chemical environment of steam generator, and they are the pure water stress corrosion cracking, intergranular attack, pitting, wastage, denting, fretting and fatigue, etc. In this report, a comprehensive review of major design factors of recirculating steam generators has been performed against the potential tube damages. Then the design characteristics of CANDU-type Wolsong steam generator were investigated in detail, including tube material, thermalhydraulic aspects, tube-to-tubesheet joint, tube supports, water chemistry and sludge management. 9 tabs., 18 figs., 38 refs. (Author) .new

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

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)

Coupled RELAP5/PANTHER/COBRA steam line break accident analysis in support of licensing DOEL 2 power uprate and steam generator replacement

International Nuclear Information System (INIS)

Zhang, J.; Bosso, S.; Henno, X.; Ouliddren, K.; Schneidesch, C.R.; Hove, W. van

2004-01-01

The nuclear reactor accident analyses using best estimate codes provide a better understanding and more accurate modeling of the key physical phenomena, which allows a more realistic evaluation of the conservatism and margins in the final safety analysis report (FSAR) accident analysis. The use of the best estimate codes and methods is necessary to meet the increasing technical, licensing and regulatory requirements for major plant changes (e.g. steam generator replacement), power uprate, core design optimization (cycle extension), as well as Periodic Safety Review. Since 1992, Tractebel Engineering (TE) has developed and applied a deterministic bounding approach to FASR accident analysis using the best estimate system thermal hydraulic code RELAP5/MOD2.5 and the subchannel thermal hydraulic code COBRA-3C. This approach has been accepted by the Belgian Safety Authorities, and turned out to be cost effective for most of the non-LOCA transient analyses. Since this approach adapts a decoupled modeling of the core responses, the analysis results often involved too large un-quantified conservatisms, due to either simplistic approximations for asymmetric accidents with strong 3D core neutronics - plant thermal hydraulics interactions, or additional penalties introduced from 'incoherent' initial/boundary conditions for separate plant and core analyses. Therefore, an external dynamic coupling between the RELAP5/MOD2.5 code and the 3-D neutronic code PANTHER was implemented since 1997 via the transient analysis code linkage program TALINK. Furthermore, a static linkage between the PANTHER code and the COBRA-3C code was developed for on-line calculation of (Departure from Nucleate Boiling Ratio (DNBR). TE intends to use the coupled code package for licensing non-symmetric FSAR accident analysis. The TE coupled code package has been applied to develop a main steam line break (MSLB) accident analysis methodology [using the TE deterministic bounding approach. The methodology

Development of a steam generator lancing system

International Nuclear Information System (INIS)

Jeong, Woo-Tae; Kim, Seok-Tae; Hong, Sung-Yull

2006-01-01

It is recommended to clean steam generators of nuclear power plants during plant outages. Under normal operations, sludge is created and constantly accumulates in the steam generators. The constituents of this sludge are different depending on each power plant characteristics. The sludge of the Kori Unit 1 steam generator, for example, was found to be composed of 93% ferrous oxide, 3% carbon and 1% of silica oxide and nickel oxide each. The research to develop a lancing system that would remove sludge deposits from the tubesheet of a steam generator was started in 1998 by the Korea Electric Power Research Institute (KEPRI) of the Korea Electric Power Corporation (KEPCO). The first commercial domestic lancing system in Korea, and KALANS-I Lancing System, was completed in 2000 for Kori Unit 1 for cleaning the tubesheet of its Westinghouse Delta-60 steam generator. Thereafter, the success of the development and site implementation of the KALANS-I lancing system for YGN Units 1 and 2 and Ulchin Units 3 and 4 was also realized in 2004 for sludge removal at those sites. The upper bundle cleaning system for Westinghouse model F steam generators is now under development

Method to detect steam generator tube leakage

International Nuclear Information System (INIS)

Watabe, Kiyomi

1994-01-01

It is important for plant operation to detect minor leakages from the steam generator tube at an early stage, thus, leakage detection has been performed using a condenser air ejector gas monitor and a steam generator blow down monitor, etc. In this study highly-sensitive main steam line monitors have been developed in order to identify leakages in the steam generator more quickly and accurately. The performance of the monitors was verified and the demonstration test at the actual plant was conducted for their intended application to the plants. (author)

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)

Comments on US LMFBR steam generator base technology

International Nuclear Information System (INIS)

Simmons, W.R.

1984-01-01

The development of steam generators for the LMFBR was recognized from the onset by the AEC, now DOE, as a difficult, challenging, and high-priority task. The highly reactive nature of sodium with water/steam requires that the sodium-water/steam boundaries of LMFBR steam generators possess a degree of leak-tightness reliability not normally attempted on a commercial scale. In addition, the LMFBR steam generator is subjected to high fluid temperatures and severe thermal transients. These requirements place great demand on materials, fabrication processes, and inspection methods; and even greater demands on the designer to provide steam generators that can meet these demanding requirements, be fabricated without unreasonable shop requirements, and tolerate off-normal effects

Steam Generator Inspection Planning Expert System

International Nuclear Information System (INIS)

Rzasa, P.

1987-01-01

Applying Artificial Intelligence technology to steam generator non-destructive examination (NDE) can help identify high risk locations in steam generators and can aid in preparing technical specification compliant eddy current test (ECT) programs. A steam Generator Inspection Planning Expert System has been developed which can assist NDE or utility personnel in planning ECT programs. This system represents and processes its information using an object oriented declarative knowledge base, heuristic rules, and symbolic information processing, three artificial intelligence based techniques incorporated in the design. The output of the system is an automated generation of ECT programs. Used in an outage inspection, this system significantly reduced planning time

Model tests of a once-through steam generator for land-blocker assessment and THEDA code verification. Final report

International Nuclear Information System (INIS)

Carter, H.R.; Childerson, M.T.; Moskal, T.E.

1983-06-01

The Babcock and Wilcox Company (B and W) operating Once-Through Steam Generators (OTSGs) have experienced leaking tubes in a region adjacent to the untubed inspection lane. The tube leaks have been attributed to an environmentally-assisted fatigue mechanism with moisture transported up the inspection lane being a major factor in the tube-failure process. B and W has developed a hardware modification (lane blockers) to mitigate the detrimental effects of inspection lane moisture. A 30-tube Laboratory Once-through Steam Generator (Designated OTSGC) was designed, fabricated, and tested. Tests were performed with and without five flat-plate lane blockers installed on tube-support plates (TSPs) 10, 11, 12, 13, and 14. The test results were utilized to determine the effectiveness of lane blockers for eliminating moisture transport to the upper tubesheet in the inspection lanes and to benchmark the predictive capabilities of a three-dimensional steam-generator computer code, THEDA

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.

Automation of steam generator services at public service electric & gas

Energy Technology Data Exchange (ETDEWEB)

Cruickshank, H.; Wray, J.; Scull, D. [Public Service Electric & Gas, Hancock`s Bridge, NJ (United States)

1995-03-01

Public Service Electric & Gas takes an aggressive approach to pursuing new exposure reduction techniques. Evaluation of historic outage exposure shows that over the last eight refueling outages, primary steam generator work has averaged sixty-six (66) person-rem, or, approximately tewenty-five percent (25%) of the general outage exposure at Salem Station. This maintenance evolution represents the largest percentage of exposure for any single activity. Because of this, primary steam generator work represents an excellent opportunity for the development of significant exposure reduction techniques. A study of primary steam generator maintenance activities demonstrated that seventy-five percent (75%) of radiation exposure was due to work activities of the primary steam generator platform, and that development of automated methods for performing these activities was worth pursuing. Existing robotics systems were examined and it was found that a new approach would have to be developed. This resulted in a joint research and development project between Westinghouse and Public Service Electric & Gas to develop an automated system of accomplishing the Health Physics functions on the primary steam generator platform. R.O.M.M.R.S. (Remotely Operated Managed Maintenance Robotics System) was the result of this venture.

3D-Printed, All-in-One Evaporator for High-Efficiency Solar Steam Generation under 1 Sun Illumination.

Science.gov (United States)

Li, Yiju; Gao, Tingting; Yang, Zhi; Chen, Chaoji; Luo, Wei; Song, Jianwei; Hitz, Emily; Jia, Chao; Zhou, Yubing; Liu, Boyang; Yang, Bao; Hu, Liangbing

2017-07-01

Using solar energy to generate steam is a clean and sustainable approach to addressing the issue of water shortage. The current challenge for solar steam generation is to develop easy-to-manufacture and scalable methods which can convert solar irradiation into exploitable thermal energy with high efficiency. Although various material and structure designs have been reported, high efficiency in solar steam generation usually can be achieved only at concentrated solar illumination. For the first time, 3D printing to construct an all-in-one evaporator with a concave structure for high-efficiency solar steam generation under 1 sun illumination is used. The solar-steam-generation device has a high porosity (97.3%) and efficient broadband solar absorption (>97%). The 3D-printed porous evaporator with intrinsic low thermal conductivity enables heat localization and effectively alleviates thermal dissipation to the bulk water. As a result, the 3D-printed evaporator has a high solar steam efficiency of 85.6% under 1 sun illumination (1 kW m -2 ), which is among the best compared with other reported evaporators. The all-in-one structure design using the advanced 3D printing fabrication technique offers a new approach to solar energy harvesting for high-efficiency steam generation. © 2017 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Operating experiences with 1 MW steam generator

Energy Technology Data Exchange (ETDEWEB)

Sano, A; Kanamori, A; Tsuchiya, T

1975-07-01

1 MW steam generator, which was planned as the first stage of steam generator development in Power Reactor and Nuclear Fuel Corp. (PNC) in Japan, is a single-unit, once-through, integrated shell and tube type with multi-helical coil tubes. It was completed in Oarai Engineering Center of PNC in March of 1971, and the various performance tests were carried out up to April, 1972. After the dismantle of the steam generator for structural inspection and material test, it was restored with some improvements. In this second 1 MW steam generator, small leak occurred twice during normal operation. After repairing the failure, the same kind of performance tests as the first steam generator were conducted in order to verify the thermal insulation effect of argon gas in downcomer zone from March to June, 1974. In this paper the above operating experiences were presented including the outline of some performance test results. (author)

Maintenance and repair of LMFBR steam generators

International Nuclear Information System (INIS)

Verriere, P.; Alanche, J.; Minguet, J.L.

1984-06-01

After some general remarks on the French fast neutron system, this paper presents the state of the program for the construction of fast reactor in France. Then, the general design of Super Phenix 1 steam generator components is outlined and, the in-service monitoring systems and protective devices with which they are equiped are briefly described. The methods used, in the event of leakage, for leak location, steam generator inspection, steam generator repair and putting the affected loop back into service, are discussed. There are two main lines of research, relating respectively to the means of water leak detection in sodium and the inspection arrangements that will be used either periodically, or following a sodium-water reaction. Finally, after a brief description of the steam generator, this paper describes the four incidents (leaks) that occurred on the Phenix steam generator in the course of 1982 and 1983, and the subsequent repair operations

Analysis of induced steam generator tube rupture using MAAP 4.0

International Nuclear Information System (INIS)

Kenton, M.; Epstein, M.; Henry, R.E.; Paik, C.; Fuller, E.

1996-01-01

The nuclear industry has initiated a program of Steam Generator Degradation Specific Management (SGDSM) to cope with the various types of corrosion that have been observed in pressurized water reactor (PWR) steam generators. In parallel, the U.S. Nuclear Regulatory Commission is promulgating revised rules on steam generator tube integrity. To support these efforts, the Electric Power Research Institute has sponsored calculations with the MAAP 4 code. The principal objective of these calculations is to estimate the peak temperatures experienced by the steam generator tubes during high-pressure severe accidents. These results are used to evaluate the potential for degraded tubes to leak or rupture. Attention was focused on station blackout (SBO) accidents with loss of turbine-driven auxiliary feedwater because these generally result in the greatest threat to the tubes

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)

US PWR steam generator management: An overview

International Nuclear Information System (INIS)

Welty, C.S. Jr.

1997-01-01

This paper provides an overview on the status of steam generator management activities in US PWRs, and includes: (1) an overview of the impact of steam generator problems; (2) a brief discussion of historical damage trends and the current damage mechanism of most concern; (3) a discussion of the elements of open-quotes steam generator managementclose quotes; and (4) a description of the approach being followed to implement a degradation-specific protocol for tubing inspection and repair. This paper was prepared in conjunction with another paper presented during the Plenary Session of this Conference, open-quotes Steam Generator Degradation: Current Mitigation Strategies for Controlling Corrosionclose quotes, and is provided as a supplement to that material

Eddy Currents Inspection of CANDU Steam Generator Tubes using Zetec's ZR-1 Robot. Experience in Romania

Energy Technology Data Exchange (ETDEWEB)

Scott Hower [Zetec Inc. (Romania); Luiza Vladu; Adrian Nichisov; Mihai Cretu [COMPCONTROL ING. (Romania)

2006-07-01

Full text of publication follows: The commercial operation of Unit 1 of Cernavoda NPP started on 2 December, 1996. The unit's reactor type is PHWR-CANDU 6 (electrical capacity 706 MWe), using natural uranium. The nuclear fuel is manufactured in Romania. The Cernavoda nuclear power plant has four CANDU - design steam generators that have been in service since 1996. The paper introduces the new ZR-1 Robot System for Inspection and Maintenance/Repair from Zetec that combines the newest state-of-the-art robotics technology with Zetec experience - based innovation to address the needs for inspection and repair of steam generators. The multipurpose ZR-1 can be easily installed to perform the necessary eddy current inspection and remain installed ready for follow-up maintenance and repair. It has superior technical performances and a modular three axis motion of arm that enables 100% coverage of tube sheet. Automated, repeatable, and precise positioning of tool heads ensures accurate delivery and reducing costly rework and reduces inspection time by 30%. The modular, light weight, and portable design permits easy assembly and disassembly through small openings and it reduces setup/tear down time by 30%. The first deployment of the new ZR-1 Robot was made in September 2004 at the Cernavoda NPP inspection outage. The unit's reactor type is PHWR-CANDU 6 (electrical capacity 706 MWe), using natural uranium; the nuclear fuel is manufactured in Romania. The Cernavoda nuclear power plant Unit 1 has four CANDU - design steam generators that have been in service since 1996. The paper presents also the Zetec's field experience and customer experience with this system. It describes the equipment setup in Cernavoda's steam generators mock-up, functional tests and calibration. Finally, provides details on the execution of the inspection, options for standardizing the inspection techniques and conclusions. (authors)

Report on US-Japan 1983 meetings on steam generators

International Nuclear Information System (INIS)

1984-04-01

This is a report on a trip to Japan by personnel of the US Nuclear Regulatory Commission in 1983 to exchange information on steam generators of nuclear power plants. Steam generators of Japanese pressurized water reactors have experienced nearly all of the forms of degradation that have been experienced in US recirculating-type steam generators, except for denting and pitting. More tubes have been plugged per year of reactor operation in Japanese than in US steam generators, but much of the Japanese tube plugging is preventative rather than the result of leaks experienced. The number of leaks per reactor year is much smaller for Japanese than for US steam generators. No steam generators have been replaced in Japan while several have been replaced in the US. The Japanese experience may be related to their very stringent inspection and maintenance programs for steam generators

Draining down of a nuclear steam generating system

International Nuclear Information System (INIS)

Jawor, J.C.

1987-01-01

The method is described of draining down contained reactor-coolant water from the inverted vertical U-tubes of a 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 is 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. The method comprises continuously 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 tube sheet 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

System for combustion of sunflower shells in industrial steam generators

International Nuclear Information System (INIS)

Todoriev, Kh.

2000-01-01

The paper presents an economically efficient solution for reconstruction of steam generators with steam production over 5 t/h using foregoing cyclone chamber for sunflower shells combustion. For average fuel caloricity 9 445 ccal/kg and sunflower shells caloricity between 3 485 and 3 750 ccal/kg, the petroleum saving is 68.78% for an average boiler efficiency 4.6 t/h steam

Loss of feed flow, steam generator tube rupture and steam line break thermohydraulic experiments

International Nuclear Information System (INIS)

Mendler, O.J.; Takeuchi, K.; Young, M.Y.

1986-10-01

The Westinghouse Model Boiler No. 2 (MB-2) steam generator test model at the Engineering Test Facility in Tampa, Florida, was reinstrumented and modified for performing a series of tests simulating steam generator accident transients. The transients simulated were: loss of feed flow, steam generator tube rupture, and steam line break events. This document presents a description of (1) the model boiler and the associated test facility, (2) the tests performed, and (3) the analyses of the test results

Loss of feed flow, steam generator tube rupture and steam line break thermohydraulic experiments

Energy Technology Data Exchange (ETDEWEB)

Mendler, O J; Takeuchi, K; Young, M Y

1986-10-01

The Westinghouse Model Boiler No. 2 (MB-2) steam generator test model at the Engineering Test Facility in Tampa, Florida, was reinstrumented and modified for performing a series of tests simulating steam generator accident transients. The transients simulated were: loss of feed flow, steam generator tube rupture, and steam line break events. This document presents a description of (1) the model boiler and the associated test facility, (2) the tests performed, and (3) the analyses of the test results.

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)

Heat transfer characteristics of horizontal steam generators under natural circulation conditions

International Nuclear Information System (INIS)

Hyvaerinen, J.

1996-01-01

This paper deals with the heat transfer characteristics of horizontal steam generators, particularly under natural circulation (decay heat removal) conditions on the primary side. Special emphasis is on the inherent features of horizontal steam generator behaviour. A mathematical model of the horizontal steam generator primary side is developed and qualitative results are obtained analytically. A computer code, called HSG, is developed to solve the model numerically, and its predictions are compared with experimental data. The code is employed to obtain for VVER 440 steam generators quantitative results concerning the dependence of primary-to-secondary heat transfer efficiency on the primary side flow rate, temperature and secondary level. It turns out that the depletion of the secondary inventory leads to an inherent limitation of the decay energy removal in VVER steam generators. The limitation arises as a consequence of the steam generator tube bundle geometry. As an example, it is shown that the grace period associated with pressurizer safety valve opening during a station black-out is 2 1/2-3 hours instead of the 5-6 hours reported in several earlier studies. (However, the change in core heat-up timing is much less-about 1 h at most.) The heat transfer limitation explains the fact that, in the Greifswald VVER 440 station black-out accident in 1975, the steam generators never boiled dry. In addition, the stability of single-phase natural circulation is discussed and insights on the modelling of horizontal steam generators with general-purpose thermal-hydraulic system codes are also presented. (orig.)

Steam generator arrangement

International Nuclear Information System (INIS)

Ssinegurski, E.

1981-01-01

A steam flow path arrangement for covering the walls of the rear gas pass of a steam generator is disclosed. The entire flow passes down the sidewalls with a minor portion then passing up through the rear wall to a superheater inlet header at an intermediate elevation. The major portion of the flow passes up the front wall and through hanger tubes to a roof header. From there the major portion passes across the roof and down the rear wall to the superheater inlet header at the intermediate elevation

PACTEL: Experiments on the behaviour of the new horizontal steam generator

Energy Technology Data Exchange (ETDEWEB)

Kouhia, J.; Riikonen, V.; Purhonen, H. [VTT Energy, Lappeenranta (Finland)

1995-12-31

Experiments were performed to study the behaviour of the PACTEL facility, a medium scale integral test loop simulating VVER 440 pressurized water reactors. The study focused on the operation of the new horizontal steam generator model installed in PACTEL. Three experiments were conducted: a small-break test to observe the steam generator behaviour over a range of primary coolant inventories, a hot leg loop seal experiment to study the cyclic behaviour of a loop seal and a loss of secondary side feedwater test to examine the effect of uncovered tubes in the steam generator. A reverse flow was observed in the lower part of the U-tube bundle of the steam generator during natural circulation. The flow reversal point dropped when the tubes uncovered, during secondary inventory reduction. (orig.). 5 refs.

