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.

An effect of corium composition variations on occurrence of a steam explosion in the TROI experiments

International Nuclear Information System (INIS)

Kim, J. W.; Park, I. K.; Hong, S. W.; Min, B. T.; Shin, Y. S.; Song, J. H.; Kim, H. D.

2003-01-01

Recently series of steam explosion experiments have been performed in the TROI facility using corium melts of various compositions. The compositions (UO 2 : ZrO 2 ) of the corium were 0 : 100, 50 : 50, 70 : 30, 80 : 20 and 87 : 13 in weight percent and the mass of the corium was about 10kg. An experiment using 0 : 100 corium (pure zirconia) caused a steam explosion. An experiment using 50 : 50 corium did not cause a steam explosion while a steam spike occurred in an experiment using 70 : 30 corium which was the eutectic point of corium. A steam spike is considered to be the fact that a triggering of a steam explosion occurred but a propagation process does not occur so as to cause a weak interaction. However, the possibility of a steam explosion with this composition can not be ruled out since many steam explosions occurred in the previous experiments. In the two experiments using 80 : 20 corium, a steam spike occurred in one experiment but no steam explosion occurred in the other experiment. However, the triggerability of a steam explosion with this composition is not clear since few steam explosions occurred in the previous experiments. And no steam explosion occurred in an experiment using 87 : 13 corium of which urania content was the greatest among the experiments performed in the TROI facility. From this, the possibility of a steam explosion or a steam spike is appeared to be high in the non-mush zone. It is considered that an explosive interaction could easily occur with the eutectic composition. Since the solidification temperature around the eutectic point is low, the melt is likely to maintain its liquid state at the time of triggering so as to cause an explosive phenomenon

Flow oscillations on the steam control valve in the middle opening condition. Clarification of the phenomena by steam flow experiment and CFD calculation

International Nuclear Information System (INIS)

Morita, Ryo; Inada, Fumio

2006-01-01

A steam control valve might cause vibrations of piping when the valve opening is in a middle condition. For rationalization of maintenance and management of the plant, the valve should be improved, but it is difficult to understand flow characteristics in detail by experiment because flow around the valve is complex 3D structure and becomes supersonic (M>1). Therefore, it is necessary to clarify the cause of the vibrations and to develop improvements by Computational Fluid Dynamics (CFD) technology. In previous researches, we clarified a mechanism of the pressure fluctuations in the middle opening condition and suggested the way to prevent the pressure fluctuations by experiments and CFD calculations. But, as we used air as a working fluid in our previous research instead of steam that is used in the power plant, we couldn't consider effects of condensation and difference of change of the quantity of state between air and steam. In this report, we have conducted steam flow experiments by multi-purpose steam experiment apparatus 'WISSH' and CFD calculations by steam flow code 'MATIS-SC' to clarify those effects. As a result, in the middle opening condition, we have observed rotating pressure fluctuations in the experiment and valve-attached flow and local high-pressure region in the CFD result. These results show the pressure fluctuations in steam experiments and CFD is same kind of the fluctuations found in air experiment and CFD. (author)

New Brunswick Market Design Committee : First interim report

International Nuclear Information System (INIS)

2001-10-01

In June 2001, a Market Design Committee was appointed by the Government of New Brunswick tp provide some advice on the best ways to implement the electricity restructuring initiatives detailed in a White Paper on energy policy in New Brunswick. The Market Design Committee is a very technical committee that was set up with a number of goals in mind: make recommendations concerning all codes and operating protocols, make recommendations on market surveillance and the establishment of a workable competitive market, take into account the reliability of supply for New Brunswick, address ways by which to avoid rate shock to existing self-generators, make recommendations for mitigation of market power in the wholesale and large retail markets, evaluate methods of stranded cost recovery, make recommendations on the requirement for reciprocity in New Brunswick electricity market design, and review the role and treatment of small-scale, on-site electricity generation and make recommendations. The final recommendations are set to be presented in April 2002. This document details the progress realized to date and outlines the plans made for the subsequent phases of the work. The different topics discussed in this document each form the basis for one section of the report. They are: market design committee constitution and process, basic market model, market power issues, transmission issues, environment and renewables, and next steps for the market design committee

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)

Geology of uranium and associated elements in New Brunswick

International Nuclear Information System (INIS)

Hassan, H.H.; Hale, W.E.; Chrzanowski, M.

1987-01-01

Eighty-two uranium occurrences have been identified in New Brunswick. Most of the data describing these occurrences is contained within assessment reports by claim holders to the provincial Department of Natural Resources and Energy. Additional information is contained within publications of the Geological Survey of Canada and in university theses. Seven metallogenic uraniferous domains are identified in New Brunswick: the Gaspe Synclinorium, Aroostook-Matapedia Anticlinorium, Chaleur Bay Synclinorium, Miramichi Anticlinorium, Fredericton Trough, Avalonian Platform and the Carboniferous Basin. Of these seven domains, the Miramichi Anticlinorium, Fredericton Trough and the Carboniferous Basin appear to be the most promising. The uranium occurrences in New Brunswick are preferentially associated with the Devono-Carboniferous rocks and appear to be related to deformational, magmatic and hydrothermal processes associated with the Acadian Orogeny. (numerous refs)

Design and operating experiences with 50MW steam generator

International Nuclear Information System (INIS)

Kawara, M.; Yamaki, H.; Kanamori, A.; Tanaka, K.; Takahashi, T.