PACTEL: Experiments on the behaviour of the new horizontal steam generator

Energy Technology Data Exchange (ETDEWEB)

Kouhia, J; Riikonen, V; Purhonen, H [VTT Energy, Lappeenranta (Finland)

1996-12-31

Experiments were performed to study the behaviour of the PACTEL facility, a medium scale integral test loop simulating VVER 440 pressurized water reactors. The study focused on the operation of the new horizontal steam generator model installed in PACTEL. Three experiments were conducted: a small-break test to observe the steam generator behaviour over a range of primary coolant inventories, a hot leg loop seal experiment to study the cyclic behaviour of a loop seal and a loss of secondary side feedwater test to examine the effect of uncovered tubes in the steam generator. A reverse flow was observed in the lower part of the U-tube bundle of the steam generator during natural circulation. The flow reversal point dropped when the tubes uncovered, during secondary inventory reduction. (orig.). 5 refs.

Experiment data report for Semiscale Mod-1 test S-28-3 (steam generator tube rupture test)

International Nuclear Information System (INIS)

Gillins, R.L.; Sackett, K.E.

1977-10-01

Recorded test data are presented for Test S-28-3 of the Semiscale Mod-1 steam generator tube rupture test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-28-3 was conducted from initial conditions of 15621 kPa and 555 K to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. Twelve steam generator tube ruptures were simulated by a controlled injection from a heated accumulator into the intact loop hot leg

Experiment data report for semiscale Mod-1 test S-28-4 (steam generator tube rupture test)

International Nuclear Information System (INIS)

Esparza, V.; Sackett, K.E.

1977-10-01

Recorded test data are presented for Test S-28-4 of the Semiscale Mod-1 steam generator tube rupture test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-28-4 was conducted from initial conditions of 15 646 kPa and 557 K to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. Thirty steam generator tube ruptures were simulated by a controlled injection from a heated accumulator into the intact loop hot leg

Steam Generator tube integrity -- US Nuclear Regulatory Commission perspective

International Nuclear Information System (INIS)

Murphy, E.L.; Sullivan, E.J.

1997-01-01

In the US, the current regulatory framework was developed in the 1970s when general wall thinning was the dominant degradation mechanism; and, as a result of changes in the forms of degradation being observed and improvements in inspection and tube repair technology, the regulatory framework needs to be updated. Operating experience indicates that the current U.S. requirements should be more stringent in some areas, while in other areas they are overly conservative. To date, this situation has been dealt with on a plant-specific basis in the US. However, the NRC staff is now developing a proposed steam generator rule as a generic framework for ensuring that the steam generator tubes are capable of performing their intended safety functions. This paper discusses the current U.S. regulatory framework for assuring steam generator (SG) tube integrity, the need to update this regulatory framework, the objectives of the new proposed rule, the US Nuclear Regulatory Commission (NRC) regulatory guide (RG) that will accompany the rule, how risk considerations affect the development of the new rule, and some outstanding issues relating to the rule that the NRC is still dealing with

CRBRP steam-generator design evolution

International Nuclear Information System (INIS)

Geiger, W.R.; Gillett, J.E.; Lagally, H.O.

1983-01-01

The overall design of the CRBRP Steam Generator is briefly discussed. Two areas of particular concern are highlighted and considerations leading to the final design are detailed. Differential thermal expansion between the shell and the steam tubes is accommodated by the tubes flexing in the curved section of the shell. Support of the tubes by the internals structure is essential to permit free movement and minimize tube wear. Special spacer plate attachment and tube hole geometry promote unimpeded axial movement of the tubes by allowing individual tubes to rotate laterally and by providing lateral movement of the spacer plates relative to the adjacent support structure. The water/steam heads of the CRBRP Steam Generator are spherical heads welded to the lower and upper tubesheets. They were chosen principally because they provide a positively sealed system and result in more favorable stresses in the tubesheets when compared to mechanically attached steamheads

Wavelet network controller for nuclear steam generators

International Nuclear Information System (INIS)

Habibiyan, H; Sayadian, A; Ghafoori-Fard, H

2005-01-01

Poor control of steam generator water level is the main cause of unexpected shutdowns in nuclear power plants. Particularly at low powers, it is a difficult task due to shrink and swell phenomena and flow measurement errors. In addition, the steam generator is a highly complex, nonlinear and time-varying system and its parameters vary with operating conditions. Therefore, it seems that design of a suitable controller is a necessary step to enhance plant availability factor. The purpose of this paper is to design, analyze and evaluate a water level controller for U-tube steam generators using wavelet neural networks. Computer simulations show that the proposed controller improves transient response of steam generator water level and demonstrate its superiority to existing controllers

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

Dynamic modelling of nuclear steam generators

International Nuclear Information System (INIS)

Kerlin, T.W.; Katz, E.M.; Freels, J.; Thakkar, J.

1980-01-01

Moving boundary, nodal models with dynamic energy balances, dynamic mass balances, quasi-static momentum balances, and an equivalent single channel approach have been developed for steam generators used in nuclear power plants. The model for the U-tube recirculation type steam generator is described and comparisons are made of responses from models of different complexity; non-linear versus linear, high-order versus low order, detailed modeling of the control system versus a simple control assumption. The results of dynamic tests on nuclear power systems show that when this steam generator model is included in a system simulation there is good agreement with actual plant performance. (author)

Steam generators of Phenix: Measurement of the hydrogen concentration in sodium for detecting water leaks in the steam generator tubes

International Nuclear Information System (INIS)

Cambillard, E.; Lacroix, A.; Langlois, J.; Viala, J.

1975-01-01

The Phenix secondary circuits are provided with measurement systems of hydrogen concentration in sodium, that allow for the detection of possible water leaks in steam generators and the location of a faulty module. A measurement device consists of : a detector with nickel membranes of 0, 3 mm wall thickness, an ion pump with a 200 l/s flow rate, a quadrupole mass spectrometer and a calibrated hydrogen leak. The temperature correction is made automatically. The main tests carried out on the leak detection systems are reported. Since the first system operation (October 24, 1973), the measurements allowed us to obtain the hydrogen diffusion rates through the steam generator tube walls. (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)

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

French steam generator

International Nuclear Information System (INIS)

Remond, A.

1986-01-01

After recalling the potential damage mode of tubes of steam generator, the author recalls the safety criteria used in France. The improvements and the process of damage prejudice and reparation for tubular bundle are presented [fr

AREVA's toolbox for long-term best performance and reliable operation of nuclear steam generators

International Nuclear Information System (INIS)

Drexler, Andreas; Weiss, Steffen; Caris, Neil; Stiepani, Christoph

2015-01-01

Long-term integrity and high performance of major plant systems and components are of uppermost importance for the successful operation of any power plant. AREVA's experience gathered with water-steam cycle chemistry treatments in more than 40 years yields the conclusion: Accumulation of corrosion products in SGs may result in local overheating and enrichment of impurities up to critical levels. This can lead to several degradation phenomena of the structural materials of the SGs. Therefore, minimization of corrosion product generation and prevention of deposit accumulation is required. The objective of AREVA's asset management program is to support operators by minimizing corrosion damage and performance losses of water-steam cycle systems and components and thereby to maximize the availability and economic performance of the plant. Such asset management program is in principle a closed cycle process. It is based on control, corrective and preventive measures. The objective of control measure is deriving a widespread assessment of the corrosion status of the steam-water cycle which yields to weak points and identifying the best suited corrective and/or preventive measures. In the subsequent steps appropriate measures which improve the current status or counteract on identified issues are identified and applied. Corrective measures, likes mechanical and/or chemical cleaning are targeting the minimization of negative influence on plant performance caused by corrosion in the steam-water cycle. Complementary to corrective measures are preventive ones, like optimization of pH strategy and AREVA's FFA technology could by applied. They are focusing on the origin of corrosion product generation. AREVA is offering a toolbox for long-term best performance and reliable operation of NPPs. (author)

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

Acoustic noises of the BOR-60 reactor steam generators when simulating leaks with argon and steam

International Nuclear Information System (INIS)

Sokolov, V.M.; Golushko, V.V.; Afanas'ev, V.A.; Grebenkin, Yu.P.; Muralev, A.B.

1985-01-01

Background acoustic noises of stea generators in different operational regimes and noises of argon and steam small leads (about 0.1 g/s) are studied to determine the possibility of designing the acoustic system for leak detection in sodium-water steamgenerators. Investigations are carried out at the 30 MW micromodule steam generator being in operation at the BOR-60 reactor as well as at the 20 MW tank type steam generator. Immersed ransduceres made of lithium niobate 6 mm in-diameter and waveguide transducers made of a stainless steel in the form of rods 10 mm in-diameter and 500 mm long are used as acoustic monitors. It is shown that the leak noise is more wide-band than the background noise of the steam generator and both high and low frequencies appear in the spectrum. The use of monitors of different types results in similar conslusions inrelation to the character of background noises and leak signals (spectral density, signal to-noise ratio) in the ase of similar bandroidths of the transduceres. A conclusion is made that the change of operational regimes leads to changes of background noise level, which can be close to the reaction of

Physics of steam generators and visit of Saint-Marcel plant

International Nuclear Information System (INIS)

Gillet, N.; Gloaguen, C.; Holcblat, A.; Borsoi, L.; Adobes, A.; David, F.; Greiner, E.; Pascal-Ribot, S.; Gauchet, J.P.; Mercier, L.; Leomy, F.

2004-01-01

This document gathers the transparencies presented at the 6. technical session of the French nuclear energy society (SFEN) in June 2004. The main topic was the physics of steam generators: 1 - description (G. Paudroux, J.Y. Guena, M. Petit); 2 - thermo-hydraulics (A. Holcblat, F. David, S. Pascal-Ribot); 3 - mechanics (N. Gillet, L. Borsoi, A. Adobes); 4 - monitoring and maintenance means (J.P. Gauchet, L. Mercier, F. Leomy); 5 - replacement (C. Gloaguen, E. Greiner). (J.S.)

LMR steam generator blowdown with RETRAN

International Nuclear Information System (INIS)

Wei, T.Y.C.

1985-01-01

One of the transients being considered in the FSAR Chapter 15 analyses of anticipated LMR transients is the fast blowdown of a steam generator upon inadvertent actuation of the liquid metal/water reaction mitigation system. For the blowdown analysis, a stand-alone steam generator model for the IFR plant was constructed using RETRAN

Review of the research proposal for the steam generator retired from Kori unit 1

Energy Technology Data Exchange (ETDEWEB)

Kim, Joung Soo; Han, Joung Ho; Kim, Hong Pyo; Lim, Yun Soo; Lee, Deok Hyun; Hwang, Seong Sik; Hur, Do Haeng [Korea Atomic Energy Research Institute, Taejeon (Korea)

2002-03-01

The tubes of the steam generator retired form Kori unit 1 have many different kinds of failures, such as denting pitting, wastage, ODSCC, PWSCC.Korea Electric Power Research Institute (KEPRI) submitted a research proposal for the steam generator to the Korea Institute S and T Evaluation and Planning (KSITEP). The KISTEP requested Korea Atomic Energy Research Institute to review the proposal by organizing a committee which should be composed of the specialists of the related domestic research institutes. Opinions of the committee on the objectives, research fields, economic benefit and validity in the research proposal were reviewed and suggested optimal research fields to be fulfilled successfully for the retired steam generator. Also, the rolls for the participants in the research works were allocated, which is critical in order to do the project effectively. 6 figs., 5 tabs. (Author)

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.)

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)

Research program plan: steam generators

International Nuclear Information System (INIS)

Muscara, J.; Serpan, C.Z. Jr.

1985-07-01

This document presents a plan for research in Steam Generators to be performed by the Materials Engineering Branch, MEBR, Division of Engineering Technology, (EDET), Office of Nuclear Regulatory Research. It is one of four plans describing the ongoing research in the corresponding areas of MEBR activity. In order to answer the questions posed, the Steam Generator Program has been organized with the three elements of non-destructive examination; mechanical integrity testing; and corrosion, cleaning and decontamination

Water box for steam generator

International Nuclear Information System (INIS)

Lecomte, Robert; Viaud, Michel.

1975-01-01

This invention relates to a water box for connecting an assembly composed of a vertical steam generator and a vertical pump to the vessel of the nuclear reactor, the assembly forming the primary cooling system of a pressurised water reactor. This invention makes it easy to dismantle the pump on the water box without significant loss of water in the primary cooling system of the reactor and particularly without it being necessary to drain the water contained in the steam generator beforehand. It makes it possible to shorten the time required for dismantling the primary pump in order to service or repair it and makes dismantling safer in that the dismantling does not involve draining the steam generator and therefore the critical storage of a large amount of cooling water that has been in contact with the fuel assemblies of the nuclear reactor core [fr

High-efficiency condenser of steam from a steam-gas mixture

Science.gov (United States)

Milman, O. O.; Krylov, V. S.; Ptakhin, A. V.; Kondratev, A. V.; Yankov, G. G.

2017-12-01

The design of a module for a high-efficiency condenser of steam with a high content (up to 15%) of noncondensable gases (NCGs) with a nearly constant steam-gas mixture (SGM) velocity during the condensation of steam has been developed. This module provides the possibility to estimate the operational efficiency of six condenser zones during the motion of steam from the inlet to the SGM suction point. Some results of the experimental tests of the pilot high-efficiency condenser module are presented. The dependence of the average heat transfer coefficient k¯ on the volumetric NCG concentration v¯ has been derived. It is shown that the high-efficiency condenser module can provide a moderate decrease in k¯ from 4400-4600 to 2600-2800 W/(m2 K) at v¯ ≈ 0.5-9.0%. The heat transfer coefficient distribution over different module zones at a heat duty close to its nominal value has been obtained. From this distribution, it can be seen that the average heat transfer coefficient decreases to 2600 W/(m2 K) at an NCG concentration v¯ = 7.5%, but the first condenser sections ( 1- 3) retain high values of k¯ at a level of no lower than 3200 W/(m2 K), and the last sections operate less well, having k¯ at a level of 1700 W/(m2 K). The dependence of the average heat transfer coefficient on the water velocity in condenser tubes has been obtained at a nearly nominal duty such that the extrapolation of this dependence to the water velocity of 2 m/s may be expected to give k¯ = 5000 W/(m2 K) for relatively pure steam, but an increase in k¯ at v¯ = 8% will be smaller. The effect of the gas removal device characteristic on the operation of the high-efficiency condenser module is described. The design developed for the steam condenser of a gas-turbine plant with a power of 25 MW, a steam flow rate of 40.2 t/h, and a CO2 concentration of up to 12% with consideration for the results of performed studies is presented.

Three-dimensional modeling of nuclear steam generator

International Nuclear Information System (INIS)

Bogdan, Z.; Afgan, N.

1985-01-01

In this paper mathematical model for steady-state simulation of thermodynamic and hydraulic behaviour of U-tube nuclear steam generator is described. The model predicts three-dimensional distribution of temperatures, pressures, steam qualities and velocities in the steam generator secondary loop. In this analysis homogeneous two phase flow model is utilized. Foe purpose of the computer implementation of the mathematical model, a special flow distribution code NUGEN was developed. Calculations are performed with the input data and geometrical characteristics related to the D-4 (westinghouse) model of U-tube nuclear steam generator built in Krsko, operating under 100% load conditions. Results are shown in diagrams giving spatial distribution of pertinent variables in the secondary loop. (author)

Three Steam Generator Replacement Projects in 1995

International Nuclear Information System (INIS)

Holz, R.; Clavier, G.

1996-01-01

Since the companies Siemens AG and Framatome S. A. joined their experience and efforts in the field of steam generator replacements and formed a consortium in 1991, the following projects were performed in 1995: Ringhals 3, Tihange 3 and Asco 1. Further projects will follow in 1996, i. e., Doel 4 and Asco 2. Currently, this European consortium is bidding for the contract to replace the steam generators at the Krsko NPP and hopes to be awarded in 1996. An overview of the way the Consortium Siemens and Framatome approaches SG replacement projects is given based on the projects performed in 1995. Various aspects of project planning, management, licensing, personnel qualification and techniques used on site will be discussed. (author)

Preliminary assessment of PWR Steam Generator modelling in RELAP5/MOD3

International Nuclear Information System (INIS)

Preece, R.J.; Putney, J.M.

1993-07-01

A preliminary assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD3 is presented. The study is based on calculations against a series of steady-state commissioning tests carried out on the Wolf Creek PWR over a range of load conditions. Data from the tests are used to assess the modelling of primary to secondary side heat transfer and, in particular, to examine the effect of reverting to the standard form of the Chen heat transfer correlation in place of the modified form applied in RELAP5/MOD2. Comparisons between the two versions of the code are also used to show how the new interphase drag model in RELAP5/MOD3 affects the calculation of SG liquid inventory and the void fraction profile in the riser

Steam generator tube rupture effects on a LOCA

International Nuclear Information System (INIS)

LaChance, J.L.

1979-01-01

A problem currently experienced in commercial operating pressurized water reactors (PWR) in the United States is the degradation of steam generator tubes. Safety questions have arisen concerning the effect of these degraded tubes rupturing during a postulated loss-of-coolant accident (LOCA). To determine the effect of a small number of tube ruptures on the behavior of a large PWR during a postulated LOCA, a series of computer simulations was performed. The primary concern of the study was to determine whether a small number (10 or less of steam generator tubes rupturing at the beginning surface temperatures. Additional reflood analyses were performed to determine the system behavior when from 10 to 60 tubes rupture at the beginning of core reflood. The FLOOD4 code was selected as being the most applicable code for use in this study after an extensive analysis of the capabilities of existing codes to perform simulations of a LOCA with concurrent steam generator tube ruptures. The results of the study indicate that the rupturing of 10 or less steam generator tubes in any of the steam generators during a 200% cold leg break will not result in a significant increase in the peak cladding temperature. However, because of the vaporization of the steam generator secondary water in the primary side of the steam generator, a significant increase in the core pressure occurs which retards the reflooding process

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)

Mathematical modeling of control system for the experimental steam generator

Science.gov (United States)

Podlasek, Szymon; Lalik, Krzysztof; Filipowicz, Mariusz; Sornek, Krzysztof; Kupski, Robert; Raś, Anita

2016-03-01

A steam generator is an essential unit of each cogeneration system using steam machines. Currently one of the cheapest ways of the steam generation can be application of old steam generators came from army surplus store. They have relatively simple construction and in case of not so exploited units - quite good general conditions, and functionality of mechanical components. By contrast, electrical components and control systems (mostly based on relay automatics) are definitely obsolete. It is not possible to use such units with cooperation of steam bus or with steam engines. In particular, there is no possibility for automatically adjustment of the pressure and the temperature of the generated steam supplying steam engines. Such adjustment is necessary in case of variation of a generator load. The paper is devoted to description of improvement of an exemplary unit together with construction of the measurement-control system based on a PLC. The aim was to enable for communication between the steam generator and controllers of the steam bus and steam engines in order to construction of a complete, fully autonomic and maintenance-free microcogeneration system.

Steam generator tube fitness-for-service guidelines

International Nuclear Information System (INIS)

Gorman, J.A.; Harris, J.E.; Lowenstein, D.B.

1995-07-01

The objectives of this project were to characterize defect mechanisms which could affect the integrity of steam generator tubes, to review and critique state-of-the-art Canadian and international steam generator tube fitness-for-service criteria and guidelines, and to obtain recommendations for criteria that could be used to assess fitness-for service guidelines for steam generator tubes containing defects in Canadian power plant service. Degradation mechanisms, that could affect CANDU steam generator tubes in Canada, have been characterized. The design standards and safety criteria that apply to steam generator tubing in nuclear power plant service in Canada and in Belgium, France, Japan, Spain, Sweden, and the USA have been reviewed and described. The fitness-for-service guidelines used for a variety of specific defect types in Canada and internationally have been evaluated and described in detail in order to highlight the considerations involved in developing such defect specific guidelines. Existing procedures for defect assessment and disposition have been identified, including inspection and examination practices. The approaches used in Canada and in Belgium, France, Japan, Spain, Sweden, and the USA for fitness-for-service guidelines were compared and contrasted for a variety of defect mechanisms. The strengths and weaknesses of the various approaches have been assessed. The report presents recommendations on approaches that may be adopted in the development of fitness-for-service guidelines for use in the dispositioning of steam generator tubing defects in Canada. (author). 175 refs., 2 tabs., 28 figs

Modeling and Simulation of U-tube Steam Generator

Science.gov (United States)

Zhang, Mingming; Fu, Zhongguang; Li, Jinyao; Wang, Mingfei

2018-03-01

The U-tube natural circulation steam generator was mainly researched with modeling and simulation in this article. The research is based on simuworks system simulation software platform. By analyzing the structural characteristics and the operating principle of U-tube steam generator, there are 14 control volumes in the model, including primary side, secondary side, down channel and steam plenum, etc. The model depends completely on conservation laws, and it is applied to make some simulation tests. The results show that the model is capable of simulating properly the dynamic response of U-tube steam generator.

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

Capability evaluation of Eddy current and ultrasonic in-service inspections of steam generator tubes. A status report of PISC III Action 5

International Nuclear Information System (INIS)

Bieth, M.; Birac, C.; Comby, R.

1998-01-01

Document summarizes the PISC III (Programme for the Inspection of Steel Components) report No. 41, full description of the PISC III Action 5 on Steam Generator Tubes Inspection, containing all details and final conclusions which are still to be approved by the PISC III Management Board. The report was prepared by the reference laboratory of PISC under guidance and with continuous contribution of the members of the Data Analysis Group (DAG) of this PISC III. There were several procedures which demonstrated good detection capability of major flaws in typical locations of the steam generator. Conclusions of the exercise indicate that capability demonstration is necessary to qualify in service inspection procedures for steam generator tubes

Steam generator leak detection using acoustic method

International Nuclear Information System (INIS)

Goluchko, V.V.; Sokolov, B.M.; Bulanov, A.N.

1982-05-01

The main requirements to meet by a device for leak detection in sodium - water steam generators are determined. The potentialities of instrumentation designed based on the developed requirements have been tested using a model of a 550 kw steam generator [fr

Mathematical modeling of control system for the experimental steam generator

Directory of Open Access Journals (Sweden)

Podlasek Szymon

2016-01-01

Full Text Available A steam generator is an essential unit of each cogeneration system using steam machines. Currently one of the cheapest ways of the steam generation can be application of old steam generators came from army surplus store. They have relatively simple construction and in case of not so exploited units – quite good general conditions, and functionality of mechanical components. By contrast, electrical components and control systems (mostly based on relay automatics are definitely obsolete. It is not possible to use such units with cooperation of steam bus or with steam engines. In particular, there is no possibility for automatically adjustment of the pressure and the temperature of the generated steam supplying steam engines. Such adjustment is necessary in case of variation of a generator load. The paper is devoted to description of improvement of an exemplary unit together with construction of the measurement-control system based on a PLC. The aim was to enable for communication between the steam generator and controllers of the steam bus and steam engines in order to construction of a complete, fully autonomic and maintenance-free microcogeneration system.

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.

Fiber Fabry-Perot Force Sensor with Small Volume and High Performance for Assessing Fretting Damage of Steam Generator Tubes.

Science.gov (United States)

Huang, Peijian; Wang, Ning; Li, Junying; Zhu, Yong; Zhang, Jie

2017-12-13

Measuring the radial collision force between the steam generator tube (SGT) and the tube support plate (TSP) is essential to assess the fretting damage of the SGT. In order to measure the radial collision force, a novel miniaturized force sensor based on fiber Fabry-Perot (F-P) was designed, and the principle and characteristics of the sensor were analyzed in detail. Then, the F-P force sensor was successfully fabricated and calibrated, and the overall dimensions of the encapsulated fiber F-P sensor were 17 mm × 5 mm × 3 mm (L × W × H). The sensor works well in humid, high pressure (10 MPa), high temperature (350 °C), and vibration (40 kHz) environments. Finally, the F-P force sensors were installed in a 1:1 steam generator test loop, and the radial collision force signals between the SGT and the TSP were obtained. The experiments indicated that the F-P sensor with small volume and high performance could help in assessing the fretting damage of the steam generator tubes.

Consolidated nuclear steam generator

International Nuclear Information System (INIS)

Jabsen, F.S.; Schluderberg, D.C.; Paulson, A.E.

1978-01-01

An improved system of providing power has a unique generating means for nuclear reactors with a number of steam generators in the form of replaceable modular units of the expendable type to attain the optimum in effective and efficient vaporization of fluid during the generating power. The system is most adaptable to undrground power plants and marine usage

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

Experiment data report for Semiscale Mod-1 Test S-28-1 (steam generator tube rupture test series)

International Nuclear Information System (INIS)

Collins, B.L.; Coppin, C.E.; Sackett, K.E.

1977-10-01

Recorded test data are presented for Test S-28-1 of the Semiscale Mod-1 steam generator tube rupture test series. These tests are among several Semiscale Mod-1 experiments conducted to investigate the thermal and hydraulic phenomena accompanying a hypothesized loss-of-coolant accident in a pressurized water reactor (PWR) system. Test S-28-1 was conducted from initial conditions of 15 767 kPa and 557 K to investigate the response of the Semiscale Mod-1 system to a depressurization and reflood transient following a simulated double-ended offset shear of the broken loop cold leg piping. During the test, cooling water was injected into the cold leg of the intact and broken loops to simulate emergency core coolant injection in a PWR. Sixty steam generator tube ruptures were simulated by a controlled injection from a heated accumulator into the intact loop hot leg

Steam generator tube integrity program. Phase I report

International Nuclear Information System (INIS)

Alzheimer, J.M.; Clark, R.A.; Morris, C.J.; Vagins, M.

1979-09-01

The results are presented of the pressure tests performed as part of Phase I of the Steam Generator Tube Integrity (SGTI) program at Battelle Pacific Northwest Laboratory. These tests were performed to establish margin-to-failure predictions for mechanically defected Pressurized Water Reactor (PWR) steam generator tubing under operating and accident conditions. Defect geometries tested were selected because they simulate known or expected defects in PWR steam generators. These defect geometries are Electric Discharge Machining (EDM) slots, elliptical wastage, elliptical wastage plus through-wall slot, uniform thinning, denting, denting plus uniform thinning, and denting plus elliptical wastage. All defects were placed in tubing representative of that currently used in PWR steam generators

Effect of heat transfer tube leak on dynamic characteristic of steam generator

International Nuclear Information System (INIS)

Sun Baozhi; Shi Jianxin; Li Na; Zheng Lusong; Liu Shanghua; Lei Yu

2015-01-01

Taking the steam generator of Daya Bay Nuclear Power Station as the research object, one-dimensional dynamic model of the steam generator based on drift flux theory and leak model of heat transfer tube were established. Steady simulation of steam generator under different conditions was carried out. Based on verifying the drift flux model and leak model of heat transfer tube, the effect of leak location and flow rate under different conditions on steam generator's key parameters was studied. The results show that the drift flux model and leak model can reflect the law of key parameter change accurately such as vapor mass fraction and steam pressure under different leak cases. The variation of the parameters is most apparent when the leak is at the entrance of boiling section and vapor mass fraction varies from 0.261 to 0.163 when leakage accounts for 5% of coolant flow rate. The successful prediction of the effect of heat transfer tube leak on dynamic characteristics of the steam generator based on drift flux theory supplies some references for monitoring and taking precautionary measures to prevent heat transfer tube leak accident. (authors)

Perspective of the Westinghouse steam generator secondary side maintenance approach

Energy Technology Data Exchange (ETDEWEB)

Ramaley, D. [Westinghouse Electric Company LLC, Cranberry Township, Pennsylvania (United States)

2012-07-01

Historically, Westinghouse had developed a set of steam generator secondary maintenance guidelines focused around performing recurring activities each outage without direct regards to the age, deposit loading, operational status, or corrosion status of the steam generator. Through the evolution of steam generator design and steam generator condition data, Westinghouse now uses a proactive assessment and planning approach for utilities. Westinghouse works with utilities to develop steam generator secondary maintenance plans for long term steam generator viability. Westinghouse has developed a portfolio of products to allow utilities to optimize steam generator operability and develop programs aimed at maintaining the steam generator secondary side in a favorable condition for successful long term operation. Judicious use of the means available for program development should allow for corrosion free operation, long term full power operation at optimum thermal efficiency, and leveling of outage expenditures over a long period of time. This paper will review the following required elements for an effective steam generator secondary side strategy: • Assessment: In order to develop an appropriate maintenance strategy, actions must be taken to obtain an accurate picture of the SG secondary side condition. • Forecasting: Using available data predictions are developed for future steam generator conditions and required maintenance actions. • Action: Cost effective engineering and maintenance actions must be completed at the appropriate time as designated by the plan. • Evaluation of Results: Following execution of maintenance tactics, it is necessary to revise strategy and develop technology enhancements as appropriate. (author)

Double wall steam generator tubing

International Nuclear Information System (INIS)

Padden, T.R.; Uber, C.F.

1983-01-01

Double-walled steam generator tubing for the steam generators of a liquid metal cooled fast breeder reactor prevents sliding between the surfaces due to a mechanical interlock. Forces resulting from differential thermal expansion between the outer tube and the inner tube are insufficient in magnitude to cause shearing of base metal. The interlock is formed by jointly drawing the tubing, with the inside wall of the outer tube being already formed with grooves. The drawing causes the outer wall of the inner tube to form corrugations locking with the grooves. (author)

U.S. Advanced Materials Development Program for steam generators

International Nuclear Information System (INIS)

Patriarca, P.; Harkness, S.D.; Duke, J.M.

1975-01-01

The selection of construction materials for LMFBR steam generators is reviewed, presenting the advantages and limitations of 2 1 / 2 Cr-1 Mo steel selected for the Clinch River Breeder Reactor Plant. These limitations indicate that further development of high-strength ferritic steels containing 9 to 12 percent Cr and the high-nickel Alloy 800 could lead to superior materials, and programs to develop these materials have been started. Combustion Engineering has surveyed the experience with the high-strength ferritic steels and prepared ingots of 26 selected compositions. Charpy V-notch tests and metallography have been used to characterize these alloys, and optimum welding rod compositions for these alloys are under development. Westinghouse-Tampa is undertaking a program to gain code acceptance of Alloy 800. A program has been set up to provide the information required for design, justification, and fabrication of reliable components. Progress has been made on characterization, the role of tertiary creep in failure, and the development of welding processes. (U.S.)

Leak suppression at steam generator man-, hand-, and eyeholes

International Nuclear Information System (INIS)

Sylvain, C.; Sutz, P.; Gemma, A.

1988-01-01

Plant unavailability associated with primary and secondary holes is approximately the same as that caused by steam generator tube defects, i.e., 0.5%. Problems encountered with steam generator man-, hand-, and eyeholes during plant operation have led Electricite de France (EdF) and Framatome to improve hole seal design and to develop robots for closing and cleaning them. The data base available in France in this field on some 150 steam generators in 900- and 1300-MW(electric) pressurized water reactors (the equivalent of 300 reactor-yr of operation) has been the base of the developments described in this paper. Incidents occurring in operation primarily concern had-and inspection holes located on the steam generator's secondary side. They include four kinds: (1) leakage detected in operation, requiring forced outages, (2) leakage detected during plant restart after a scheduled shutdown and resulting in a restart delay, (3) pitting of seal mating surfaces, not inducing any leakage but jeopardizing subsequent compliance and requiring difficult and costly repairs, and (4) seizing of screws or bolts. New primary and secondary hole stud tightening and maintenance machines help to improve the efficiency of the in-service closing operations. They provide savings of up to 80% on labor, duration of operations, and exposure

Forming a cohesive steam generator maintenance strategy

International Nuclear Information System (INIS)

Poudroux, G.

1991-01-01

In older nuclear plants, steam generator tube bundles are the most fragile part of the reactor coolant system. Steam generator tubes are subject to numerous types of loading, which can lead to severe degradation (corrosion and wear phenomena). Preventive actions, such as reactor coolant temperature reduction or increasing the plugging limit and their associated analyses, can increase steam generator service life. Beyond these preventive actions, the number of affected tubes and the different locations of the degradations that occur often make repair campaigns necessary. Framatome has developed and qualified a wide range of treatment and repair processes. They enable careful management of the repair campaigns, to avoid reaching the maximum steam generator tube plugging limit, while optimizing the costs. Most of the available repair techniques allow a large number of affected tubes to be treated. Here we look only at those techniques that should be taken into account when defining a maintenance strategy. (author)

Condensate polisher application for PWR steam generator corrosion control

International Nuclear Information System (INIS)

Sawochka, S.G.; Leibovitz, J.; Siegwarth, D.P.; Pearl, W.L.

1981-01-01

The evolution of corrosion attack modes particularly in recirculating U-tube PWR steam generators has dictated a thorough review of the advantages and disadvantages of condensate polishing. Analytical modeling techniques to qualitatively predict crevice chemistry variations resulting from steam generator bulk water variations have allowed valuable insights to be developed. Modeling results complemented by steam generator and laboratory corrosion data will be employed to set condensate demineralizer effluent specifications consistent with control of steam generator corrosion. Laboratory and plant studies are being performed to demonstrate achievability of necessary effluent specifications. (author)

Fracture toughness determination in steam generator tubes

International Nuclear Information System (INIS)

Bergant M; Yawny, A; Perez Ipina, J

2012-01-01

The assessment of the structural integrity of steam generator tubes in nuclear power plants deserved increasing attention in the last years due to the negative impact related to their failures. In this context, elastic plastic fracture mechanics (EPFM) methodology appears as a potential tool for the analysis. The application of EPFM requires, necessarily, knowledge of two aspects, i.e., the driving force estimation in terms of an elastic plastic toughness parameter (e.g., J) and the experimental measurement of the fracture toughness of the material (e.g., the material J-resistance curve). The present work describes the development of a non standardized experimental technique aimed to determine J-resistance curves for steam generator tubes with circumferential through wall cracks. The tubes were made of Incoloy 800 (Ni: 30.0-35.0; Cr: 19.0-23.0; Fe: 35.5 min, % in weight). Due to its austenitic microstructure, this alloy shows very high toughness and is widely used in applications where a good corrosion resistance in aqueous environment or an excellent oxidation resistance in high temperature environment is required. Finally, a procedure for the structural integrity analysis of steam generator tubes with crack-like defects, based on a FAD diagram (Failure Assessment Diagram), is briefly described (author)

Babcock and Wilcox Canada steam generators past, present and future

International Nuclear Information System (INIS)

Smith, J.C.

1998-01-01

The steam generators in all of the domestic CANDU Plants, and most of the foreign CANDU plants, were supplied by Babcock and Wilcox Canada, either on their own or in co-operation with local manufacturers. More than 200 steam generators have been supplied. In addition, Babcock and Wilcox Canada has taken the technology which evolved out of the CANDU steam generators and has adapted the technology to supply of replacement steam generators for PWR's. There is enough history and operating experience, plus laboratory experience, to point to the future directions which will be taken in steam generator design. This paper documents the steam generators which have been supplied, the experience in operation and maintenance, what has worked and not worked, and how the design, materials, and operating and maintenance philosophy have evolved. The paper also looks at future requirements in the market, and the continuing research and product development going on at Babcock and Wilcox to address the future steam generator requirements. (author)

Conceptual design of once-through helical steam generator for SMART

Energy Technology Data Exchange (ETDEWEB)

Kim, Yong Wan; Kim, J. I.; Kim, J. H. [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-12-01

Conceptual design of once-through helical steam generator for the integral reactor SMART is developed. The once-through helical steam generator requires quite different design concepts from the steam generators used in loop type commercial reactors. In this study the design requirements satisfying the operating conditions of the steam generator are derived, and the arrangements and the dimensions of the major parts are determined. By describing the design procedure, the cost of redesign and the costs of developments of similar new steam generators are minimized. The three dimensional models developed make it possible to preview the interferences of the steam generator components and to minimize the possibility of significant design changes in the next design stage by the preliminary strength analysis of the major parts. A methodology for evaluation of flow induced vibration of steam generator tubes has been developed and a preliminary flow induced vibration analysis has been performed. 24 refs., 54 figs., 9 tabs. (Author)

Chemical cleaning - essential for optimal steam generator asset management

International Nuclear Information System (INIS)

Ammann, Franz

2009-01-01

Accumulation of deposits in Steam Generator is intrinsic during the operation of Pressurized Water Reactors. Such depositions lead to reduction of thermal performance, loss of component integrity and, in some cases, to power restrictions. Accordingly, removal of such deposits is an essential part of the asset management program of Steam Generators. Every plant has specific conditions, history and constraints which must be considered when planning and performing a chemical cleaning. Typical points are: -Constitution of the deposits or sludge - Sludge load - Sludge distribution in the steam generator - Existing or expected corrosion problems - Amount and tendency of fouling for waste treatment The strategy for chemical cleaning is developed from these points. The range of chemical cleaning treatments starts with very soft cleanings which can remove approximately 100kg per steam generator and ends with full scale, i.e., hard, cleanings which can remove several thousand kilograms of deposits from a steam generator. Dependent upon the desired goal for the operating plant and the steam generator material condition, the correct cleaning method can be selected. This requires flexible cleaning methods that can be adapted to the individual needs of a plant. Such customizing of chemical cleaning methods is a crucial factor for an optimized asset management program of steam generators in a nuclear power plant

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

Modelling of steam condensation in the primary flow channel of a gas-heated steam generator

International Nuclear Information System (INIS)

Kawamura, H.; Meister, G.

1982-10-01

A new simulation code has been developed for the analysis of steam ingress accidents in high temperatures reactors which evaluates the heat transfer in a steam generator headed by a mixture of helium and water steam. Special emphasis is laid on the analysis of steam condensation in the primary circuit of the steam generator. The code takes wall and bulk condensation into account. A new method is proposed to describe the entrainment of water droplets in the primary gas flow. Some typical results are given. Steam condensation in the primary channel may have a significant effect on temperature distributions. The effect on the heat transferred by the steam generator, however, is found to be not so prominent as might be expected. The reason is discussed. A simplified code will also be described, which gives results with reasonable accuracy within much shorter execution times. This code may be used as a program module in a program simulating the total primary circuit of a high temperature reactor. (orig.) [de

Optimum fuel allocation in parallel steam generator systems

International Nuclear Information System (INIS)

Bollettini, U.; Cangioli, E.; Cerri, G.; Rome Univ. 'La Sapienza'; Trento Univ.