1975-01-01

The main purpose of the 50 MW steam generator is to have experiences of manufacturing and operation with large scale steam generator including necessary research and development works which can be reflected on the design and fabrication of 'Monju' (Japan 300 MWe prototype LMFBR). The detailed design of the 50 MW steam, generator was begun on March, 1972 and succeeded in the demonstration of 72 hours continuous operation with full power on June, 1974. It has been successfully operated since then, the performances of which have been evaluated through various kinds of tests. In this paper, the following items are mainly discussed system design, thermal and hydraulic design, structure and fabrication and some experiences on testing operation including cleaning and sodium flushing of equipment, sodium level control system, the behavior of hydrogen detection system and general outlook of the performance. (author)

Design and operating experiences with 50MW steam generator

Energy Technology Data Exchange (ETDEWEB)

Kawara, M; Yamaki, H; Kanamori, A; Tanaka, K; Takahashi, T

1975-07-01

The main purpose of the 50 MW steam generator is to have experiences of manufacturing and operation with large scale steam generator including necessary research and development works which can be reflected on the design and fabrication of 'Monju' (Japan 300 MWe prototype LMFBR). The detailed design of the 50 MW steam, generator was begun on March, 1972 and succeeded in the demonstration of 72 hours continuous operation with full power on June, 1974. It has been successfully operated since then, the performances of which have been evaluated through various kinds of tests. In this paper, the following items are mainly discussed system design, thermal and hydraulic design, structure and fabrication and some experiences on testing operation including cleaning and sodium flushing of equipment, sodium level control system, the behavior of hydrogen detection system and general outlook of the performance. (author)

New Coleoptera records from New Brunswick, Canada: Lycidae

Directory of Open Access Journals (Sweden)

Reginald Webster

2012-04-01

Full Text Available Eight species of Lycidae are newly recorded from New Brunswick, Canada, bringing the total number of species known from the province to 16. The first documented records from New Brunswick are provided for Greenarius thoracicus (Randall Erotides scuptilis (Say, and Calopteron terminale (Say reported by Majka et al. (2011. Eropterus arculus Green, Lopheros crenatus (Germar, and Calochromus perfacetus (Say are reported for the first time in the Maritime provinces. Collection data, habitat data, and distribution maps are presented for all these species.

Advisory panel on the proposed New Brunswick - Quebec electricity transaction

International Nuclear Information System (INIS)

Ganong, D.A.

2010-01-01

NB Power was faced with major financial and technical challenges in generation that would lead to significant rate increases over several years to come. In 2009, the governments of New Brunswick and Quebec proposed an agreement whereby Hydro-Quebec would acquire most of the assets of NB Power and provide a guaranteed electricity supply at a set price to the province. The proposal involved the sale of NB Power's hydro generation facilities, 2 diesel peaking plants and the Point Lepreau nuclear generating facility. It also included a guaranteed bulk electricity supply of 14 Terawatt-hours per year by Hydro-Quebec at a blended rate of 7.35 cents per kilowatt-hour. The new Brunswick government would retain ownership of the fossil-fuelled power plants, the NB Power transmission assets and the NB Power distribution assets and customer sales. A 6-person independent panel was asked to examine the proposal. This report addresses the proposed agreement as presented to the Panel up to January 26, 2010. The panel conducted an independent and objective examination of the financial implications; long-term effects on electricity rates; New Brunswick's control of energy policies; environmental impacts; short-and long-term risks and avoided risks; and benefits and costs to the provincial economy. The Panel drew its conclusions and recommendations on the basis of available information. The Panel concluded that the benefits to New Brunswick contribute to real and positive value to New Brunswick over business as usual. tabs., figs.

Preliminary condensation pool experiments with steam using DN80 and DN100 blowdown pipes

International Nuclear Information System (INIS)

Laine, J.; Puustinen, M.

2004-03-01

The report summarizes the results of the preliminary steam blowdown experiments. Altogether eight experiment series, each consisting of several steam blows, were carried out in autumn 2003 with a scaled-down condensation pool test rig designed and constructed at Lappeenranta University of Technology. The main purpose of the experiments was to evaluate the capabilities of the test rig and the needs for measurement and visualization devices. The experiments showed that a high-speed video camera is essential for visual observation due to the rapid condensation of steam bubbles. Furthermore, the maximum measurement frequency of the current combination of instrumentation and data acquisition system is inadequate for the actual steam tests in 2004. (au)

Leakage experiences with 1 MW steam generator

International Nuclear Information System (INIS)

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

1975-01-01

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

Transforming the New Brunswick Energy Hub: An Analysis on Renewable Energy

Science.gov (United States)

Gunter, Christopher

This research examines the benefits and disadvantages of instituting a shift from fossil fuel dependence to renewable sources of energy in New Brunswick. The New Brunswick Energy Hub is a complex system acting under the mandate of the White Paper New Brunswick Energy Policy. In my research, I consider information derived from statistical indicators developed by Patlitzianas, Doukas, Kagiannas and Psarras (2008) and compare these findings to the efficacy of energy policies in Germany, Denmark and Spain. These countries are similar to New Brunswick in climate and organizational complexity (US Department of Commerce, 2009). Weighing the outcomes of this comparative study, I discuss my recommendations highlighting the environmental and economic benefits. My research investigates subsidies in each country that allowed them early economic and environmental advantages. Specific regional considerations, such as Denmark's trend of selling energy technology for profit over domestic applications, inform my conclusions. The future New Brunswick Energy Policy should focus on creating favorable conditions for renewable energy development to occur. Some proven conditions include infrastructure development subsidies and the development and annual review of a competitive open access transmission tariff. With the expiry of the current White Paper comes the necessity of this investigation, and the opportunity to address the growing financial and environmental concerns that many politicians and policy planners have failed to deal with in past policies.

Extended layup of steam generators during a refurbishment outage

International Nuclear Information System (INIS)

Marks, C.R.; Little, M.D.; Slade, J.; Gendron, T.