1991-01-01

An optimization procedure was developed to allocate fuels into parallel steam generators. The procedure takes into account the level of performance deterioration connected with the loading history (fossil fuel allocation and maintenance) of each steam generator. The optimization objective function is the system hourly cost, overall steam demand being satisfied. Costs are due to fuel and electric power supply and to plant depreciation and maintenance as well. In order to easily updata the state of each steam generator, particular care was put in the general formulation of the steam production function by adopting a special efficiency-load curve description based on a deterioration scaling parameter. The influence of the characteristic time interval length on the optimum operation result is investigated. A special implementation of the method based on minimum cost paths is suggested

U.S. LMFBR steam generators materials considerations and waterside chemistry issues

Energy Technology Data Exchange (ETDEWEB)

Spalaris, C N

1975-07-01

This report describes the materials and waterside chemistry topics most relevant to the steam generator system for the Clinch River Breeder Reactor Plant. Development programs necessary to support or confirm design and plant operating conditions are summarized, together with selected test results obtained to date. (author)

U.S. LMFBR steam generators materials considerations and waterside chemistry issues

International Nuclear Information System (INIS)

Spalaris, C.N.

1975-01-01

This report describes the materials and waterside chemistry topics most relevant to the steam generator system for the Clinch River Breeder Reactor Plant. Development programs necessary to support or confirm design and plant operating conditions are summarized, together with selected test results obtained to date. (author)

The Creys Malville FBR Super Phenix steam generators

International Nuclear Information System (INIS)

Baque, P.; Zuber, T.; Saur, J.M.; Cambillard, E.

1980-08-01

After briefly recalling the French experience on sodium steam generators, the authors describe the design concepts of the Superphenix units and give their main characteristics. A short summary of the realized R and D program precedes the description of the four 750-MWt steam generators, the fabrication of which is in progress by Creusot-Loire at Chalon sur Saone (France). The studies started for the next French fast breeder reactors and their steam generators are mentioned

Regulation of the water level in the steam generator using modal control

International Nuclear Information System (INIS)

Benoit, Guy.

1981-11-01

The nuclear power reactors type P.W.R. (900 MWe) have three steam generators (S.G.). The problem of the water level in the S.G. is analogous to that for a system with non-minimum phase. This causes a serious trouble for the stability of the regulation, which is actually realized by using the PID regulator. The first part of this study is devoted to construct a mathematical model which represents the S.G. This model is simulated on a digital computer, which order is six. The validity of this model is checked using actual measured signals which have been collected from the BUGEY III power reactor. In the second part, the mathematical representation for simulating the regulation of the level in the S.G. using the modal control is given. The simulation of the actual system is given in the third part. This actual system is composed from the S.G. as well as the PI and PID for regulating the water level. As results from this study, it can be concluded that, the modal control improves the regulation of the water level. The accuracy of the steam flow measurement at low rate is poor. So, the actual regulating system using the measurements has a reduced performance performance. The control modal which is represented in this study overcome this problem [fr

NRC integrated program for the resolution of Unresolved Safety Issues A-3, A-4 and A-5 regarding steam generator tube integrity: Final report

International Nuclear Information System (INIS)

1988-09-01

This report presents the results of the NRC integrated program for the resolution of Unresolved Safety Issues (USIs) A-3, A-4, and A-5 regarding steam generator tube integrity. A generic risk assessment is provided and indicates that risk from steam generator tube rupture (SGTR) events is not a significant contributor to total risk at a given site, nor to the total risk to which the general public is routinely exposed. This finding is considered to be indicative of the effectiveness of licensee programs and regulatory requirements for ensuring steam generator tube integrity in accordance with 10 CFR 50, Appendices A and B. This report also identifies a number of staff-recommended actions that the staff finds can further improve the effectiveness of licensee programs in ensuring the integrity of steam generator tubes and in mitigating the consequences of an SGTR. As part of the integrated program, the staff issued Generic Letter 85-02 encouraging licensees of pressurized water reactors (PWRs) to upgrade their programs, as necessary, to meet the intent of the staff-recommended actions; however, such actions do not constitute NRC requirements. In addition, this report describes a number of ongoing staff actions and studies involving steam generator issues which are being pursued to provide added assurance that risk from SGTR events will continue to be small. 146 refs., 5 figs., 11 tabs

Research perspectives on the evaluation of steam generator tube integrity

International Nuclear Information System (INIS)

Muscara, J.; Diercks, D. R.; Majumdar, S.; Kupperman, D. S.; Bakhtiari, S.; Shack, W. J.

2001-01-01

Industry effects have been largely successful in managing degradation of steam generator tubes due to wastage, pitting, and denting, but fretting, SCC and intergranular attack have proved more difficult to manage. Although steam generator replacements are proceeding there is substantial industry interest in operating with degraded steam generators, and significant numbers of plants will continue to do so. In most cases degradation of steam generator tubing by stress corrosion cracking is still managed by plug or repair on detection, because current NDE techniques for characterization of flaws are not accurate enough to permit continued operation. This paper reviews some of the historical background that underlies current steam generator degradation management strategies and outlines some of the additional research that must be done to provide more effective management of degradation in current generators and provide greater assurance of satisfactory performance in replacement steam generators

Development of the On-line Acoustic Leak Detection Tool for the SFR Steam Generator Protection

International Nuclear Information System (INIS)

Kim, Tae-Joon; Jeong, Ji-Young; Kim, Jong-Man; Kim, Byung-Ho; Kim, Seong-O

2007-01-01

The successful detection of a water/steam into a sodium leak in the SFR SG (steam generator) at an early phase of a leak origin depends on the fast response and sensitivity of a leak detection system. This intention of an acoustic leak detection system is stipulated by a key impossibility of a fast detecting of an intermediate leak by the present nominal systems such as the hydrogen meter. Subject of this study is to introduce the detection performance of an on-line acoustic leak detection tool discriminated by a back-propagation neural network with a preprocessing of the 1/m Octave band analysis, and to introduce the status of an on-line development being developed with the acoustic leak detection tool(S/W) in KAERI. For a performance test, it was used with the acoustic signals for a sodium-water reaction from the injected steam into water experiments in KAERI, the acoustic signals injected from the water into the sodium obtained in IPPE, and the background noise of the PFR superheater

Third international seminar on horizontal steam generators

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-12-31

The Third International Seminar on Horizontal Steam Generators held on October 18-20, 1994 in Lappeenranta, consisted of six sessions dealing with the topics: thermal hydraulic experiments and analyses, primary collector integrity, management of primary-to-secondary leakage accidents, feedwater collector replacement and discussion of VVER-440 steam generator safety issues.

Third international seminar on horizontal steam generators

International Nuclear Information System (INIS)

1995-01-01

The Third International Seminar on Horizontal Steam Generators held on October 18-20, 1994 in Lappeenranta, consisted of six sessions dealing with the topics: thermal hydraulic experiments and analyses, primary collector integrity, management of primary-to-secondary leakage accidents, feedwater collector replacement and discussion of VVER-440 steam generator safety issues

Third international seminar on horizontal steam generators

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-12-31

The Third International Seminar on Horizontal Steam Generators held on October 18-20, 1994 in Lappeenranta, consisted of six sessions dealing with the topics: thermal hydraulic experiments and analyses, primary collector integrity, management of primary-to-secondary leakage accidents, feedwater collector replacement and discussion of VVER-440 steam generator safety issues.

Steam generator thermal-hydraulics

International Nuclear Information System (INIS)

Inch, W.W.; Scott, D.A.; Carver, M.B.

1980-01-01

This paper discusses a code for detailed numerical modelling of steam generator thermal-hydraulics, and describes related experimental programs designed to promote in-depth understanding of three-dimensional two-phase flow. (auth)

Safety significance of steam generator tube degradation mechanisms

Energy Technology Data Exchange (ETDEWEB)

Roussel, G; Mignot, P [AIB-Vincotte Nuclear - AVN, Brussels (Belgium)

1991-07-01

Steam generator (SG) tube bundle is a part of the Reactor Coolant Pressure Boundary (RCPB): this means that its integrity must be maintained. However, operating experience shows various types of tube degradation to occur in the SG tubing, which may lead to SG tube leaks or SG tube ruptures and create a loss of primary system coolant through the SG, therefore providing a direct path to the environment outside the primary containment structure. In this paper, the major types of known SG tube degradations are described and analyzed in order to assess their safety significance with regard to SG tube integrity. In conclusion: The operational reliability and the safety of the PWR steam generator s requires a sufficient knowledge of the degradation mechanisms to determine the amount of degradation that a tube can withstand and the time that it may remain in operation. They also require the availability of inspection techniques to accurately detect and characterize the various degradations. The status of understanding of the major types of degradation summarized in this paper shows and justifies why efforts are being performed to improve the management of the steam generator tube defects.

Development of steam generator manufacturing technology

International Nuclear Information System (INIS)

Grant, J.A.

1979-01-01

In 1968 Babcock and Wilcox (Operations) Ltd., received an order from the CEGB to design, manufacture, install and commission 16 Steam Generators for 2 x 660 Mw (e) Advanced Gas Cooled Reactor Power Station at Hartlepool. This order was followed in 1970 by a similar order for the Heysham Power Station. The design and manufacture of the Steam Generators represented a major advance in technology and the paper discusses the methods by which a manufacturing facility was developed, by the Production Division of Babcock, to produce components to a quality, complexity and accuracy unique in the U.K. commercial boilermaking industry. The discussion includes a brief design background, a description of the Steam Generators and a view of the Production Division background. This is followed by a description of the organisation of the technological development and a consideration of the results. (author)

The testing of a steam-water separating device used for vertical steam generators

International Nuclear Information System (INIS)

Ding Xunshen; Cui Baoyuan; Xue Yunkui; Liu Shixun

1989-01-01

The air-water screening tests of a steam-water separating device used for vertical steam generators at low pressure are introduced. The article puts emphasis on the qualification test of the steam-water separating device at hot conditions in a high temperature and pressure water test rig. The performance of the comprehensive test of the steam-water separating device indicates that the humidity of the steam at the drier exit is much less than the specified amount of 0.25%

Code Assessment of SPACE 2.19 using LSTF Steam Generator Tube Rupture Test

Energy Technology Data Exchange (ETDEWEB)

Kim, Minhee; Kim, Seyun [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

The SPACE is a best estimated two-phase three-field thermal-hydraulic analysis code used to analyze the safety and performance of pressurized water reactors. As a result of the development, the 2.19 version of the code was released through the successive various verification and validation works. The present work is on the line of expanding the work by Kim et al. In this study, results produced by the SPACE 2.19 code were compared with the experimental data from JAERI's LSTF Test Run LSTF SB-SG-06 experiment simulating a Steam Generator Tube Rupture (SGTR) transient. In order to identify the predictability of SPACE 2.19, the LSTF steam generator tube rupture test was simulated. To evaluate the computed results, LSTF SB-SG-06 test data simulating the SGTR and the RELAP5/ MOD3.1 are used. The calculation results indicate that the SPACE 2.19 code predicted well the sequence of events and the major phenomena during the transient, such as the asymmetric loop behavior, reactor coolant system cooldown and heat transfer by natural circulation, the primary and secondary system depressurization by the pressurizer auxiliary spray and the steam dump using the intact loop steam generator relief valve.

Steam generator in the SNR-project

International Nuclear Information System (INIS)

van Westenbrugge, J.K.

1979-01-01

The design philosophy of steam generators for 1300 MWe LMFBR's is presented. The basis for this philosophy is the present experience with the licensing of the SNR-300. This experience is reported. The approach for the steam generators for the 1300 MWe LMFBR is elaborated on, both for accident prevention and damage limitation, for the component itself as well as for the system design. Both Design Base Accident and Hypothetical Accidents are discussed. 8 refs

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

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

A Receding Horizon Controller for the Steam Generator Water Level

International Nuclear Information System (INIS)

Na, Man Gyun; Lee, Yoon Joon

2003-01-01

In this work, the receding horizon control method was used to control the water level of nuclear steam generators and applied to two linear models and also a nonlinear model of steam generators. A receding horizon control method is to solve an optimization problem for finite future steps at current time and to implement the first optimal control input as the current control input. The procedure is then repeated at each subsequent instant. The dynamics of steam generators is very different according to power levels. The receding horizon controller is designed by using a reduced linear steam generator model fixed over a certain power range and applied to a Westinghouse-type (U-tube recirculating type) nuclear steam generator. The proposed controller designed at a fixed power level shows good performance for any other power level within this power range. The steam generator shows actually nonlinear characteristics. Therefore, the proposed algorithm is implemented for a nonlinear model of the nuclear steam generator to verify its real performance and also shows good responses

Proceedings of steam generator sludge deposition in recirculating and once through steam generator upper tube bundle and support plates

International Nuclear Information System (INIS)

Baker, R.L.; Harvego, E.A.

1992-01-01

The development of remedial measures of shot peening have given nuclear utilities viable measures to address primary water stress corrosion cracking to extend steam generator life. The nuclear utility industry is now faced with potential replacement of steam generators in nuclear power plants due to stress corrosion cracking and intergranular attach in crevice locations on the secondary side of steam generators at tube support plates and at the crevice at the top of the tube sheet. Significant work has been done on developing and understanding of the effects of sludge buildup on the corrosion process at these locations. This session was envisioned to provide a forum for the development of an understanding of the mechanisms which control the transport and deposition of sludge on the secondary side of steam generators. It is hoped that this information will aid utilities in monitoring the progression of fouling of these crevices by further knowledge in where to look for the onset of support plate crevice fouling. An understanding of the progression of fouling from upper tube support plates to those lower in the steam generator where higher temperatures cause the corrosion process to initiate first can aid the nuclear utility industry in developing remedial measures for this condition and in providing a forewarning of when to apply such remedial measures

Material choices for the commercial fast reactor steam generators

International Nuclear Information System (INIS)

Willby, C.; Walters, J.

1978-01-01

Experience with fast reactor steam generators has shown them to be critical components in achieving a high availability. This paper presents the designers views on the use of ferritic materials for steam generators and describes the proposed design of the steam generators for the Commercial Fast Reactor (CFR), prototype of which are to be inserted in the Prototype Fast Reactor at Dounreay. (author)

Super-hydrophilic copper sulfide films as light absorbers for efficient solar steam generation under one sun illumination

Science.gov (United States)

Guo, Zhenzhen; Ming, Xin; Wang, Gang; Hou, Baofei; Liu, Xinghang; Mei, Tao; Li, Jinhua; Wang, Jianying; Wang, Xianbao

2018-02-01

Solar steam technology is one of the simplest, most direct and effective ways to harness solar energy through water evaporation. Here, we report the development using super-hydrophilic copper sulfide (CuS) films with double-layer structures as light absorbers for solar steam generation. In the double-layer structure system, a porous mixed cellulose ester (MCE) membrane is used as a supporting layer, which enables water to get into the CuS light absorbers through a capillary action to provide continuous water during solar steam generation. The super-hydrophilic property of the double-layer system (CuS/MCE) leads to a thinner water film close to the air-water interface where the surface temperature is sufficiently high, leading to more efficient evaporation (˜80 ± 2.5%) under one sun illumination. Furthermore, the evaporation efficiencies still keep a steady value after 15 cycles of testing. The super-hydrophilic CuS film is promising for practical application in water purification and evaporation as a light absorption material.

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

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)

Steam generator deposit control program assessment at Comanche Peak

International Nuclear Information System (INIS)

Stevens, J.; Fellers, B.; Orbon, S.

2002-01-01

Comanche Peak has employed a variety of methods to assess the effectiveness of the deposit control program. These include typical methods such as an extensive visual inspection program and detailed corrosion product analysis and trending. In addition, a recently pioneered technique, low frequency eddy current profile analysis (LFEC) has been utilized. LFEC provides a visual mapping of the magnetite deposit profile of the steam generator. Analysis of the LFEC results not only provides general area deposition rates, but can also provide local deposition patterns, which is indicative of steam generator performance. Other techniques utilized include trending of steam pressure, steam generator hideout-return, and flow assisted corrosion (FAC) results. The sum of this information provides a comprehensive assessment of the deposit control program effectiveness and the condition of the steam generator. It also provides important diagnostic and predictive information relative to steam generator life management and mitigative strategies, such as special cleaning procedures. This paper discusses the techniques employed by Comanche Peak Chemistry to monitor the effectiveness of the deposit control program and describes how this information is used in strategic planning. (authors)

Commissioning and maintenance experience on mechanical equipment in steam generators of captive power plant at HWP, Manuguru (Paper No. 5.3)

International Nuclear Information System (INIS)

Bhatnagar, R.; Sinha, Ashok; Mohan Rao, A.C.

1992-01-01

Heavy Water Project (Manuguru) is having a captive power plant to cater to the demands of steam and power for the main plant. During the commissioning and initial run of the steam generators and their auxiliaries, teething/initial problems were encountered in nearly all the equipment of the steam generators. This paper briefly describes some of the major problems faced during the commissioning of the steam generators. (author). 4 figs

Corrosion Processes of the CANDU Steam Generator Materials in the Presence of Silicon Compounds

International Nuclear Information System (INIS)

Lucan, Dumitra; Fulger, Manuela; Velciu, Lucian; Lucan, Georgiana; Jinescu, Gheorghita

2006-01-01

The feedwater that enters the steam generators (SG) under normal operating conditions is extremely pure but, however, it contains low levels (generally in the μg/l concentration range) of impurities such as iron, chloride, sulphate, silicate, etc. When water is converted into steam and exits the steam generator, the non-volatile impurities are left behind. As a result of their concentration, the bulk steam generator water is considerably higher than the one in the feedwater. Nevertheless, the concentrations of corrosive impurities are in general sufficiently low so that the bulk water is not significantly aggressive towards steam generator materials. The impurities and corrosion products existing in the steam generator concentrate in the porous deposits on the steam generator tubesheet. The chemical reactions that take place between the components of concentrated solutions generate an aggressive environment. The presence of this environment and of the tubesheet crevices lead to localized corrosion and thus the same tubes cannot ensure the heat transfer between the fluids of the primary and secondary circuits. Thus, it becomes necessary the understanding of the corrosion process that develops into SG secondary side. The purpose of this paper is the assessment of corrosion behavior of the tubes materials (Incoloy-800) at the normal secondary circuit parameters (temperature = 2600 deg C, pressure = 5.1 MPa). The testing environment was demineralized water containing silicon compounds, at a pH=9.5 regulated with morpholine and cyclohexyl-amine (all volatile treatment - AVT). The paper presents the results of metallographic examinations as well as the results of electrochemical measurements. (authors)

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

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.).

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.)

Nuclear steam generator sludge lance method and apparatus

International Nuclear Information System (INIS)

Shirey, R.A.; Murray, D.E.

1991-01-01

This paper describes a sludge lancing system for removing sludge deposits from an interior region of a steam generator. It comprises: a peripheral fluid injection means for injecting a fluid at a high pressure about a periphery of the steam generator, the peripheral fluid injection means comprising at least one elongated fluid conduit, at least one injection nozzle and a joint positioned at a predetermined point along the elongated fluid conduit for permitting the peripheral fluid injection means to bend to a predetermined angle at the joint within the steam generator; a reciprocable fluid injection means for injecting a fluid at a high pressure toward the sludge deposits and dislodging the sludge deposits; and a supporting means positioned within the interior of the steam generator for supporting the reciprocable fluid injection means throughout the reciprocation of the reciprocable fluid injection means

Hydrogen-based power generation from bioethanol steam reforming

Energy Technology Data Exchange (ETDEWEB)

Tasnadi-Asztalos, Zs., E-mail: tazsolt@chem.ubbcluj.ro; Cormos, C. C., E-mail: cormos@chem.ubbcluj.ro; Agachi, P. S. [Babes-Bolyai University, Faculty of Chemistry and Chemical Engineering, 11 Arany Janos, Postal code: 400028, Cluj-Napoca (Romania)

2015-12-23

This paper is evaluating two power generation concepts based on hydrogen produced from bioethanol steam reforming at industrial scale without and with carbon capture. The power generation from bioethanol conversion is based on two important steps: hydrogen production from bioethanol catalytic steam reforming and electricity generation using a hydrogen-fuelled gas turbine. As carbon capture method to be assessed in hydrogen-based power generation from bioethanol steam reforming, the gas-liquid absorption using methyl-di-ethanol-amine (MDEA) was used. Bioethanol is a renewable energy carrier mainly produced from biomass fermentation. Steam reforming of bioethanol (SRE) provides a promising method for hydrogen and power production from renewable resources. SRE is performed at high temperatures (e.g. 800-900°C) to reduce the reforming by-products (e.g. ethane, ethene). The power generation from hydrogen was done with M701G2 gas turbine (334 MW net power output). Hydrogen was obtained through catalytic steam reforming of bioethanol without and with carbon capture. For the evaluated plant concepts the following key performance indicators were assessed: fuel consumption, gross and net power outputs, net electrical efficiency, ancillary consumptions, carbon capture rate, specific CO{sub 2} emission etc. As the results show, the power generation based on bioethanol conversion has high energy efficiency and low carbon footprint.

Hydrogen-based power generation from bioethanol steam reforming

International Nuclear Information System (INIS)

Tasnadi-Asztalos, Zs.; Cormos, C. C.; Agachi, P. S.

2015-01-01

This paper is evaluating two power generation concepts based on hydrogen produced from bioethanol steam reforming at industrial scale without and with carbon capture. The power generation from bioethanol conversion is based on two important steps: hydrogen production from bioethanol catalytic steam reforming and electricity generation using a hydrogen-fuelled gas turbine. As carbon capture method to be assessed in hydrogen-based power generation from bioethanol steam reforming, the gas-liquid absorption using methyl-di-ethanol-amine (MDEA) was used. Bioethanol is a renewable energy carrier mainly produced from biomass fermentation. Steam reforming of bioethanol (SRE) provides a promising method for hydrogen and power production from renewable resources. SRE is performed at high temperatures (e.g. 800-900°C) to reduce the reforming by-products (e.g. ethane, ethene). The power generation from hydrogen was done with M701G2 gas turbine (334 MW net power output). Hydrogen was obtained through catalytic steam reforming of bioethanol without and with carbon capture. For the evaluated plant concepts the following key performance indicators were assessed: fuel consumption, gross and net power outputs, net electrical efficiency, ancillary consumptions, carbon capture rate, specific CO 2 emission etc. As the results show, the power generation based on bioethanol conversion has high energy efficiency and low carbon footprint

Hydrogen-based power generation from bioethanol steam reforming

Science.gov (United States)

Tasnadi-Asztalos, Zs.; Cormos, C. C.; Agachi, P. S.

2015-12-01

This paper is evaluating two power generation concepts based on hydrogen produced from bioethanol steam reforming at industrial scale without and with carbon capture. The power generation from bioethanol conversion is based on two important steps: hydrogen production from bioethanol catalytic steam reforming and electricity generation using a hydrogen-fuelled gas turbine. As carbon capture method to be assessed in hydrogen-based power generation from bioethanol steam reforming, the gas-liquid absorption using methyl-di-ethanol-amine (MDEA) was used. Bioethanol is a renewable energy carrier mainly produced from biomass fermentation. Steam reforming of bioethanol (SRE) provides a promising method for hydrogen and power production from renewable resources. SRE is performed at high temperatures (e.g. 800-900°C) to reduce the reforming by-products (e.g. ethane, ethene). The power generation from hydrogen was done with M701G2 gas turbine (334 MW net power output). Hydrogen was obtained through catalytic steam reforming of bioethanol without and with carbon capture. For the evaluated plant concepts the following key performance indicators were assessed: fuel consumption, gross and net power outputs, net electrical efficiency, ancillary consumptions, carbon capture rate, specific CO2 emission etc. As the results show, the power generation based on bioethanol conversion has high energy efficiency and low carbon footprint.

Glas generator for the steam gasification of coal with nuclear generated heat

International Nuclear Information System (INIS)

Buchner, G.

1980-01-01

The use of heat from a High Temperature Reactor (HTR) for the steam gasification of coal saves coal, which otherwise is burnt to generate the necessary reaction heat. The gas generator for this process, a horizontal pressure vessel, contains a fluidized bed of coal and steam. An ''immersion-heater'' type of heat exchanger introduces the nuclear generated heat to the process. Some special design problems of this gasifier are presented. Reference is made to the present state of development of the steam gasification process with heat from high temperature reactors. (author)

Chemical cleaning an essential part of steam generator asset management

International Nuclear Information System (INIS)

Amman, Franz

2008-01-01

Chemical Cleaning an essential part of Steam Generator asset management accumulation of deposits is intrinsic for the operation of Steam Generators in PWRs. Such depositions often lead to reduction of thermal performance, loss of component integrity and, in some cases to power restrictions. Accordingly removal of such deposits is an essential part of the asset management of the Steam Generators in a Nuclear Power Plant. Every plant has its individual condition, history and constraints which need to be considered when planning and performing a chemical cleaning. Typical points are: - Sludge load amount and constitution of the deposits - Sludge distribution in the steam generator - Existing or expected corrosion problems - Amount and tendency of fouling for waste treatment Depending on this points the strategy for chemical cleaning shall be evolved. the range of treatment starts with very soft cleanings with a removal of approx 100 kg per steam generator and goes to a full scale cleaning which can remove up to several thousand kilograms of deposits from a steam generator. Depending on the goal to be achieved and the steam generator present an adequate cleaning method shall be selected. This requires flexible and 'customisable' cleaning methods that can be adapted to the individual needs of a plant. Such customizing of chemical cleaning methods is an essential factor for an optimized asset management of the steam generator in a nuclear power plant

1.5 W green light generation by single-pass second harmonic generation of a single-frequency tapered diode laser

DEFF Research Database (Denmark)

Jensen, Ole Bjarlin; Andersen, Peter E.; Sumpf, Bernd

2009-01-01

More than 1.5 W of green light at 531 nm is generated by singlepass second harmonic generation in periodically poled MgO:LiNbO3. The pump laser is a high power tapered laser with a distributed Bragg reflector etched in the ridge section of the laser to provide wavelength selectivity. The output...... power of the single-frequency tapered laser is 9.3 W in continuous wave operation. A conversion efficiency of 18.5 % was achieved in the experiments....

Comparative biochemical analysis after steam pretreatment of lignocellulosic agricultural waste biomass from Williams Cavendish banana plant (Triploid Musa AAA group).

Science.gov (United States)

Kamdem, Irénée; Jacquet, Nicolas; Tiappi, Florian Mathias; Hiligsmann, Serge; Vanderghem, Caroline; Richel, Aurore; Jacques, Philippe; Thonart, Philippe

2015-11-01

The accessibility of fermentable substrates to enzymes is a limiting factor for the efficient bioconversion of agricultural wastes in the context of sustainable development. This paper presents the results of a biochemical analysis performed on six combined morphological parts of Williams Cavendish Lignocellulosic Biomass (WCLB) after steam cracking (SC) and steam explosion (SE) pretreatments. Solid (S) and liquid (L) fractions (Fs) obtained from SC pretreatment performed at 180°C (SLFSC180) and 210°C (SLFSC210) generated, after diluted acid hydrolysis, the highest proportions of neutral sugar (NS) contents, specifically 52.82 ± 3.51 and 49.78 ± 1.39%w/w WCLB dry matter (DM), respectively. The highest proportions of glucose were found in SFSC210 (53.56 ± 1.33%w/w DM) and SFSC180 (44.47 ± 0.00%w/w DM), while the lowest was found in unpretreated WCLB (22.70 ± 0.71%w/w DM). Total NS content assessed in each LF immediately after SC and SE pretreatments was less than 2%w/w of the LF DM, thus revealing minor acid autohydrolysis consequently leading to minor NS production during the steam pretreatment. WCLB subjected to SC at 210 °C (SC210) generated up to 2.7-fold bioaccessible glucan and xylan. SC and SE pretreatments showed potential for the deconstruction of WCLB (delignification, depolymerization, decrystallization and deacetylation), enhancing its enzymatic hydrolysis. The concentrations of enzymatic inhibitors, such as 2-furfuraldehyde and 5-(hydroxymethyl)furfural from LFSC210, were the highest (41 and 21 µg ml(-1), respectively). This study shows that steam pretreatments in general and SC210 in particular are required for efficient bioconversion of WCLB. Yet, biotransformation through biochemical processes (e.g., anaerobic digestion) must be performed to assess the efficiency of these pretreatments. © The Author(s) 2015.

Nuclear steam generator tubesheet shield

International Nuclear Information System (INIS)

Nickerson, J.H.D.; Ruhe, A.

1982-01-01

The invention involves improvements to a nuclear steam generator of the type in which a plurality of U-shaped tubes are connected at opposite ends to a tubesheet and extend between inlet and outlet chambers, with the steam generator including an integral preheater zone adjacent to the downflow legs of the U-shaped tubes. The improvement is a thermal shield disposed adjacent to an upper face of the tubesheet within the preheater zone, the shield including ductile cladding material applied directly to the upper face of the tubesheet, with the downflow legs of the U-shaped tubes extending through the cladding into the tubesheet

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

Steam generator replacement at Surry Power Station

International Nuclear Information System (INIS)

McKay, H.S.

1982-01-01

The purposes of the steam generator repair program at Surry Power Station were to repair the tube degradation caused by corrosion-related phenomena and to restore the integrity of the steam generators to a level equivalent to new equipment. The repair program consisted of (1) replacing the existing lower-shell assemblies with new ones and (2) adding new moisture separation equipment to the upper-shell assemblies. These tasks required that several pieces of reactor coolant piping, feedwater piping, main steam piping, and the steam generator be cut and refurbished for reinstallation after the new lower shell was in place. The safety implications and other potential effects of the repair program both during the repair work and after the unit was returned to power were part of the design basis of the repair program. The repair program has been completed on Unit 2 without any adverse effects on the health and safety of the general public or to the personnel engaged in the repair work. Before the Unit 1 repair program began, a review of work procedures and field changes for the Unit 2 repair was conducted. Several major changes were made to avoid recurrence of problems and to streamline procedures. Steam generator replacements was completed on June 1, 1981, and the unit is presently in the startup phase of the outrage

Steam generator life management

International Nuclear Information System (INIS)

King, P.; McGillivray, R.; Reinhardt, W.; Millman, J.; King, B.; Schneider, W.

2003-01-01

'Full-Text:' Steam Generator Life Management responsibility embodies doing whatever is necessary to maintain the steam generation equipment of a nuclear plant in effective, reliable service. All comes together in that most critical deliverable, namely the submission of the documentation which wins approval for return to service after an outage program. Life management must address all aspects of SG reliability over the life of the plant. Nevertheless, the life management activities leading up to return to service approval is where all of it converges. Steam Generator Life Management activities entail four types of work, all equally important in supporting the objective of successful operation. These activities are i) engineering functions; including identification of inspection and maintenance requirements, outage planning and scope definition plus engineering assessment, design and analysis as necessary to support equipment operation, ii) fitness of service work; including the expert evaluation of degradation mechanisms, disposition of defects for return to service or not, and the fitness for service analysis as required to justify ongoing operation with acceptable defects, iii) inspection work; including large scale eddy current inspection of tubing, the definition of defect size and character, code inspections of pressure vessel integrity and visual inspections for integrity and iv) maintenance work; including repairs, retrofits, cleaning and modifications, all as necessary to implement the measures defined during activities i) through iii). The paper discusses the approach and execution of the program for the achievement of the above objectives and particularly of items i) and ii). (author)

Emergency makeup of nuclear steam generators in blackout conditions

International Nuclear Information System (INIS)

Korolev, A.V.; Derevyanko, O.V.

2014-01-01

The paper describes an original solution for using steam energy to organize makeup of NPP steam generators in blackout conditions. The proposed solution combines a disk friction turbine and an axial turbine in a single housing to provide a high overall technical effect enabling the replenishment of nuclear steam generators with steam using the pump turbine drive assembly. The application of the design is analyzed and its efficiency and feasibility are shown

Evaluation of Oconee steam-generator debris. Final report

International Nuclear Information System (INIS)

Rigdon, M.A.; Rubright, M.M.; Sarver, L.W.

1981-10-01

Pieces of debris were observed near damaged tubes at the 14th support plate elevation in the Oconee 1-B steam generator. A project was initiated to evaluate the physical and chemical nature of the debris, to identify its source, and to determine its role in tube damage at this elevation. Various laboratory techniques were used to characterize several debris and mill scale samples. Data from these samples were then compared with each other and with literature data. It was concluded that seven of eight debris samples were probably formed in the steam generator. Six of these samples were probably formed by high temperature aqueous corrosion early in the life of the steam generator. The seventh sample was probably formed by the deposition and spalling of magnetite on the Inconel steam generator tubes. None of the debris samples resembled any of the mill scale samples

Steam generators under construction for the SNR-300 power plant

Energy Technology Data Exchange (ETDEWEB)

Essebaggers, J

1975-07-01

The prototype straight tube and the helical coil-steam generator has been designed and fabricated of which the straight tube steam generator has been successfully tested for over 3000 hours at prototypical conditions and is presently being dismantled for detailed examination of critical designed features. The prototype helical coil steam generator is presently under testing in the 50 MWt test facility at TNO-Hengelo with approximately 500 hours of operation at full load conditions. In an earlier presentation the design and fabrication of the prototype steam generators have been presented, while for this presentation the production units for SNR-300 will be discussed. Some preliminary information will be presented at this meeting of the dismantling operations of the prototype straight tube steam generator. (author)

Steam generators under construction for the SNR-300 power plant

International Nuclear Information System (INIS)

Essebaggers, J.

1975-01-01

The prototype straight tube and the helical coil-steam generator has been designed and fabricated of which the straight tube steam generator has been successfully tested for over 3000 hours at prototypical conditions and is presently being dismantled for detailed examination of critical designed features. The prototype helical coil steam generator is presently under testing in the 50 MWt test facility at TNO-Hengelo with approximately 500 hours of operation at full load conditions. In an earlier presentation the design and fabrication of the prototype steam generators have been presented, while for this presentation the production units for SNR-300 will be discussed. Some preliminary information will be presented at this meeting of the dismantling operations of the prototype straight tube steam generator. (author)

Control system for fluid heated steam generator

Science.gov (United States)

Boland, J.F.; Koenig, J.F.

1984-05-29

A control system for controlling the location of the nucleate-boiling region in a fluid heated steam generator comprises means for measuring the temperature gradient (change in temperature per unit length) of the heating fluid along the steam generator; means for determining a control variable in accordance with a predetermined function of temperature gradients and for generating a control signal in response thereto; and means for adjusting the feedwater flow rate in accordance with the control signal.

Nuclear steam generator tube to tubesheet joint optimization

International Nuclear Information System (INIS)

McGregor, Rod

1999-01-01

Industry-wide problems with Stress Corrosion Cracking in the Nuclear Steam Generator tube-to-tubesheet joint have led to costly repairs, plugging, and replacement of entire vessels. To improve corrosion resistance, new and replacement Steam Generator developments typically employ the hydraulic tube expansion process (full depth) to minimize tensile residual stresses and cold work at the critical transition zone between the expanded and unexpanded tube. These variables have undergone detailed study using specialized X-ray diffraction and analytical techniques. Responding to increased demands from Nuclear Steam Generator operators and manufacturers to credit the leak-tightness and strength contributions of the hydraulic expansion, various experimental tasks with complimentary analytical modelling were applied to improve understanding and control of tube to hole contact pressure. With careful consideration to residual stress impact, design for strength/leak tightness optimization addresses: Experimentally determined minimum contact pressure levels necessary to preclude incipient leakage into the tube/hole interface. The degradation of contact pressure at surrounding expansions caused by the sequential expansion process. The transient and permanent contact pressure variation associated with tubesheet hole dilation during Steam Generator operation. An experimental/analytical simulation has been developed to reproduce cyclic Steam Generator operating strains on the tubesheet and expanded joint. Leak tightness and pullout tests were performed during and following simulated Steam Generator operating transients. The overall development has provided a comprehensive understanding of the fabrication and in-service mechanics of hydraulically expanded joints. Based on this, the hydraulic expansion process can be optimized with respect to critical residual stress/cold work and the strength/leakage barrier criteria. (author)

Steam generator tubing NDE performance

International Nuclear Information System (INIS)

Henry, G.; Welty, C.S. Jr.

1997-01-01

Steam generator (SG) non-destructive examination (NDE) is a fundamental element in the broader SG in-service inspection (ISI) process, a cornerstone in the management of PWR steam generators. Based on objective performance measures (tube leak forced outages and SG-related capacity factor loss), ISI performance has shown a continually improving trend over the years. Performance of the NDE element is a function of the fundamental capability of the technique, and the ability of the analysis portion of the process in field implementation of the technique. The technology continues to improve in several areas, e.g. system sensitivity, data collection rates, probe/coil design, and data analysis software. With these improvements comes the attendant requirement for qualification of the technique on the damage form(s) to which it will be applied, and for training and qualification of the data analysis element of the ISI process on the field implementation of the technique. The introduction of data transfer via fiber optic line allows for remote data acquisition and analysis, thus improving the efficiency of analysis for a limited pool of data analysts. This paper provides an overview of the current status of SG NDE, and identifies several important issues to be addressed

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 Group Project. Annual report, 1982

International Nuclear Information System (INIS)

Clark, R.A.; Lewis, M.

1984-02-01

The Steam Generator Group Project (SGGP) is an NRC program joined by additional sponsors. The SGGP utilizes a steam generator removed from service at a nuclear plant (Surry 2) as a vehicle for research on a variety of safety and reliability issues. This report is an annual summary of progress of the program for 1982. Information is presented on the Steam Generator Examination Facility (SGEF), especially designed and constructed for this research. Loading of the generator into the SGEF is then discussed. The report then presents radiological field mapping results and personnel exposure monitoring. This is followed by information on field reduction achieved by channel head decontaminations. The report then presents results of a secondary side examination through shell penetrations placed prior to transport, confirming no change in generator condition due to transport. Decontamination of the channel head is discussed followed by plans for eddy current testing and removal of the plugs placed during service. Results of a preliminary profilometry examination are then provided

Functional performance of the helical coil steam generator, Consolidated Nuclear Steam Generator (CNSG) IV system. Executive summary report

International Nuclear Information System (INIS)

Watson, G.B.

1975-10-01

The objective of this project was to study the functional performance of the CNSG - IV helical steam generator to demonstrate that the generator meets steady-state and transient thermal-hydraulic performance specifications and that secondary flow instability will not be a problem. Economic success of the CNSG concepts depends to a great extent on minimizing the size of the steam generator and the reactor vessel for ship installation. Also, for marine application the system must meet stringent specifications for operating stability, transient response, and control. The full-size two-tube experimental unit differed from the CNSG only in the number of tubes and the mode of primary flow. In general, the functional performance test demonstrated that the helical steam generator concept will exceed the specified superheat of 35F at 100% load. The experimental measured superheat at comparable operating conditions was 95F. Testing also revealed that available computer codes accurately predict trends and overall performance characteristics

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.