2009-01-01

In May 2008, Point Lepreau Generating Station (PLGS), owned and operated by New Brunswick Power Nuclear (NBPN), entered an extended refurbishment outage initially expected to last approximately 18 months. NBPN had the two inter-related goals with respect to layup of the steam generators during this period: equipment preservation and inspection interval modification. The steam generators were to be preserved such that there was no loss of operating life due to corrosion of either the tubing (Alloy 800NG) or other internal components (with carbon steel being the limiting material with respect to corrosion). Additionally, NBPN desired that the time in layup not count as operating time in setting the schedule for future inspections. That is, a key goal of the steam generator layup is that the future inspection interval be based on operating time, not calendar time. The NBPN approach consists of the following four steps: A review of industry operating experience with long outages (including both PWRs and PHWRs); The development of technically based layup strategies and procedures; A mid-outage review of the implementation of the layup strategies and procedures; and A post-outage review to determine if the actual conditions in the steam generators will support modification of the inspection interval. This paper discusses the results of the first three of these steps. At this time, the plant is still in the refurbishment outage. Throughout the outage evaluation process, the following issues have been the main focus of the reviews: The potential for degradation (pitting and cracking) of steam generator tubes; The potential for general corrosion of carbon and low alloy steel internals; Oxidation of deposits (which could subsequently lead to oxidizing conditions during operation, possibly leading to tube degradation). This paper discusses the industry operating experience reviewed, the pre-outage assessments, and the mid-outage assessments. Current outage planning places the

Wet steam turbines for nuclear generating stations -design and operating experience

International Nuclear Information System (INIS)

Usher, J.

1977-01-01

Lecture to the Institution of Nuclear Engineers, 11 Jan. 1977. The object of this lecture was to give an account of some design features of large wet steam turbines and to show by describing some recent operational experience how their design concepts were fulfilled. Headings are as follows: effects of wet steam cycle on turbine layout and operation (H.P. turbine, L.P. turbine); turbine control and operation; water separators; and steam reheaters. (U.K.)

Physician characteristics and prescribing for elderly people in New Brunswick: relation to patient outcomes.

OpenAIRE

Davidson, W; Molloy, D W; Bédard, M

1995-01-01

OBJECTIVE: To examine the relation between physician characteristics, prescribing behaviour and patient outcomes. DESIGN: Descriptive study linking four provincial databases. SETTING: New Brunswick. PARTICIPANTS: All 366 general practitioners (GPs) (accounting for 40% of all physicians with a general licence in New Brunswick) who ordered at least 200 prescriptions for elderly beneficiaries of the New Brunswick Prescription Drug Program and saw at least 20 elderly patients in an office setting...

Design, development and operating experience with wet steam turbines

International Nuclear Information System (INIS)

Bolter, J.R.

1989-01-01

The paper first describes the special characteristics of wet steam units. It then goes on to discuss the principal features of the units manufactured by the author's company, the considerations on which the designs were based, and the development work carried out to validate them. Some of the design features such as the separator/reheater units and the arrangements for water extraction in the high pressure turbine are unconventional. An important characteristic of all nuclear plant is the combination of high capital cost and low fuel cost, and the consequent emphasis placed on high availability. The paper describes some service problems experienced with wet steam plant and how these were overcome with minimum loss of generation. The paper also describes a number of the developments for future wet steam plant which have evolved from these experiences, and from research and development programmes aimed at increasing the efficiency and reliability of both conventional and wet steam units. Blading, rotor construction and separator/reheater units are considered. (author)

Operating experience of the EBR-II steam generating system

International Nuclear Information System (INIS)

Buschman, H.W.; Penney, W.H.; Quilici, M.D.; Radtke, W.H.

1981-01-01

The Experimental Breeder Reactor II (EBR-II) is a Liquid Metal Fast Breeder Reactor (LMFBR) with integrated power producing capability. Superheated steam is produced by eight natural circulation evaporators, two superheaters, and a conventional steam drum. Steam throttle conditions are 438 C (820 F) and 8.62 MPa (1250 psi). The designs of the evaporators and superheaters are essentially identical; both are counterflow units with low pressure nonradioactive sodium on the shell side. Safety and reliability are maximized by using duplex tubes and tubesheets. The performance of the system has been excellent and essentially trouble free. The operating experience of EBR-II provides confidence that the technology can be applied to commercial LMFBR's for an abundant supply of energy for the future. 5 refs

78 FR 14300 - Proposed CERCLA Settlement Relating to the 1244 White Drive Site in North Brunswick, Middlesex...

Science.gov (United States)

2013-03-05

... Drive Site in North Brunswick, Middlesex County, NJ AGENCY: Environmental Protection Agency (EPA... 1244 White Drive Site (``Site''), located in North Brunswick, Middlesex County, New Jersey. Under the... 10007-1866. Comments should reference the 1244 White Drive Site, located in North Brunswick, Middlesex...

New Brunswick electricity market rules : summary

International Nuclear Information System (INIS)

2004-02-01

The electricity market rules for New Brunswick were reviewed with particular reference to two broad classifications. The first classification is based on the roles and responsibilities of the system operator (SO) in facilitating the Bilateral Contract market, as well as the role of market participants in participating in the Bilateral Contract market. The second classification is based on the roles and responsibilities of each of the SO, market participants and transmitters in maintaining the reliability of the integrated electricity system and ensuring a secure supply of electricity for consumers in New Brunswick. The market rules consist of 10 chapters entitled: (1) introduction to the market rules and administrative rules of general application, (2) market participation and the use of the SO-controlled grid, (3) market administration, (4) technical and connection requirements, testing and commissioning, (5) system reliability, (6) operational requirements, (7) settlement, (8) connection of new or modified facilities, (9) transmission system planning, investment and operation, and (10) definitions and interpretation

New Coleoptera records from New Brunswick, Canada: Cerambycidae.

Science.gov (United States)

Webster, Reginald P; Sweeney, Jon D; Demerchant, Ian; Silk, Peter J; Mayo, Peter

2012-01-01

Five species of Cerambycidae, Acmaeops discoideus (Haldeman), Anelaphus villosus (Fabricius), Phymatodes species (CNC sp. n. #1), Sarosesthes fulminans (Fabricius), and Urgleptus signatus (LeConte) are newly recorded for New Brunswick, Canada. All but Acmaeops villosus are new to the Maritime provinces. Phymatodes testaceus (Linnaeus) is removed from the faunal list of the province as a result of mislabeled specimens, records of Phymatodes maculicollis LeConte are presented confirming the presence of this species in New Brunswick, and the first recent records ofNeospondylis upiformis (Mannerheim) are presented. Additional records are given for the recently recorded Phymatodes aereus (Newman), indicating a wider distribution in the province. Collection data, habitat data, and distribution maps are presented for each species.

A model for successful research partnerships: a New Brunswick experience.

Science.gov (United States)

Tamlyn, Karen; Creelman, Helen; Fisher, Garfield

2002-01-01

The purpose of this paper is to present an overview of a partnership model used to conduct a research study entitled "Needs of patients with cancer and their family members in New Brunswick Health Region 3 (NBHR3)" (Tamlyn-Leaman, Creelman, & Fisher, 1997). This partial replication study carried out by the three authors between 1995 and 1997 was a needs assessment, adapted with permission from previous work by Fitch, Vachon, Greenberg, Saltmarche, and Franssen (1993). In order to conduct a comprehensive needs assessment with limited resources, a partnership between academic, public, and private sectors was established. An illustration of this partnership is presented in the model entitled "A Client-Centred Partnership Model." The operations of this partnership, including the strengths, the perceived benefits, lessons learned by each partner, the barriers, and the process for conflict resolution, are described. A summary of the cancer care initiatives undertaken by NBHR3, which were influenced directly or indirectly by the recommendations from this study, is included.

Preliminary condensation pool experiments with steam using DN80 and DN100 blowdown pipes[VIDEO CAMERAS

Energy Technology Data Exchange (ETDEWEB)

Laine, J.; Puustinen, M. [Lappeenranta University of Technology (Finland)

2004-03-01

The report summarizes the results of the preliminary steam blowdown experiments. Altogether eight experiment series, each consisting of several steam blows, were carried out in autumn 2003 with a scaled-down condensation pool test rig designed and constructed at Lappeenranta University of Technology. The main purpose of the experiments was to evaluate the capabilities of the test rig and the needs for measurement and visualization devices. The experiments showed that a high-speed video camera is essential for visual observation due to the rapid condensation of steam bubbles. Furthermore, the maximum measurement frequency of the current combination of instrumentation and data acquisition system is inadequate for the actual steam tests in 2004. (au)

New Brunswick Site annual environmental report for calendar year 1991, New Brunswick, New Jersey

International Nuclear Information System (INIS)

1992-09-01

This document describes the environmental monitoring program at the New Brunswick Site (NBS) and surrounding area, implementation of the program, and monitoring results for 1991. The site, near New Brunswick,, New Jersey, is a 5.6-acre vacant, fenced, and grass-covered area. Environmental monitoring of NBS began in 1981 when the site was part of the US Department of Energy's (DOE) Surplus Facilities Management Program. In 1990 responsibility for NBS was transferred to the Formerly Utilized Sites Remedial Action Program (FUSP.4P). FUSRAP is a DOE program to identify and decontaminate or otherwise control sites where residual radioactive materials remain from the,early years of the nation's atomic energy program or from commercial operations causing conditions that Congress has authorized DOE to remedy. The environmental monitoring program at NBS includes sampling networks for radon and thoron in air; external gamma radiation exposure; and radium-226, radium-228, thorium-228, thorium-230, thorium-232, americium-241, cesium-137, plutonium-239, and total uranium in surface water, sediment, and groundwater. Several nonradiological parameters are also measured in groundwater, surface water, and sediments. Monitoring results are compared with applicable Environmental Protection Agency standards, DOE derived concentration guides, dose limits, and other requirements in DOE orders. Environmental standards are established to protect public health and the environment

New Coleoptera records from New Brunswick, Canada: Cerambycidae

Directory of Open Access Journals (Sweden)

Reginald Webster

2012-04-01

Full Text Available Five species of Cerambycidae, Acmaeops discoideus (Haldeman, Anelaphus villosus (Fabricius, Phymatodes species (CNC sp. n. #1, Sarosesthes fulminans (Fabricius, and Urgleptus signatus (LeConte are newly recorded for New Brunswick, Canada. All but A. villosus are new to the Maritime provinces. Phymatodes testaceus (Linnaeus is removed from the faunal list of the province as a result of mislabeled specimens, records of Phymatodes maculicollis LeConte are presented confirming the presence of this species in New Brunswick, and the first recent records of Neospondylis upiformis (Mannerheim are presented. Additional records are given for the recently recorded Phymatodes aereus (Newman, indicating a wider distribution in the province. Collection data, habitat data, and distribution maps are presented for each species.

Experiment Operating Specification for the Semiscale MOD-2C feedwater and steam line break experiment series. Appendix S-FS-6 and 7

International Nuclear Information System (INIS)

Boucher, T.J.; Owca, W.A.

1985-05-01

This document is the Semiscale MOD-2C feedwater and steam line break experiment series Experiment Operating Specification Appendix for tests S-FS-6 and S-FS-7. Test S-FS-6 is the third test in the series and simulates a 100% break in a steam generator bottom feedwater line downstream of the check valve accompanied by compounding factors (such as check valve failure, loss-of-offsite power at SIS and SIS delayed until low steam generator pressure signal). The test is terminated after plant stabilization and recovery procedures including unaffected loop steam and feed, pressurizer heater operation, pressurizer auxiliary spray operation, and normal charging/letdown operation. Test S-FS-7 is the fourth test in the series and simulates a 14.3% break in a steam generator bottom feedwater line downstream of the check valve, accompanied by compounding factors. The test is terminated after plant stabilization procedures including unaffected loop steam and feed, pressurizer heater operation, and normal charging/letdown operation. The test was followed by an affected loop secondary refill after isolating the break. The Appendix contains information on the major fluid systems, initial experiment conditions, experiment boundary conditions, and sequence of experiment events. Also included is a discussion of the scaling criteria and philosophy used to develop the experiment initial and boundary conditions and system configuration

Flow oscillations on the steam control valve in the middle opening condition. Clarification of the effects of valve body and valve seat by steam experiments

International Nuclear Information System (INIS)