Severe transient tests on operation steam generators: Analysis of the fluid structure dynamic thermal interaction

International Nuclear Information System (INIS)

Billon, F.; David, J.; Procaccia, H.

1983-01-01

The operating efficiency of steam generators (S.G.s) and their structural integrity depend on the design configurations of the feedwater spray within the S.G., and on the operating procedure. To check the merit of some design modifications, and to verify the fluid-structure interaction with a view to preserve the S.G.s integrity during severe operating transients, a special instrumentation that admits the determination of the instantaneous thermal hydraulic characteristics of the flow in the secondary water and the S.G. tube sheet, has been installed by EDF on one steam generator of Tricastin unit 1 power plant. In parallel, FRAMATOME has developped a computer code, TEMPTRON, that allows the calculations of the thermal loads and the consequent stresses in the most sollicited zones of the steam generator during transient operation of the plant. This code divides the S.G. into three parts: - the first concerns the S.G.s region above the downcomer, zone where the mixing between hot water and cold feedwater occurs, - the second is the downcomer itself which is divided into n segments, - the third concerns the tube sheet zone which is also divided into n segments. The most severe transient test performed is the auxiliary cold feedwater injection into the steam generator during a hot standby of the plant: two levels of flow rate have been realised: 55 and 110 m 3 /h of 42 0 C feedwater. The tests have shown that if the cold feedwater injection occurs when the steam generator water level is below feedwater ring, the lowest fluid temperature reached at tube sheet inlet is about 230 0 C. (orig.)

A study on improving the performance of steam generator using thermal analysis

International Nuclear Information System (INIS)

Li, Zhen Zhe; Heo, Kwang Su; Choi, Jun Hoo; Seol, Seoung Yun

2008-01-01

Steam generation mechanism is the key technology of domestic steam cleaner. Not only weight and price of steam cleaner but also the performance of steam generation mechanism must be considered to improve the competitive power of the products. In order to find out the mechanism which can be used to improve the performance of steam generator, the process of steam generation was studied at first. In the following step, possibility of control, safety of mechanism and etc were compared about the two candidated steam generation mechanism. Finally, the merit and drawback of each mechanism were summarized

Status of steam generator tubing integrity at Jaslovske Bohunice NPP

International Nuclear Information System (INIS)

Cepcek, S.

1997-01-01

Steam generator represents one of the most important component of nuclear power plants. Especially, loss of tubing integrity of steam generators can lead to the primary coolant leak to secondary circuit and in worse cases to the unit shut down or to the PTS events occurrence. Therefore, to ensure the steam generator tubing integrity and the current knowledge about tube degradation propagation and development is of the highest importance. In this paper the present status of steam generator tubing integrity in operated NPP in Slovak Republic is presented

Steam generator tube performance

International Nuclear Information System (INIS)

Tatone, O.S.; Tapping, R.L.; Stipan, L.

1992-03-01

A survey of steam generator operating experience for 1986 has been carried out for 184 pressurized water and pressurized heavy-water reactors, and 1 water-cooled, graphite-moderated reactor. Tubes were plugged at 75 of the reactors (40.5%). In 1986, 3737 tubes were plugged (0.14% of those in service) and 3148 tubes were repaired by sleeving. A small number of reactors accounted for the bulk of the plugged tubes, a phenomenon consistent with previous years. For 1986, the available tubesheet sludge data for 38 reactors has been compiled into tabular form, and sludge/deposit data will be incorporated into all future surveys

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.)

Assesment of integrity of WWER steam generator tubes

International Nuclear Information System (INIS)

Splichal, K.; Brozova, A.; Zdarek, J.

1992-01-01

Full text: The leak rates measurement project was held to give experimental data enabling the Czechoslovak Atomic Agency Inspection to decree the change in the Technical Specification allowable limit of steam generator activity release on secondary side. The WWER types of nuclear power plants in Czechoslovakia have horizontal steam generators. The tubes studying in frame of the project belong to steam generator WWER- 440 type, the diameter of tube is 16 mm, the wall thickness 1.4 mm. The subject of the project was the measurement of service leak rates of typical in service cracks. Secondary side stress corrosion cracks were determined as the typical crack created in service condition. These cracks were prepared in tubes artificially by exposition in chloride environment accompanied by an internal stress. The experimental device consisted of a pressure vessel connected with pressure water loop, a cooling vessel for leakage medium and a measuring vessel. The leak rates were determined as a slope of plots the leakage volume - time. Inside the pressure vessel the steam generator operation environment was simulated. It means: primary side of tube 12.5 MPa, Z90 deg. C, secondary side -4.6MPa, 250 deg. C, water service quality. We observed reduce of leak rate in course of time in each experiment. We suppose the tubes were stopped up by deposits formed in manufacturing of crack and in experiment. Our opinion has been proved by fractography. Project results in recommendation for in service leak rate limit based on safety factors with respect to critical crack lengths and for determination of tube plugging criteria. (author)

Steam generator and circulator model for the HELAP code

International Nuclear Information System (INIS)

Ludewig, H.

1975-07-01

An outline is presented of the work carried out in the 1974 fiscal year on the GCFBR safety research project consisting of the development of improved steam generator and circulator (steam turbine driven helium compressor) models which will eventually be inserted in the HELAP (1) code. Furthermore, a code was developed which will be used to generate steady state input for the primary and secondary sides of the steam generator. The following conclusions and suggestions for further work are made: (1) The steam-generator and circulator model are consistent with the volume and junction layout used in HELAP, (2) with minor changes these models, when incorporated in HELAP, could be used to simulate a direct cycle plant, (3) an explicit control valve model is still to be developed and would be very desirable to control the flow to the turbine during a transient (initially this flow will be controlled by using the existing check valve model); (4) the friction factor in the laminar flow region is computed inaccurately, this might cause significant errors in loss-of-flow accidents; and (5) it is felt that HELAP will still use a large amount of computer time and will thus be limited to design basis accidents without scram or loss of flow transients with and without scram. Finally it may also be used as a test bed for the development of prototype component models which would be incorporated in a more sophisticated system code, developed specifically for GCFBR's

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.)

New steam generators slated for nuclear units

International Nuclear Information System (INIS)

Anon.

1992-01-01

This article is a brief discussion of Duke Power's plans to replace steam generators at its McGuire and Catawba nuclear units. A letter of intent to purchase (from Babcock and Wilcox) the 12 Westinghouse steam generators has been signed, but no constructor has been selected at this time. This action is brought about by the failures of more than 3000 tubes in these units

An effect of downcomer feedwater fraction on steam generator performance with an axial flow economizer

International Nuclear Information System (INIS)

Jung, Byung Ryul; Park, Hu Shin; Chung, Duk Muk; Baik, Se Jin

2000-01-01

The effects of feedwater flow fraction introduced into the downcomer region have been evaluated in terms of steam generator performance based on the same steam generator thermal output for the Korea Standard Nuclear Power Plant (KSNP) steam generator. The KSNP steam generator design has an integral axial flow economizer which is designed such that most of the feedwater is introduced through the economizer region and only a portion of feedwater through the downcomer region. The feedwater flow introduced into the downcomer region is not normally controlled during the power operation. However, the actual feedwater fraction into the downcomer region may differ from the design flow depending on the as-built system and component characteristics. Investigated in this paper were the downcomer feedwater flow effects on the steam pressure, circulation ratio, internal void fraction and velocity distribution in the tube bundle region at the steady state operation using SAFE and ATHOS3 codes. The results show that the steam pressure increases and the resultant total feedwater flow increases with reducing the downcomer feedwater flow fraction for the same steam generator thermal output. The slight off-design condition of downcomer feedwater flow fraction renders no significant effect on the steam generator performance such as circulation ratios, steam qualities, void fractions and internal velocity distributions. The evaluation shows that the slight off-design downcomer feedwater flow fraction deviation up to ± 5% is acceptable for the steam generator performance

Failures of fine tubes of steam generators and the essential defects

International Nuclear Information System (INIS)

Kawano, Shinji; Ebisawa, Toru; Sato, Susumu.

1976-01-01

Light water reactors were sold to Japan as their economy and safety have been established, but the average availability of 11 reactors in Japan during 7 year operation is only 53%, and it is being proved that there are questions in the safety and economy. In this report, the failures of fine tubes of steam generators are discussed from the standpoint of the corrosion of materials. First, the functions and construction of the fine tubes of steam generators in PWRs are explained. The failures of the fine tubes of steam generators became frequent since the beginning of 1970s as large capacity nuclear power stations have started the operation. When the fine tubes are pierced with holes during operation and the radioactivity in primary coolant leaks into secondary coolant, it is detected with radioactivity monitors. In order to find out the broken tubes, eddy current flaw detectors are used, and the tubes on which flaws were detected we plugged by explosion welding. In these works, many manual operations are included, and the radiation exposure of workers and the difficulties in the operations are the problems. The cases of the tube failures in Japan and foreign countries, the causes and the countermeasures are described. Chemical corrosion, thermal stress cycle, shaving off due to eddy flow, and stress corrosion are the probable causes. The safety of steam generators is essentially in extremely poor state. The seriousness of the tube failures in steam generators is emphasized. (Kako, I.)

Chemical cleaning of nuclear (PWR) steam generators

International Nuclear Information System (INIS)

Welty, C.S. Jr.; Mundis, J.A.

1982-01-01

This paper reports on a significant research program sponsored by a group of utilities (the Steam Generator Owners Group), which was undertaken to develop a process to chemically remove corrosion product deposits from the secondary side of pressurized water reactor (PWR) power plant steam generators. Results of this work have defined a process (solvent system and application methods) that is capable of removing sludge and tube-to-tube support plate crevice corrosion products generated during operation with all-volatile treatment (AVT) water chemistry. Considers a plant-specific test program that includes all materials in the steam generator to be cleaned and accounts for the physical locations (proximity and contact) of those materials. Points out that prior to applying the process in an operational unit, the utility, with the participation of the NSSR vendor, must define allowable total corrosion to the materials of construction of the unit

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

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.

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

The progress of test and study for steam dryer in vertical steam generator

International Nuclear Information System (INIS)

Ding Xunshen

1993-01-01

Constructions, tests and test results are reviewed for three types of steam generator dryer that are concentric vertical corrugated separator, centrifugal conic separator and chevron separator. The last type is considered as the best one in comparison, which has been applied to Qinshan 300 MW steam generator. A number of pertinent remarks to draining scheme, hydraulic loss reduction, and conduct of test are given based on experiences

Thermal-hydraulic characteristic of the PGV-1000 steam generator

International Nuclear Information System (INIS)

Ubra, O.; Doubek, M.

1995-01-01

Horizontal steam generators are typical parts of nuclear power plants with pressure water reactor type VVER. By means of this computer program, a detailed thermal-hydraulic study of the horizontal steam generator PGV-1000 has been carried out and a special attention has been paid to the thermal-hydraulics of the secondary side. A set of important steam generator characteristics has been obtained and analyzed. Some of the interesting results of the analysis are presented in the paper. (author)

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

Performance tests and efficiency analysis of Solar Invictus 53S - A parabolic dish solar collector for direct steam generation

Science.gov (United States)

Jamil, Umer; Ali, Wajahat

2016-05-01

This paper presents the results of performance tests conducted on Solar Invictus 53S `system'; an economically effective solar steam generation solution designed and developed by ZED Solar Ltd. The system consists of a dual axis tracking parabolic solar dish and bespoke cavity type receiver, which works as a Once Through Solar Steam Generator `OTSSG' mounted at the focal point of the dish. The overall performance and efficiency of the system depends primarily on the optical efficiency of the solar dish and thermal efficiency of the OTSSG. Optical testing performed include `on sun' tests using CCD camera images and `burn plate' testing to evaluate the sunspot for size and quality. The intercept factor was calculated using a colour look-back method to determine the percentage of solar rays focused into the receiver. Solar dish tracking stability tests were carried out at different times of day to account for varying dish elevation angles and positions, movement of the sunspot centroid was recorded and logged using a CCD camera. Finally the overall performance and net solar to steam efficiency of the system was calculated by experimentally measuring the output steam parameters at varying Direct Normal Insolation (DNI) levels at ZED Solar's test facility in Lahore, Pakistan. Thermal losses from OTSSG were calculated using the known optical efficiency and measured changes in output steam enthalpy.

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)

Specific features of emergency processes associated with water leacs into sodium in a reverse steam generator

International Nuclear Information System (INIS)

Sroelov, V.S.; Nikol'skij, R.V.; Chernobrovkin, Yu.V.; Privalov, Yu.V.; Bocharin, P.P.; Shtynda, Yu.E.

1986-01-01

Experimental and theoretical data characterizing the development of emergency processes arising in the course of water leaks into sodium in a reverse steam generator (sodium in tubes, water in intertube space) are considered. The results of calculations performed for BOR-60 reactor steam generator at initial leaks of 0.01 and 0.55 g/s are presented. It is shown that in the reverse steam generator the development of accident occurs much slower than in steam generators of traditional design. At same stage of accident sodium is displaced from the damaged tube and as a result the destruction of tube material discontinues. The conclusion is drawn that by the development of emergency protection systems for reverse steam generator the requirements for sensitivity and fast response of leak detectors could be reduced

Clinch river breeder reactor plant steam generator water quality

Energy Technology Data Exchange (ETDEWEB)

Van Hoesen, D; Lowe, P A

1975-07-01

The recent problems experienced by some LWR Steam Generators have drawn attention to the importance of system water quality and water/ steam side corrosion. Several of these reactor plants have encountered steam generator failures due to accelerated tube corrosion caused, in part, by poor water quality and corrosion control. The CRBRP management is aware of these problems, and the implications that they have for the Clinch River Breeder Reactor Plant (CPBRP) Steam Generator System (SGS). Consequently, programs are being implemented which will: (1) investigate the corrosion mechanisms which may be present in the CRBRP SGS; (2) assure steam generator integrity under design and anticipated off-normal water quality conditions; and (3) assure that the design water quality levels are maintained at all times. However, in order to understand the approach being used to examine this potential problem, it is first necessary to look at the CRBRP SGS and the corrosion mechanisms which may be present.

Clinch river breeder reactor plant steam generator water quality

International Nuclear Information System (INIS)

Van Hoesen, D.; Lowe, P.A.

1975-01-01

The recent problems experienced by some LWR Steam Generators have drawn attention to the importance of system water quality and water/ steam side corrosion. Several of these reactor plants have encountered steam generator failures due to accelerated tube corrosion caused, in part, by poor water quality and corrosion control. The CRBRP management is aware of these problems, and the implications that they have for the Clinch River Breeder Reactor Plant (CPBRP) Steam Generator System (SGS). Consequently, programs are being implemented which will: 1) investigate the corrosion mechanisms which may be present in the CRBRP SGS; 2) assure steam generator integrity under design and anticipated off-normal water quality conditions; and 3) assure that the design water quality levels are maintained at all times. However, in order to understand the approach being used to examine this potential problem, it is first necessary to look at the CRBRP SGS and the corrosion mechanisms which may be present

Dryout occurrence in a helically coiled steam generator for nuclear power application

Directory of Open Access Journals (Sweden)

Santini L.

2014-03-01

Full Text Available Dryout phenomena have been experimentally investigated in a helically coiled steam generator tube. The experiences carried out in the present work are part of a wide experimental program devoted to the study of a GEN III+ innovative nuclear power plant [1].The experimental facility consists in an electrically heated AISI 316L stainless steel coiled tube. The tube is 32 meters long, 12.53 mm of inner diameter, with a coil diameter of 1m and a pitch of 0.79 m, resulting in a total height of the steam generator of 8 meters. The thermo-hydraulics conditions for dryout investigations covered a spectrum of mass fluxes between 199 and 810 kg/m2s, the pressures ranges from 10.7 to 60.7 bar, heat fluxes between 43.6 to 209.3 kW/m2.Very high first qualities dryout, between 0.72 and 0.92, were found in the range of explored conditions, comparison of our results with literature available correlations shows the difficulty in predicting high qualities dryout in helical coils., immediately following the heading. The text should be set to 1.15 line spacing. The abstract should be centred across the page, indented 15 mm from the left and right page margins and justified. It should not normally exceed 200 words.

Studying the processes of sodium-water interaction in the BOR-60 reactor micromodule steam generator

International Nuclear Information System (INIS)

Tsykanov, V.A.; Antipin, G.K.; Borisov, V.V.

1981-01-01

Main results of experimental studies of emergency regimes of micromodule steam generator (MSG) at small and big leaks of water into sodium, realized using the 30 MW MSG, operating in the BOR-o0 reactor, are considered. The aims of the study are as follows: the modelling of macroleak in ''Nadja'' steam generator for the BN-350 reactor; testing the conceptions of alarm signalling and MSG protection; testing under real conditions of new perspective systems of leak detection; gaining the experimence and development of the ways to eliminate the consequences of accident caused by big water leak into sodium; accumulation of knowledge on restoration of MSG operating ability after accident; experimental test of calculational techniques for big leak accidents to use them in future for calculational studies of similar situations at other reactors equipped with sodium-water steam generators; refinement of characteristics of hydrodynamic and thermal effects interaction zone for big leak in real circuit during the plant operation. A series of experiments with the imitation of water leak into sodium by means of argon and steam supply through injection devices, located before the steam superheater module of one of the sections and between evaporator module of the same section, is conducted. The range of steam flow rate is 0.02-0.45 g/s. Duration of steam supply is 100-400 s. A conclusion is made that the results obtained can be used for steam generator of the BN-350 reactor [ru

Steam generator replacement at the Obrigheim nuclear power station

International Nuclear Information System (INIS)

Pickel, E.; Schenk, H.; Huemmler, A.

1984-01-01

The Obrigheim Nuclear Power Station (KWO) is equipped with a dual-loop pressurized water reactor of 345 MW electric power; it was built by Siemens in the period 1965 to 1968. By the end of 1983, KWO had produced some 35 billion kWh in 109,000 hours of operation. Repeated leaks in the heater tubes of the two steam generators had occurred since 1971. Both steam generators were replaced in the course of the 1983 annual revision. Kraftwerk Union AG (KWU) was commissioned to plant and carry out the replacement work. Despite the leakages the steam generators had been run safely and reliably over a period of 14 years until their replacement. Replacing the steam generators was completed within twelve weeks. In addition to the KWO staff and the supervising crew of KWU, some 400 external fitters were employed on the job at peak work-load periods. For the revision of the whole plant, work on the emergency systems and replacement of the steam generators a maximum number of approx. 900 external fitters were employed in the plant in addition to some 250 members of the plant crew. The exposure dose of the personnel sustained in the course of the steam generator replacement was 690 man-rem, which was clearly below previous estimates. (orig.) [de

An automated eddy current in-service inspection system for nuclear steam generator tubing

International Nuclear Information System (INIS)

Wells, N.S.