Morita, Ryo; Inada, Fumio

2007-01-01

A steam control valve might cause vibrations of piping when the valve opening is in a middle condition. For rationalization of maintenance and management of the plant, the valve should be improved, but it is difficult to understand flow characteristics in detail by experiment because flow around the valve is complex 3D structure and becomes supersonic (M>1). Therefore, it is necessary to clarify the cause of the vibrations and to develop the countermeasures by CFD (Computational Fluid Dynamics) technology. In previous researches, we clarified a mechanism of the pressure fluctuations in the middle opening condition and suggested the new valve shape (named 'Extended Valve') that can suppress the pressure fluctuations by air experiments and CFD calculations. Then, we also conducted steam experiments and CFD calculations to understand the differences between air and the steam, and found that the pressure fluctuations in the middle opening condition also occurred in the steam tests and the differences between the air and steam were not remarkable. In this report, to clarify the effects of valve and valve seat shape in steam flow condition, we conduct the steam experiments with various valve and seat shape. As a result, we find the change of the valve seat can decrease the amplitude of pressure fluctuations, but can not quite suppress the pressure fluctuations in the middle opening condition. Then, we apply the 'Extended Valve' to clarify the valve shape effect, and find that the extended valve suppresses the pressure fluctuations in the middle opening condition completely and decreases the pressure amplitude drastically. (author)

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)

New Coleoptera records from New Brunswick, Canada: Gyrinidae, Carabidae, and Dytiscidae

Directory of Open Access Journals (Sweden)

Reginald Webster

2012-04-01

Full Text Available Dineutus assimilis Kirby and Dineutus discolor Aubé of the Family Gyrinidae are newly reported from New Brunswick, Canada. Four species of Carabidae, Agonum (Agonum piceolum (LeConte, Bembidion (Pseudoperyphus rufotinctum Chaudoir, Harpalus (Harpalus opacipennis (Haldeman, and Pterostichus (Melanius castor Goulet & Bousquet are newly reported from New Brunswick and the Maritime provinces, and one species of Dytiscidae, Liodessus noviaffinis Miller, is newly recorded for the province. Collection, habitat data, and distribution maps are presented for each species.

HTGR steam generator development

International Nuclear Information System (INIS)

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

1976-01-01

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

Large-scale wind power in New Brunswick : a regional scenario study towards 2025

International Nuclear Information System (INIS)

2008-08-01

This paper discussed the large-scale development of wind power in New Brunswick and evaluated Danish experiences with wind development as a template for developing wind resources in the Maritimes region. The study showed that New Brunswick and the Maritimes region have good wind resources, and that the province will gain significant economic benefits from deploying between 5500 and 7500 MW of wind power capacity by 2025. Wind power development will contribute to the security of supply in the region and reduce air pollution. Carbon regulation and renewable portfolio standards will improve the competitiveness of wind power. Electricity generated by wind power plants in the Maritimes can be sold to other provinces in Canada, as well as to the heavily populated New England region of the United States. A high level of cooperation between markets in the Maritimes area and neighbouring New England and Quebec systems will be required in addition to load flow analyses of electricity systems. Denmark's experiences with developing wind power indicate that existing market designs must be restructured to allow for higher levels of competition. A strong system operator is required to integrate wind power into the system. It was concluded that strong political leadership is required to ensure the sustainable development of the region. 5 refs., 4 tabs., 9 figs

Steam generator tube failures: experience with water-cooled nuclear power reactors during 1976

International Nuclear Information System (INIS)

Tatone, O.S.; Pathania, R.S.

1978-02-01

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

Letter from Kirk Kessler, EPS to Galo Jackson, USEPA. Subject: Results of the July 2011 Sampling in the Former Brunswick-Altamaha Canal South of the LCP Chemicals Site, Brunswick, Ga, Revision Dated April 19, 2012

Science.gov (United States)

Results report of sediment and fish tissue samples collected from the former Brunswick-Altamaha Canal, in a segment of the canal south of the LCP Chemicals Site in Brunswick, Georgia in July 2010. Region ID: 04 DocID: 10843428, DocDate: 04-19-2012

Proceedings of the NEA/CSNI-UNIPEDE Specialist Meeting on Operating Experience with Steam Generators

Energy Technology Data Exchange (ETDEWEB)

NONE

1991-07-01

The long history of operating experience with pressurized water reactors has indicated that the steam generators are of primary importance in nuclear power plant design and operation; this is furthermore confirmed by analyzing the data of the Incident Reporting System (IRS). It is for this reason that the OECD/NEA Committee on the Safety of Nuclear Installations organizes, in cooperation with UNIPEDE, a Specialist Meeting on 'Operating Experience with Steam Generators'. This Specialist Meeting, held in Brussels, Belgium, in September 1991, is hosted by the Belgian Government and AIB-Vincotte Nuclear. In addition to being a follow-up to the October 1984 meeting (organized by the CSNI and UNIPEDE in Stockholm, Sweden), this Meeting reviews the current state-of-the-art of steam generator technology thus providing a forum for the exchange of related experience in operation, inspection, maintenance, repair, modifications, replacement, and licensing requirements pertaining to steam generators. Forty-seven papers are presented in eight sessions entitled: Operating Experience (two sessions), Structural Integrity and Licensing Issues, Analysis and Prediction of Degradation Mechanisms, Inservice Inspection Methods, Preventive and Corrective Actions (two sessions) and Replacement of Steam Generators. There are furthermore two panel sessions entitled 'Observed Degradation Mechanisms and Licensing Positions', and 'Inspection, Repair and Replacement Strategies'. These proceedings consist of a compilation of the papers presented at the Meeting, which is attended by more than one hundred and fifty participants from fifteen countries and several international organisations.

Recent operating experiences with steam generators in Japanese NPPs

International Nuclear Information System (INIS)

Yashima, Seiji

1997-01-01

In 1994, the Genkai-3 of Kyushu Electric Power Co., Inc. and the Ikata-3 of Shikoku Electric Power Co., Inc. started commercial operation, and now 22 PWR plants are being operated in Japan. Since the first PWR plant now 22 PWR plants are being operated in was started to operate, Japanese PWR plants have had an operating experience of approx. 280 reactor-years. During that period, many tube degradations have been experienced in steam generators (SGs). And, in 1991, the steam generator tube rupture (SGTR) occurred in the Mihama-2 of Kansai Electric Power Co., Inc. However, the occurrence of tube degradation of SGs has been decreased by the instructions of the MITI as regulatory authorities, efforts of Electric Utilities, and technical support from the SG manufacturers. Here the author describes the recent SGs in Japan about the following points. (1) Recent Operating Experiences (2) Lessons learned from Mihama-2 SGTR (3) SG replacement (4) Safety Regulations on SG (5) Research and development on SG

Challenge and opportunity: developing a gas distribution industry in New Brunswick

Energy Technology Data Exchange (ETDEWEB)

Pleckaitis, A.J. [Enbridge Gas New Brunswick, Fredericton, NB (Canada)

2001-07-01

The largest gas distributor in Canada is Enbridge, which also operates the longest oil pipeline system in the world. In this presentation, the author discussed the challenges facing Enbridge Gas New Brunswick (EGNB). The limited partnership, the 20-year renewable franchise and the business plan was briefly examined. The state of development of EGNB was described with the back-bone distribution system having been completed and the first customers attached. EGNB is in the process of developing the supporting infrastructure. A map of the Eastern region of the country was shown, followed by a map of New Brunswick displaying the work accomplished to date. A brief overview of the business model was provided and the challenges identified. The customer value chain was illustrated, and a graph showed the natural gas price advantage in New Brunswick. Next, the author discussed the situation in Nova Scotia, where the gas distribution franchise was awarded in December 1999 to Sempra Atlantic Gas. Looking to the future, the author mentioned supply and regional strategy. The final thought was for greater cooperation between government and industry. figs.

School Psychology in New Brunswick in 2016

Science.gov (United States)

Mureika, Juanita

2016-01-01

School psychology in New Brunswick experienced a surge of growth and development in the early part of the 21st century; however, dwindling numbers and recent government initiatives are presenting serious challenges to our ability to continue to provide the quality tiered services that we want to offer to the school community.

Steam sterilization does not require saturated steam

NARCIS (Netherlands)

van Doornmalen Gomez Hoyos, J. P.C.M.; Paunovic, A.; Kopinga, K.

2017-01-01

The most commonly applied method to sterilize re-usable medical devices in hospitals is steam sterilization. The essential conditions for steam sterilization are derived from sterilization in water. Microbiological experiments in aqueous solutions have been used to calculate various time–temperature

New Brunswick Power Nuclear Corporation update 2005

International Nuclear Information System (INIS)

White, R.M.; Eagles, E.R.; Pilkington, W.S.

2005-01-01

A brief presentation will be made on the operations and business activities over the previous year with a discussion of the current status of the NB Power group of companies. The New Brunswick Government has implemented the new 'Electricity Act' which has resulted restructuring of NB Power, opening of the electricity market to wholesale completion and the separation of the transmission system operation from NB Power. On October 1, 2005 the restructuring of NB Power was implemented to change NB Power from a single integrated utility into NB Power Holding Corporation with four subsidiary operating companies including NB Power Nuclear Corporation, NB Power Distribution and Customer Service Corporation, NB Power Transmission Corporation and NB Power Generation Corporation. As part of the Electricity Act, the transmission system reliability, operation and market control functions have been moved into a separate company, the NB System Operator, outside of the NB Power group of companies. A review of Point Lepreau's operational activities will include presentation of the capacity factor, availability and safety results with a summary of significant issues, planned outages and unplanned outages. An update on the current status of Power Reactor Operating License renewal and the strategies for renewal timing will be presented. Planning for refurbishment has continued with a major focus on addressing the recommendations made by Dr. Robin Jeffrey in his report to the Province of New Brunswick. These recommendations included three options for replacement of the Point Lepreau capacity and energy; 1) improve refurbishment contract arrangements with AECL, 2) solicit external investment in refurbishment and the station and 3) update the case for fossil fuel alternatives. The NB Power Holding Corporation Board of Directors have provided the appropriate information on the options to the owner (Province of New Brunswick) for consideration. A decision on the future of the Point Lepreau

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.

New safety experiments in decommissioned superheated steam reactor at Karlstein

International Nuclear Information System (INIS)

Koerting, K.

1986-01-01

This article gives a concise summary of the Status Report of the Superheated Steam Reactor Safety Program (PHDR) Project, held at KfK on Dec. 5, 1985. The results discussed dealt with fire experiments, shock tests simulating airplane crashes, temperature shocks in the reactor pressure vessel, studies of crack detection in pressure vessels and blasting experiments associated with nuclear plant decommissioning

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.

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.)

The preliminary results of steam explosion experiments in TROI

International Nuclear Information System (INIS)

Song, J.H.; Park, I.K.; Chang, Y.J.; Min, B.T.; Hong, S.W.; Kim, H.D.

2001-01-01

Korea Atomic Energy Research Institute (KAERI) launched an intermediate scale steam explosion experiment named 'Test for Real corium Interaction with water (TROI)' using reactor material to investigate the effect of material composition, multi-dimensional melt-water interaction, and hydrogen generation. The melt-water interaction is confined in a pressure vessel with the multi-dimensional fuel and water pool geometry. The cold crucible technology, where the mixture of oxide powder in a water-cooled cage is heated by high frequency induction, is employed. It minimizes unwanted inclusion of impurities during the melting process. The data acquisition system and instrumentations which measure the static and dynamic pressure, temperatures of melt and water are set up. In the first series of tests using several kg of ZrO 2 , melt water interaction is made in a heated water pool at 95 Celsius degrees without triggering. A steam spike pressure at about 10 bar is observed. The morphology of debris shows that there was a mild local steam explosion. The melt water interaction was monitored by video cameras. The UO 2 tests are scheduled around March of 2001, in parallel with the improvements of the design of test facility. (authors)

Nuclear power in New Brunswick

International Nuclear Information System (INIS)

O'Connor, A.J.