1981-06-01

A prototype steam generator in-service inspection system incorporating remotely-controlled instrumentation linked by a digital transmission line to an instrument and control trailer outside the reactor containment has been designed and manufactured and is presently undergoing field tests. The (Monel 400) steam generator tubes are scanned two at a time using absolute eddy current probes controlled by two remotely-operated probe drives at a scanning speed of 0.5 m/s. The probes are positioned on the tubesheet by a light-weight (1.5 kg) microprocessor-operated tubesheet walker mechanism. Digitized control and data signals are transmitted up to 300 m to the control trailer. There the control and analysis computers extract the relevant signal information and present it in condensed form as labelled graphics on CRT consoles for on-line visual assessment. Hard copy output is also provided for each tube scanned (one per minute). Condensed data is archived on magnetic tapes for additional off-line analysis and comparisons with other inspections

Steam producing plant concept of 4S for oil sand extraction

International Nuclear Information System (INIS)

Matsuyama, Shinichiro; Nishiguchi, Youhei; Sakashita, Yoshiaki; Kasuga, Shoji; Kawashima, Masatoshi

2009-01-01

Plant concept of small fast reactor '4S' applying to continuous steam production for recovery of crude oil from oil sands was investigated. Assuming typical steam assisted gravity drainage (SAGD) plant whose production scale is 120,000 barrels per day of a crude oil, concept of nuclear steam supply system consisting of eight reactor modules for steam production and three reactor modules for electric generation of the 4S with a thermal rating of 135 MWt was established without any essential or significant design change from the preceding 4S with a thermal rating of 30 MWt. The 4S, provided for an oil sand extraction, will reduce greenhouse gas emission significantly, and has not much burden for development and licensing and has economic competitiveness. (author)

Technical development and its application on steam generator replacement

International Nuclear Information System (INIS)

Morita, Sadahiko; Hanzawa, Katsumi; Sato, Hajime; Kannoto, Yasuo.

1995-01-01

Twenty-two PWR nuclear power plants are now under commercial operation in Japan. Eight of these plants are scheduled to have their steam generators replaced by up-graded units as a social responsibility for improved reliability, economy and easier maintenance. To carry out steam generator replacement, main coolant pipe cutting and restoration techniques, remote controlled welding machines and other remote controlled equipment, templating techniques with which the new steam generator primary nozzles will fit the existing primary pipes correctly were developed. An adequate training program was carried out to establish these techniques and they were then applied in replacement work on site. The steam generators of the three plants were replaced completely in 1994. These newly developed techniques are to be applied in upcoming plants and replaced plants will be much reliable. (author)

Research perspectives on the evaluation of steam generator tube integrity

International Nuclear Information System (INIS)

Muscara, J.; Diercks, D.R.; Majumdar, S.; Kupperman, D.S.; Bakhtiari, S.; Shack, W.J.

2002-01-01

Industry efforts have been largely successful in managing degradation of steam generator tubes due to wastage, pitting, and denting, but fretting, stress corrosions cracking (SCC) and intergranular attack have proved more difficult to manage. Although steam generator replacements are proceeding, there is substantial industry interest in operating with degraded steam generators, and significant numbers of plants will continue to do so. In most cases degradation of steam generator tubing by stress corrosion cracking is still managed by 'plug or repair on detection' because current NDE techniques for characterization of flaws and the knowledge of SCC crack growth rates are not accurate enough to permit continued operation. Replacement generators with improved designs and materials have performed well to date, but previous experience with the appearance of some types of SCC in Alloy 600 after 10 years or more of operation and laboratory results suggest additional understanding of corrosion performance of these materials is needed. This paper reviews some of the historical background that underlies current steam generator degradation management strategies and outlines some of the additional research that must be done to provide more effective management of degradation in current generators and provide greater assurance of satisfactory performance in replacement steam generators. (author)

Corrosion problems of PWR steam generators

International Nuclear Information System (INIS)

Urbancik, L.; Kostal, M.

Literature data are assessed on corrosion failures of steam generator tubes made of INCONEL 600 or INCOLOY 800. It was found that both alloys with high nickel content showed good stability in a corrosion environment while being sensitive to carbide formation on grain boundaries. The gradual depletion of chromium results from the material and corrosion resistance deteriorates. INCOLOY 800 whose chromium carbide precipitation on grain boundaries in pure water and steam is negligible up to 75O degC and which is not subject to corrosion attacks in the above media and in an oxidizing environment at a temperature to about 700 degC shows the best corrosion resistance. Its favourable properties were tested in long-term operation in the Peach Bottom 1 nuclear power plant where no failures due to corrosion of this material have been recorded since 1967. In view of oxygenic-acid surface corrosion, it is necessary to work in a neutral or slightly basic environment should any one of the two alloys be used for steam generator construction. The results are summed up of an analysis conducted for the Beznau I NOK reactor. Water treatment with ash-free amines can be used as prevention against chemical corrosion mechanisms, although the treatment itself does not ensure corrosion resistance of steam generator key components. (J.B.)

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)

Design Evolution and Verification of the A-3 Chemical Steam Generator

Science.gov (United States)

Kirchner, Casey K.

2009-01-01

Following is an overview of the Chemical Steam Generator system selected to provide vacuum conditions for a new altitude test facility, the A-3 Test Stand at Stennis Space Center (SSC) in Bay St. Louis, MS. A-3 will serve as NASA s primary facility for altitude testing of the J-2X rocket engine, to be used as the primary propulsion device for the upper stages of the Ares launch vehicles. The Chemical Steam Generators (CSGs) will produce vacuum conditions in the test cell through the production and subsequent supersonic ejection of steam into a diffuser downstream of the J-2X engine nozzle exit. The Chemical Steam Generators chosen have a rich heritage of operation at rocket engine altitude test facilities since the days of the Apollo program and are still in use at NASA White Sands Test Facility (WSTF) in New Mexico. The generators at WSTF have been modified to a degree, but are still very close to the heritage design. The intent for the A-3 implementation is to maintain this heritage design as much as possible, making minimal updates only where necessary to substitute for obsolete parts and to increase reliability. Reliability improvements are especially desired because the proposed system will require 27 generators, which is nine times the largest system installed in the 1960s. Improvements were suggested by the original design firm, Reaction Motors, by NASA SSC and NASA WSTF engineers, and by the A-3 test stand design contractor, Jacobs Technology, Inc. (JTI). This paper describes the range of improvements made to the design to date, starting with the heritage generator and the minor modifications made over time at WSTF, to the modernized configuration which will be used at A-3. The paper will discuss NASA s investment in modifications to SSC s E-2 test facility fire a full-scale Chemical Steam Generator in advance of the larger steam system installation at A-3. Risk mitigation testing will be performed in early 2009 at this test facility to verify that the CSGs

Some aspects affecting fast reactor steam generator integrity considered from a utility viewpoint

Energy Technology Data Exchange (ETDEWEB)

Bolt, P R

1975-07-01

The important conditions affecting fast reactor steam generator integrity are discussed. In addition to the need for high integrity levels when the steam generator is first delivered to the power station site, the equally important aspect of demonstrating retention of continued high levels of integrity throughout the operating life of the station is described. The functional and related conditions that are believed important to the selection of a design type which can offer adequately high levels of integrity are given. Some of the data needs of a utility concerned with fast reactor S.G.U. design assessment are described, particular emphasis being given to areas believed to have a significant effect on steam generator reliability and integrity. (author)

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)

Coupled Calculations in Helical Steam Generator: Validation on Legacy Data

Energy Technology Data Exchange (ETDEWEB)

Merzari, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Yuan, Haomin [Argonne National Lab. (ANL), Argonne, IL (United States); Kraus, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Solberg, Jerome [Argonne National Lab. (ANL), Argonne, IL (United States); Ferencz, Robert M. [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-09-30

The NEAMS program aims to develop an integrated multi-physics simulation capability “pellet-to-plant” for the design and analysis of future generations of nuclear power plants. In particular, the Reactor Product Line code suite's multi-resolution hierarchy is being designed to ultimately span the full range of length and time scales present in relevant reactor design and safety analyses, as well as scale from desktop to petaflop computing platforms. Flow-induced vibration (FIV) is widespread problem in energy systems because they rely on fluid movement for energy conversion. Vibrating structures may be damaged as fatigue or wear occurs. Given the importance of reliable components in the nuclear industry, flow-induced vibration has long been a major concern in safety and operation of nuclear reactors. In particular, nuclear fuel rods and steam generators have been known to suffer from flow-induced vibration and related failures. Advanced reactors, such as integral Pressurized Water Reactors (PWRs) considered for Small Modular Reactors (SMR), often rely on innovative component designs to meet cost and safety targets. One component that is the subject of advanced designs is the steam generator, some designs of which forego the usual shell-and-tube architecture in order to fit within the primary vessel. In addition to being more cost- and space-efficient, such steam generators need to be more reliable, since failure of the primary vessel represents a potential loss of coolant and a safety concern. A significant amount of data exists on flow-induced vibration in shell-and-tube heat exchangers, and heuristic methods are available to predict their occurrence based on a set of given assumptions. In contrast, advanced designs have far less data available. Advanced modeling and simulation based on coupled structural and fluid simulations have the potential to predict flow-induced vibration in a variety of designs, reducing the need for expensive experimental

Repair technology for steam generator tubes

Energy Technology Data Exchange (ETDEWEB)

Kim, Seung Ho; Jung, Hyun Kyu; Jung, Seung Ho; Kim, Chang Hoi; Jung, Young Moo; Seo, Yong Chil; Kim, Jung Su; Seo, Moo Hong

2001-02-01

The most commonly used sleeving materials are thermally treated Alloy 600 and thermally treated Alloy 690 Alloy. Currently, thermally treated Alloy 690 and Alloy 800 are being offered although Alloy 800 has not been licensed in the US. To install sleeve, joint strength, leak tightness, PWSCC resistance, evaluation on process parameter range and the effect of equipments and procedures on repair plan and radiation damage have to be investigated before sleeving. ABB CE provides three type of leak tight Alloy 690 TIG welded and PLUSS sleeve. Currently, Direct Tube Repair technique using Nd:YAG laser has been developed by ABB CE and Westinghouse. FTI has brazed and kinetic sleeve designs for recirculating steam generator and hydraulic and rolled sleeve designs for one-through steam generators. Westinghouse provides HEJ, brazed and laser welded sleeve design. When sleeve is installed in order to repair the damaged S/G tubes, it is certain that defects can be occurred due to the plastic induced stress and thermal stress. Therefore it is important to minimize the residual stress. FTI provides the electrosleeve technique as a future repair candidate using electroplating.

Repair technology for steam generator tubes

International Nuclear Information System (INIS)

Kim, Seung Ho; Jung, Hyun Kyu; Jung, Seung Ho; Kim, Chang Hoi; Jung, Young Moo; Seo, Yong Chil; Kim, Jung Su; Seo, Moo Hong

2001-02-01

The most commonly used sleeving materials are thermally treated Alloy 600 and thermally treated Alloy 690 Alloy. Currently, thermally treated Alloy 690 and Alloy 800 are being offered although Alloy 800 has not been licensed in the US. To install sleeve, joint strength, leak tightness, PWSCC resistance, evaluation on process parameter range and the effect of equipments and procedures on repair plan and radiation damage have to be investigated before sleeving. ABB CE provides three type of leak tight Alloy 690 TIG welded and PLUSS sleeve. Currently, Direct Tube Repair technique using Nd:YAG laser has been developed by ABB CE and Westinghouse. FTI has brazed and kinetic sleeve designs for recirculating steam generator and hydraulic and rolled sleeve designs for one-through steam generators. Westinghouse provides HEJ, brazed and laser welded sleeve design. When sleeve is installed in order to repair the damaged S/G tubes, it is certain that defects can be occurred due to the plastic induced stress and thermal stress. Therefore it is important to minimize the residual stress. FTI provides the electrosleeve technique as a future repair candidate using electroplating

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

Emergency systems and protection equipment of modular steam generators for fast reactors

International Nuclear Information System (INIS)

Matal, O.

The requirements are discussed for accident protection of modular steam generators for fast reactors. Accident protection is assessed for a modular through-flow steam generator and for a natural circulation modular steam generator. Benefits and constraints are shown and possible improvements are outlined for accident protection of liquid sodium fired modular steam generators. (Kr)

Experimental and theoretical investigations on safety of the SNR - straight-tube design steam generator with sodium-water reactions

International Nuclear Information System (INIS)

Dumm, K.; Sauermann, F.; Schnitker, W.; Welter, A.

A number of large sodium-water reaction tests has been performed in a steam generator model in order to verify the layout criteria of the SNR straight-tube design steam generators under accident conditions. The experimental setup is described. The test results and their applicability to the SNR steam generators are given and discussed. (U.S.)

Steam generator replacement at Bruce A: approach, results, and lessons learned

International Nuclear Information System (INIS)

Tomkiewicz, W.; Savage, B.; Smith, J.

2008-01-01

Steam Generator Replacement is now complete in Bruce A Units 1 and 2. In each reactor, eight steam generators were replaced; these were the first CANDU steam generator replacements performed anywhere in the world. The plans for replacement were developed in 2004 and 2005, and were summarized in an earlier paper for the CNS Conference held in November, 2006. The present paper briefly summarizes the methodologies and special processes used such as metrology, cutting and welding and heavy lifting. The paper provides an update since the earlier report and focuses on the project achievements to date, such as: - A combination of engineered methodology, laser metrology and precise remote machining led to accurate first time fit-ups of each new replacement steam generator and steam drums - Lessons learned in the first unit led to schedule improvements in the second unit - Dose received was lowest recorded for any steam generator replacement project. The experience gained and lessons learned from Units 1 and 2 will be valuable in planning and executing future replacement steam generator projects. A video was presented

Fretting-wear characteristics of steam generator tubes contacting with foreign object

International Nuclear Information System (INIS)

Jo, Jong Chull; Jhung, Myung Jo; Kim, Woong Sik; Kim, Hho Jung

2003-01-01

Fretting-wear characteristics of steam generator tubes contacting with foreign object has been investigated in this study. The operating steam generator shell-side flow field conditions are obtained from three-dimensional steam generator flow calculation using a well-validated steam generator thermal-hydraulic analysis computer code. Modal analyses are performed for the finite element modelings of tubes to get the natural frequency, corresponding mode shape and participation factor. The wear rate of a steam generator tube caused by foreign object is calculated using the Archard formula and the remaining life of the tube is predicted. In addition, the effects of internal pressure and flow velocity on the remaining life of the tube are discussed in this paper

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

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

Corrosion problems in PWR steam generators

International Nuclear Information System (INIS)

Weber, J.; Suery, P.

1976-01-01

Examinations on pulled steam generator tubes from the Swiss nuclear power plants Beznau I and II, together with some laboratory tests, may be summarized as follows: Corrosion problems in vertical U-tube steam generators with Alloy 600 as tube material are localized towards relatively narrow regions above the tube sheet where thermohydraulic conditions and, as a consequence thereof, chemical conditions are uncontrolled. Within these zones Alloy 600 is not sufficienthy resistent to caustic or phosphate attack (caustic stress corrosion cracking and general corrosion, resp.). The mechanisms of several corrosion phenomena are not fully understood. (orig.) [de

Experience with modular steam generator production and application of new testing methods

International Nuclear Information System (INIS)

Olesovsky

Experience is reviewed gained at the Trebic IBZKG plant with the production of modular steam generators. The plant started producing steam generators for the Jaslovske Bohunice nuclear power plant in 1965. In addition to the steam generator for the A-1, the plant also produced a loop for the Melekess power plant and a steam generator for the BOR-60 reactor. Operating experience gained so far allowed improving the quality of the BOR steam generator, especially in the tube-tube plate joint. A double tube plate was used and the welded joint shape was changed. As a result of high requirements on the quality of welded joints, the steam generator has successfully been in operation for more then 10,000 hours. The existing experience was utilized in designing a new steam generator named Nadya. Many design and technological requirements were presented concerning the Nadya generator and many new checking operations have been included in technology. (Kr)

Evolution of management activities and performance of the Point Lepreau Steam Generators

International Nuclear Information System (INIS)

Slade, J.; Keating, J.; Gendron, T.

2007-01-01

The Point Lepreau steam generators have been in service since 1983 when the plant was commissioned. During the first thirteen years of operation, Point Lepreau steam generator maintenance issues led to 3-4% unplanned plant incapability Steam generator fouling, corrosion, and the introduction of foreign materials during maintenance led to six tube leaks, two unplanned outages, two lengthy extended outages, and degraded thermal performance during this period. In recognition of the link between steam generator maintenance activities and plant performance, improvements to steam generator management activities have been continuously implemented since 1987. This paper reviews the evolution of steam generator management activities at Point Lepreau and the resulting improved trends in performance. Plant incapability from unplanned steam generator maintenance has been close to zero since 1996. The positive trends have provided a strong basis for the management strategies developed for post-refurbishment operation. (author)

Steam generator inspection activities at the EPRI NDE Center

International Nuclear Information System (INIS)

Krzywosz, K.

1988-01-01

Various types of corrosion and mechanical damage continue to affect the availability of both recirculating and once-through steam generators. Both the tube bundle and its supporting structure are affected. Intergranular attack and stress corrosion cracking (SCC) are the corrosion-assisted tube-wall damage mechanisms of most concern at this time. Fatigue cracking and fretting at antivibration bars are currently the mechanical damage forms causing most concern. Improved NDE equipment and techniques are providing better detection and characterization of adverse conditions within the steam generators and doing it at an earlier stage. This allows timely corrective action. To maintain the projected life expectancy of existing and new steam generators, remedial measures have been implemented. These measures include shot- or roto-peening, U-bend stress relief, chemical cleaning of secondary side, and sleeving of tubes. The improved NDE technology will also be instrumental in monitoring and assessing the effectiveness of the remedial measures. The revision of guidance documents for steam generator in-service inspection (ISI) is providing more relevant information to support this complex operation. A multitasked project is described that includes evaluation of steam generator tube NDE technology, transfer of this technology to utilities, and rapid response utility assistance

Utilizing Philippine Calatrava coal-diesel oil mixture (CDOM) as alternative fuel for industrial steam generator

Energy Technology Data Exchange (ETDEWEB)

Archie B. Maglaya [De La Salle University, Manila (Philippines). Department of Mechanical Engineering

2005-01-01

The fast depletion of fuel oil and the continuous increase in the demand for power is a global issue. In the Philippines, the demand for diesel oil is expected to increase significantly in a 20-year period as projected by the Department of Energy. In line with the Philippine Government's thrust to lessen the dependence on imported energy, the agenda for the search for alternative fuel is highly prioritized. Thus, this paper presents the results of the study on performance analysis and efficiency test of a diesel oil fired industrial steam generator using Philippine Calatrava coal-diesel oil mixture (CDOM) as alternative fuel. A computer program was developed in HyperText Markup Language (HTML{copyright}) and JavaScript{copyright} to aid the computation of the adiabatic flame temperature from the governing system of equations based on the heat interaction between CDOM fuel, combustion air and products of combustion to determine the most desirable alternative fuel. Actual experimentation for the determination of CDOM fuel properties was also conducted to verify the alternative fuel selected through theoretical calculations. Results showed that the CDOM fuel with a particle size passing 75 {mu}m (-200 mesh) sieve having a proportion of 5% pulverized coal-95% diesel oil and 10% pulverized coal-90% diesel oil could be handled throughout the test with no degradation of the industrial steam generator. The steam generator efficiency using diesel oil is close to the steam generator efficiency using both CDOM fuels. 20 refs., 5 figs., 4 tabs.