1984-06-01

New Brunswick Power is a medium-utility with gross production for the past fiscal year for domestic and external sales of about 16.5 billion kilowatt hours. Of this figure 33.5% was supplied through nuclear generation. The financial risks involved with the production of the Point Lepreau nuclear generating station were discussed. Further, questions of assurances given for schedule and cost, licencing, and long-term plant performance of the Point Lepreau no. 2 unit were addressed. These were discussed with particular emphasis on the competition for the New England market

New Brunswick air quality monitoring results for the year 2000

International Nuclear Information System (INIS)

Hughes, R.

2002-01-01

Monitoring data for air quality in New Brunswick in 2000 is presented in this document. Designed for the general public, it summarizes the air quality results for 2000 and focuses on air quality assessment as it relates to existing air quality standards and objectives. The report also contains the long term trend data for representative sites. The New Brunswick Air Quality Regulation of the Clean Air Act specifies the air quality standards applicable for carbon monoxide, sulphur dioxide, hydrogen sulphide, nitrogen dioxide and total suspended particulate. The monitoring was conducted at 58 sites in 8 regional monitoring networks throughout the province. Thirteen additional sites were selected for the measurement of acid rain. Despite the fact that no standards were in effect in 2000 in New Brunswick for several substances, including inhalable particulate matter, ground-level ozone, volatile organic compounds, selected semi-volatile organic compounds, trace elements in particulate matter and mercury in air and precipitation, their levels were measured at some locations. The results indicate that emissions trends, variations in industrial output, changing process or emission control technologies, and weather conditions throughout the year explain most of the variations in results between regions. As a rule, compliance with standards is good. Acid rain continued to impact, especially in the southwestern districts of the province. Since 1996, the levels of mercury in precipitation has slightly declined. 39 refs., 43 figs

Brunswick-2 water chemistry. Interim report

International Nuclear Information System (INIS)

Miller, A.D.

1981-04-01

This study summarizes and interprets the nearly half million data points obtained through January of 1978 from the continuous monitoring equipment and data acquisition computers at Brunswick-2. Dissolved oxygen, specific conductance, and pH levels of 12 separate sample points were measured and correlated to plant operation, leading to a more complete understanding of the water chemistry of boiling water reactors. The measured parameters were characterized for various reactor power levels, startups, shutdowns, resin intrusions, etc

77 FR 11086 - Revised Notice of Intent To Prepare a Draft Environmental Impact Statement for the Brunswick...

Science.gov (United States)

2012-02-24

... evaluate the impacts of the proposed CSDR alternatives to reduce coastal storm damages from beach erosion... Prepare a Draft Environmental Impact Statement for the Brunswick County Beaches, NC, Coastal Storm Damage... conducting a General Reevaluation Report (GRR) for the Brunswick County Beaches, NC, Coastal Storm Damage...

Analysis of KROTOS KS-2 and KS-4 steam explosion experiments with TEXAS-VI

Energy Technology Data Exchange (ETDEWEB)

Chen, Ronghua, E-mail: rhchen@mail.xjtu.edu.cn [State Key Laboratory of Multiphase Flow in Power Engineering, School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049 (China); Wang, Jun [Nuclear Engineering and Engineering Physics, College of Engineering, University of Wisconsin Madison, WI 53706 (United States); Su, G.H.; Qiu, Suizheng [State Key Laboratory of Multiphase Flow in Power Engineering, School of Nuclear Science and Technology, Xi’an Jiaotong University, Xi’an 710049 (China); Corradini, M.L., E-mail: Corradini@engr.wisc.edu [Nuclear Engineering and Engineering Physics, College of Engineering, University of Wisconsin Madison, WI 53706 (United States)

2016-12-01

Highlights: • The KS-2 and KS-4 steam explosion experiments were analyzed by TEXAS-VI. • The coarse mixing status up to the explosion triggering time was well predicted by TEXAS-VI. • The predicted dynamic explosion pressure was in good agreement with the experimental results. - Abstract: TEXAS-VI is a transient, three-field, one-dimensional mechanistic model for the steam explosion phenomena. A fuel solidification model and associated fragmentation criteria of the solidifying particle for both the mixing phase and explosion phase were developed and incorporated into TEXAS-VI to account for solidification. In the present study, TEXAS-VI was used to analyze the KS-2 and KS-4 steam explosion experiments, which were performed in the KROTOS facility as part of the OECD-SERENA-2 program. In the simulation, the KROTOS experimental facility was modeled as Eulerian control volumes based on the facility geometry. The molten corium jet was divided up into a series of LaGrangian master particles equal to the initial jet diameter. Both the mixing phase and the explosion phase of the experiments were simulated by TEXAS-VI. Comparison to test data indicates that the fuel jet kinematics and the vapor volume during the mixing phase were well predicted by TEXAS-VI. The TEXAS-VI prediction of the dynamic explosion pressure at different axial locations in the test was also in good agreement with the experimental results. The maximum pressure of KS-2 and KS-4 predicted by TEXAS-VI were 16.7 MPa and 41.9 MPa, respectively. The KS-4 maximum steam explosion pressure predicted by TEXAS-VI was higher than that of KS-2, which was consistent with experiment observation. The observed differences of the dynamic explosion pressure between the KS-2 and KS-4 experiments were also successfully simulated by TEXAS-VI. This suggests that TEXAS-VI is able to analyze the effect of prototypic melt compositions on the steam explosion phenomena. Additional benchmarking and evaluations are ongoing.

Experience on detection of leakages in LMFBR-steam generators

International Nuclear Information System (INIS)

Smit, C.C.