Dynamic study of steam generation from low-grade waste heat in a zeolite–water adsorption heat pump

International Nuclear Information System (INIS)

Xue, Bing; Meng, Xiangrui; Wei, Xinli; Nakaso, Koichi; Fukai, Jun

2015-01-01

A novel zeolite–water adsorption heat pump system based on a direct-contact heat exchange method to generate steam from low-grade waste gas and water has been proposed and examined experimentally. Superheated steam (200 °C, 0.1 MPa) is generated from hot water (70–80 °C) and dry air (100–130 °C). A dynamic model for steam generation process is developed to describe local mass and heat transfer. This model features a three-phase calculation and a moving water–gas interface. The calculations are carried out in the zeolite–water and zeolite–gas regions. Model outputs are compared with experimental results for validation. The thermal response inside the reactor and mass of steam generated is well predicted. Numerical results show that preheat process with low-temperature steam is an effective method to achieve local equilibrium quickly, thus generation process is enhanced by prolonging the time and increasing mass of the generated steam. Besides, high-pressure steam generation up to 0.5 MPa is possible from the validated dynamic model. Future work could be emphasized on enhancing high-pressure steam generation with preheat process or mass recovery operation

Large-leak sodium-water reaction analysis for steam generators

International Nuclear Information System (INIS)

Sakano, K.; Shindo, Y.; Hori, M.

1975-01-01

The guillotine rupture of 4 tubes is assumed as a design basis regarding the large-leak sodium-water reaction in the system of the MONJU steam generator. Three kinds of analyses were performed with the view to showing the integrity of the steam generator system on the reaction. The first one is the analysis of the initial pressure spike, assuming the initial guillotine rupture of 1 tube. The analysis was performed by utilizing one-dimensional sphere-cylinder model code SWAC-7 and two-dimensional axisymmetric code PISCES 2DL. The second one is the analysis of the secondary peak pressure and its propagation in the system, assuming the instantaneous guillotine rupture of 4 tubes. The third one is the analysis of the dynamic deformation of the steam generator shell. The integrity of the steam generator system was shown by the analyses. (author)

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

Modular sludge collection system for a nuclear steam generator

International Nuclear Information System (INIS)

Appleman, R.H.; Bein, J.D.; Powasaki, F.S.

1986-01-01

A sludge collection system is described for a vertically oriented nuclear steam generator wherein vapors produced in the steam generator pass through means for separating entrained liquid from the vapor prior to the vapor being discharged from the steam generator. The sludge collection system comprises: an upwardly open chamber for collecting the separated liquid and feedwater entering the steam generator; upwardly open sludge collecting containers positioned within the chamber, wherein each of the containers includes a top rim encompassing an opening leading to the interior of each container; generally flat, perforated covers, each of the covers being positioned over one of the openings such that a gap is formed between the cover and the adjacent top rim; sludge agitating means on at least one of the containers; and sludge removal means on at least one of the containers

Large-leak sodium-water reaction analysis for steam generators

Energy Technology Data Exchange (ETDEWEB)

Sakano, K; Shindo, Y; Hori, M

1975-07-01

The guillotine rupture of 4 tubes is assumed as a design basis regarding the large-leak sodium-water reaction in the system of the MONJU steam generator. Three kinds of analyses were performed with the view to showing the integrity of the steam generator system on the reaction. The first one is the analysis of the initial pressure spike, assuming the initial guillotine rupture of 1 tube. The analysis was performed by utilizing one-dimensional sphere-cylinder model code SWAC-7 and two-dimensional axisymmetric code PISCES 2DL. The second one is the analysis of the secondary peak pressure and its propagation in the system, assuming the instantaneous guillotine rupture of 4 tubes. The third one is the analysis of the dynamic deformation of the steam generator shell. The integrity of the steam generator system was shown by the analyses. (author)

46 CFR 61.15-5 - Steam piping.

Science.gov (United States)

2010-10-01

... 46 Shipping 2 2010-10-01 2010-10-01 false Steam piping. 61.15-5 Section 61.15-5 Shipping COAST... Periodic Tests of Piping Systems § 61.15-5 Steam piping. (a) Main steam piping shall be subjected to a... removed and the piping thoroughly examined. (b) All steam piping subject to pressure from the main boiler...

Application of Integrated Neural Network Method to Fault Diagnosis of Nuclear Steam Generator

International Nuclear Information System (INIS)

Zhou Gang; Yang Li

2009-01-01

A new fault diagnosis method based on integrated neural networks for nuclear steam generator (SG) was proposed in view of the shortcoming of the conventional fault monitoring and diagnosis method. In the method, two neural networks (ANNs) were employed for the fault diagnosis of steam generator. A neural network, which was used for predicting the values of steam generator operation parameters, was taken as the dynamics model of steam generator. The principle of fault monitoring method using the neural network model is to detect the deviations between process signals measured from an operating steam generator and corresponding output signals from the neural network model of steam generator. When the deviation exceeds the limit set in advance, the abnormal event is thought to occur. The other neural network as a fault classifier conducts the fault classification of steam generator. So, the fault types of steam generator are given by the fault classifier. The clear information on steam generator faults was obtained by fusing the monitoring and diagnosis results of two neural networks. The simulation results indicate that employing integrated neural networks can improve the capacity of fault monitoring and diagnosis for the steam generator. (authors)

TRAC analysis of steam-generator overfill transients for TMI-1

International Nuclear Information System (INIS)

Bassett, B.

1983-01-01

A reactor safety issue concerning the overfilling of once-through steam generators leading to combined primary/secondary blowdown has been raised recently. A series of six calculations, performed with the LWR best-estimate code, TRAC-PD2, on a Babcock and Wilcox Plant (TMI-1), was performed to investigate this safety issue. The base calculation assumed runaway main feedwater to one steam generator causing it to overfill and to break the main steam line. Four additional calculations build onto the base case with combinations of a pump-seal failure, a steam-generator tube rupture, and the pilot-operated relief valve not reseating. A sixth calculation involved only the rupture of a single steam-generator tube. The results of these analyses indicate that for the transients investigated, the emergency cooling system provided an adequate make-up coolant flow to mitigate the accidents

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

Steam generators and fuel engineering utilizing solid, liquid, gaseous and special fuels

Energy Technology Data Exchange (ETDEWEB)

Thor, G

1983-01-01

Provided were technological specifications and details in the design of brown coal fired steam generators, produced in the German Democratic Republic. These steam generators range in their capacity between 1.6 and more than 1,000 t/h. The appropriate coal feeding systems, water supply and cleaning equipment, coal pulverizers and ash removal units are also manufactured. Various schemes show the design of a 25 to 64 t/h, a 320 t/h and an 815 t/h brown coal steam generator. Specifications are given for series of fuel pulverizers available, for the water circulation system and steam evaporators. The VEB Dampferzeugerbau Berlin also offers steam generators for saliniferous brown coal with a steam capacity up to 125 t/h, steam generators for pulverized black coal with a capacity up to 350 t/h and oil and gas fired generators up to 250 t/h. The company has experience in combustion of biomass (sugar cane waste) with oil in steam generators of more than 100 t/h capacity, and in projecting firing systems for other biofuels including rice, peanut and coconut hulls, wood and bark. Multi-biofuel firing in combination with coal for steam generation is also regarded as possible. (In English)

The ageing of CANDU steam generator due to localized corrosion

International Nuclear Information System (INIS)

Lucan, D.; Fulger, M.; Jinescu, Ghe.

2001-01-01

The principal types of corrosion are presented which can occur in CANDU steam generator. There are also presented the operation conditions, the specifications referring to the water chemistry and the construction materials of Steam Generator, the factors that have a great influence on the corrosion behaviour during the whole exploitation period of this equipment. The most important elements of CANDU Steam Generator ageing management program are also discussed. (R. P.)

Design and performance of BWC replacement steam generators for PWR systems

Energy Technology Data Exchange (ETDEWEB)

Klarner, R.; Steinmoeller, F.; Millman, J.; Schneider, W. [Babcock and Wilcox Canada, Cambridge, Ontario (Canada)

1998-07-01

In recent years, Babcock and Wilcox Canada (BWC) has provided a number of PWR Replacement Steam Generators (RSGS) to replace units that had experienced extensive Alloy 600 tube degradation. BWC RSG units are in operation at Northeast Utilities' Millstone Unit 2, Rochester Gas and Electric's Ginna Station, Duke Energy's Catawba Unit 1, McGuire Unit 1 and 2, Florida Power and Light's St. Lucie Unit 1 and Commonwealth Edison's Byron 1 Station. Extensive start-up performance characteristics have been obtained for Millstone 2, Ginna, McGuire 1, and Catawba 1 RSGS. The Millstone 2, Ginna and Catawba 1 RSGs have also undergone extensive inspections following their first cycle of operation. The design and start-up performance characteristics of these RSGs are presented. The BWC Replacement Steam generators were designed to fit the existing envelope of pressure boundary dimensions to ensure licensability and integration into the Nuclear Steam Supply System. The RSGs were provided with a tube bundle of Alloy 690TT tubing, sized to match or exceed the original steam generator (OSG) thermal performance including provision for the reduced thermal conductivity of Alloy 690 relative to Alloy 600. The RSG tube bundle configurations provide a higher circulation design relative to the OSG, and feature corrosion resistant lattice grid and U-bend tube supports which provide effective anti-vibration support. The tube bundle supports accommodate relatively unobstructed flow and allow unrestrained structural interactions during thermal transients. Efficient steam separators assure low moisture carryover as well as high circulation. Performance measurements obtained during start-up verify that the BWC RSGs meet or exceed the specified thermal and moisture carryover performance requirements. RSG water level stability results at nor-mal operation and during plant transients have been excellent. Visual and ECT inspections have confirmed minimal deposition and 100

Inspection and repair of steam generator tubing with a robot

International Nuclear Information System (INIS)

Boehm, H.H.; Foerch, H.

1985-01-01

During inspection and repair of steam generator tubing, radiation exposure to personnel is an unrequested endowment. To combat this intrinsic handicap, a robot has been designed for deployment in all operations inside the steam generator water chamber. This measure drastically reduces entering time and also improves inspection capabilities with regard to the accuracy and reproduction of the desired tube address. The inherent flexibility of the robot allows for performing various inspection and repair techniques: eddy-current testing of tubing; ultrasonic testing of tubing; visual examination of tube ends; profilometry measurements; tube plugging; plug removal; tube extraction; sleeving of tubes; tube end repair; chemical cleaning; and thermal treatment. Plant experience has highlighted the following features of the robot: 1) short installation and demounting periods; 2) installation independent of manhole location; 3) installation possible from outside the steam generator; 4) only one relocation required to address all the tube positions; 5) fast and highly accurate positioning; 6) operational surveillance not required; and 7) drastic reduction of radiation exposure to personnel during repair work

Accident alarm in steam generators in sodium cooled fast reactor power plants. II

International Nuclear Information System (INIS)

Matal, O.; Martoch, J.; Taraba, O.; Hanke, V.

1978-01-01

Conditions were simulated in the economizer of a steam generator of water leaks in sodium at a sodium flow of O.62x10 -3 to 1.24x10 -3 m 3 /s and a sodium temperature of 320 to 380 degC by injecting water at a pressure of 6 to 10 MPa which roughly corresponds to conditions in an economizer of an actual steam generator with leaks within the limits of 0.01 to 0.3 g/s. The leak was recorded by acoustic detectors at all observed sodium flow rates and temperatures. The mean signal-to-noise ratio was in all cases greater than 2. At the assumed 25 dB noise level of the real steam generator of micromodular design it may be assumed that using existing acoustic detectors with waveguides a 0.02 g/s leak of water into sodium may be detected. The measurements showed that the technical standard of the equipment is at least as good as that of the flowmeter system of accident monitoring. (J.B.)

Key findings from the artist project on aerosol retention in a dry steam generator

International Nuclear Information System (INIS)

Dehbi, Abedeloahab; Suckow, Deltef; Lind, Tettaliisa; Guentat, Salih; Danner, Steffen; Mukin, Roman

2016-01-01

A steam generator tube rupture (SGTR) event with a stuck-open safety relief valve constitutes one of the most serious accident sequences in pressurized water reactors (PWRs) because it may create an open path for radioactive aerosol release into the environment. The release may be mitigated by the deposition of fission product particles on a steam generator's (SG's) dry tubes and structures or by scrubbing in the secondary coolant. However, the absence of empirical data, the complexity of the geometry, and the controlling processes have, until recently, made any quantification of retention difficult to justify. As a result, past risk assessment studies typically took little or no credit for aerosol retention in SGTR sequences. To provide these missing data, the Paul Scherrer Institute (PSI) initiated the Aerosol Trapping In Steam GeneraTor (ARTIST) Project, which aimed to thoroughly investigate various aspects of aerosol removal in the secondary side of a breached steam generator. Between 2003 and 2011, the PSI has led the ARTIST Project, which involved intense collaboration between nearly 20 international partners. This summary paper presents key findings of experimental and analytical work conducted at the PSI within the ARTIST program

Key findings from the artist project on aerosol retention in a dry steam generator

Energy Technology Data Exchange (ETDEWEB)

Dehbi, Abedeloahab; Suckow, Deltef; Lind, Tettaliisa; Guentat, Salih; Danner, Steffen; Mukin, Roman [Nuclear Energy and Safety Research Department, Paul Scherrer Institute, Villigen (Switzerland)

2016-08-15

A steam generator tube rupture (SGTR) event with a stuck-open safety relief valve constitutes one of the most serious accident sequences in pressurized water reactors (PWRs) because it may create an open path for radioactive aerosol release into the environment. The release may be mitigated by the deposition of fission product particles on a steam generator's (SG's) dry tubes and structures or by scrubbing in the secondary coolant. However, the absence of empirical data, the complexity of the geometry, and the controlling processes have, until recently, made any quantification of retention difficult to justify. As a result, past risk assessment studies typically took little or no credit for aerosol retention in SGTR sequences. To provide these missing data, the Paul Scherrer Institute (PSI) initiated the Aerosol Trapping In Steam GeneraTor (ARTIST) Project, which aimed to thoroughly investigate various aspects of aerosol removal in the secondary side of a breached steam generator. Between 2003 and 2011, the PSI has led the ARTIST Project, which involved intense collaboration between nearly 20 international partners. This summary paper presents key findings of experimental and analytical work conducted at the PSI within the ARTIST program.

Fourth international seminar on horizontal steam generators

Energy Technology Data Exchange (ETDEWEB)

Tuomisto, H. [ed.] [IVO Group, Vantaa (Finland); Purhonen, H. [ed.] [VTT, Espoo (Finland); Kouhia, V. [ed.] [Lappeenranta Univ. of Technology (Finland)

1997-12-31

The general objective of the International Seminars of Horizontal Steam Generator Modelling has been the improvement in understanding of realistic thermal hydraulic behaviour of the generators when performing safety analyses for VVER reactors. The main topics presented in the fourth seminar were: thermal hydraulic experiments and analyses, primary collector integrity, feedwater distributor replacement, management of primary-to-secondary leakage accidents and new developments in the VVER safety technology. The number of participants, representing designers and manufacturers of the horizontal steam generators, plant operators, engineering companies, research organizations, universities and regulatory authorities, was 70 from 10 countries.

Fourth international seminar on horizontal steam generators

Energy Technology Data Exchange (ETDEWEB)

Tuomisto, H [ed.; IVO Group, Vantaa (Finland); Purhonen, H [ed.; VTT, Espoo (Finland); Kouhia, V [ed.; Lappeenranta Univ. of Technology (Finland)

1998-12-31

The general objective of the International Seminars of Horizontal Steam Generator Modelling has been the improvement in understanding of realistic thermal hydraulic behaviour of the generators when performing safety analyses for VVER reactors. The main topics presented in the fourth seminar were: thermal hydraulic experiments and analyses, primary collector integrity, feedwater distributor replacement, management of primary-to-secondary leakage accidents and new developments in the VVER safety technology. The number of participants, representing designers and manufacturers of the horizontal steam generators, plant operators, engineering companies, research organizations, universities and regulatory authorities, was 70 from 10 countries.

Fourth international seminar on horizontal steam generators

International Nuclear Information System (INIS)

Tuomisto, H.; Purhonen, H.; Kouhia, V.

1997-01-01

The general objective of the International Seminars of Horizontal Steam Generator Modelling has been the improvement in understanding of realistic thermal hydraulic behaviour of the generators when performing safety analyses for VVER reactors. The main topics presented in the fourth seminar were: thermal hydraulic experiments and analyses, primary collector integrity, feedwater distributor replacement, management of primary-to-secondary leakage accidents and new developments in the VVER safety technology. The number of participants, representing designers and manufacturers of the horizontal steam generators, plant operators, engineering companies, research organizations, universities and regulatory authorities, was 70 from 10 countries

Dynamic analysis of CHASNUPP steam generator structure during shipping

International Nuclear Information System (INIS)

Han Liangbi; Xu Jinkang; Zhou Meiwu; He Yinbiao

1998-07-01

The dynamic analysis of CHASNUPP steam generator during shipping is described, including the simplified mathematical model, acceleration power spectrum of ocean wave induced random vibration, the dynamic analysis of steam generator structure under random loading, the applied computer code and calculated results

Hideout of sea water impurities in steam generator tube deposits: laboratory and field studies

International Nuclear Information System (INIS)

Balakrishnan, P.V.; Turner, C.W.; Thompson, R.; Sawochka, S.

1996-01-01

Sea water impurities hide out within thin (∼10 μm) deposits on steam generator tubes, as demonstrated by both laboratory studies using segments of fouled steam generator tubes pulled in 1992 from Crystal River-3 nuclear power station and field hideout return studies performed during recent plant shutdowns. Laboratory tests performed at 279 o C (534 o F) and heat fluxes ranging from 35 to 114 kW/m 2 (11,100 - 36,150 Btu/h.ft 2 ), conditions typical of the lower tubesheet to the first support plate region of a once-through steam generator, showed that impurity hideout can occur in thin free-span tube deposits. The extent of hideout increased with increasing heat flux. Soluble species, such as sodium and chloride ions, returned promptly to the bulk water from the deposits when the heat flux was turned off, whereas less soluble species, such as calcium sulfate and magnesium hydroxide, returned more slowly. Recent field hideout return studies performed at Crystal River-3 where the water level in the steam generators was maintained below the first tube support plate during the shutdown, thus wetting only the thin deposits in the free span and the small sludge pile, corroborate the laboratory findings, showing that hideout does indeed occur in the free-span regions of the tubes. These findings suggest that hideout within tube deposits has to be accounted for in the calculation of crevice chemistry from hideout return studies and in controlling the bulk chemistry using the molar ratio criterion. (author). 3 refs., 4 tabs., 3 figs

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.

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

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

Study on steam separation in steam generators of a NPP with the WWER-440 reactors

International Nuclear Information System (INIS)

Dmitriev, A.I.; Kolzov, Yu.V.; Titov, V.F.; Dubrovin, A.V.; Ilyushin, V.F.; Volkov, A.P.

1977-01-01

The separation characteristics as well as the actual level position in steam generators with and without a submerged holy sheet have been determined at a WWER-440 reactor nuclear power plant. It has been shown, that without changing the design of steam generators their load at the WWER-440 reactor nuclear power plant can be increased by about 10%. In this case the vapour humidity does not exceed the permissible value equal to 0.25%. The submerged holy sheet considerably decreases load irregularity and swelling of the water-steam mixture layer

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

Steam generators and furnaces

Energy Technology Data Exchange (ETDEWEB)

Swoboda, E

1978-04-01

The documents published in 1977 in the field of steam generators for conventional thermal power plants are classified according to the following subjects: power industry and number of power plants, planning and operation, design and construction, furnaces, environmental effects, dirt accumulation and corrosion, conservation and scouring, control and automation, fundamental research, and materials.