1975-01-01

One of the advantages of long time on full size LMFBR-components is that experience is gained nut only or, the behaviour of components at normal conditions, but also on the operational consequences (real or imaginary) disturbances. One of the most difficult situations that do occur during steam generator operation is the sudden appearance of a leak indication on the hydrogen detectors. It is possible to connect an automatic trip action to the hydrogen detector however, there are reasons not to do so. Spurious signals, which unfortunately do occur rather frequently, can cause unnecessary shut downs. In the case of a very small leak it can be very difficult to locate the leaking steam generator module and to get an impression of the size of the leak. The time available to confirm the leak, locate the component and to take the proper measures is strongly dependent on the leaking rate or translated into a visual signal, on the rate of rise of the hydrogen level shown on the instrument. During the operation of the 50 MW-SCTF at Hengelo experience was obtained with leak indications caused by real and imaginary leaks

Experience on detection of leakages in LMFBR-steam generators

Energy Technology Data Exchange (ETDEWEB)

Smit, C C

1975-07-01

One of the advantages of long time on full size LMFBR-components is that experience is gained nut only or, the behaviour of components at normal conditions, but also on the operational consequences (real or imaginary) disturbances. One of the most difficult situations that do occur during steam generator operation is the sudden appearance of a leak indication on the hydrogen detectors. It is possible to connect an automatic trip action to the hydrogen detector however, there are reasons not to do so. Spurious signals, which unfortunately do occur rather frequently, can cause unnecessary shut downs. In the case of a very small leak it can be very difficult to locate the leaking steam generator module and to get an impression of the size of the leak. The time available to confirm the leak, locate the component and to take the proper measures is strongly dependent on the leaking rate or translated into a visual signal, on the rate of rise of the hydrogen level shown on the instrument. During the operation of the 50 MW-SCTF at Hengelo experience was obtained with leak indications caused by real and imaginary leaks.

A three-dimensional laboratory steam injection model allowing in situ saturation measurements. [Comparing steam injection and steam foam injection with nitrogen and without nitrogen

Energy Technology Data Exchange (ETDEWEB)

Demiral, B.M.R.; Pettit, P.A.; Castanier, L.M.; Brigham, W.E.

1992-08-01

The CT imaging technique together with temperature and pressure measurements were used to follow the steam propagation during steam and steam foam injection experiments in a three dimensional laboratory steam injection model. The advantages and disadvantages of different geometries were examined to find out which could best represent radial and gravity override flows and also fit the dimensions of the scanning field of the CT scanner. During experiments, steam was injected continuously at a constant rate into the water saturated model and CT scans were taken at six different cross sections of the model. Pressure and temperature data were collected with time at three different levels in the model. During steam injection experiments, the saturations obtained by CT matched well with the temperature data. That is, the steam override as observed by temperature data was also clearly seen on the CT pictures. During the runs where foam was present, the saturation distributions obtained from CT pictures showed a piston like displacement. However, the temperature distributions were different depending on the type of steam foam process used. The results clearly show that the pressure/temperature data alone are not sufficient to study steam foam in the presence of non-condensible gas.

Lyme disease risk in dogs in New Brunswick

OpenAIRE

Bjurman, Natalie K.; Bradet, Gina; Lloyd, Vett K.

2016-01-01

This study assessed the seroprevalence of anti-Borrelia burgdorferi antibodies in New Brunswick dogs. Testing of 699 serum samples from dogs across the province revealed a 6% province-wide seropositivity, more than 6 times higher than that found in 2008. The rapid increase in seropositivity indicates increased Lyme disease risk to both canine and human health.

Du Nouveau dans l'enseignement professional au Nouveau Brunswick

Science.gov (United States)

Le Blanc, L. F.

1974-01-01

In French, the author traces the development of vocational education in New Brunswick; despite a general decrease in school enrollment, adult vocational training has increased. A "modular" carpentry course illustrates an individualized approach minimizing the problem caused by unilingual instruction of students, some of whom speak…

Compilation of Water-Resources Data and Hydrogeologic Setting for Brunswick County, North Carolina, 1933-2000

Science.gov (United States)

Fine, Jason M.; Cunningham, William L.

2001-01-01

Water-resources data were compiled for Brunswick County, North Carolina, to describe the hydrologic conditions of the County. Hydrologic data collected by the U.S. Geological Survey as well as data collected by other governmental agencies and reviewed by the U.S. Geological Survey are presented. Data from four weather stations and two surface-water stations are summarized. Data also are presented for land use and land cover, soils, geology, hydrogeology, 12 continuously monitored ground-water wells, 73 periodically measured ground-water wells, and water-quality measurements from 39 ground-water wells. Mean monthly precipitation at the Longwood, Shallotte, Southport, and Wilmington Airport weather stations ranged from 2.19 to 7.94 inches for the periods of record, and mean monthly temperatures at the Longwood, Southport, and Wilmington Airport weather stations ranged from 43.4 to 80.1 degrees Fahrenheit for the periods of record. An evaluation of land-use and land-cover data for Brunswick County indicated that most of the County is either forested land (about 57 percent) or wetlands (about 29 percent). Cross sections are presented to illustrate the general hydrogeology beneath Brunswick County. Water-level data for Brunswick County indicate that water levels ranged from about 110 feet above mean sea level to about 22 feet below mean sea level. Chloride concentrations measured in aquifers in Brunswick County ranged from near 0 to 15,000 milligrams per liter. Chloride levels in the Black Creek and Cape Fear aquifers were measured at well above the potable limit for ground water of 250 milligrams per liter set by the U.S. Environmental Protection Agency for safe drinking water.

Deliberate ignition of hydrogen-air-steam mixtures in condensing steam environments

International Nuclear Information System (INIS)

Blanchat, T.K.; Stamps, D.W.

1997-05-01

Large scale experiments were performed to determine the effectiveness of thermal glow plug igniters to burn hydrogen in a condensing steam environment due to the presence of water sprays. The experiments were designed to determine if a detonation or accelerated flame could occur in a hydrogen-air-steam mixture which was initially nonflammable due to steam dilution but was rendered flammable by rapid steam condensation due to water sprays. Eleven Hydrogen Igniter Tests were conducted in the test vessel. The vessel was instrumented with pressure transducers, thermocouple rakes, gas grab sample bottles, hydrogen microsensors, and cameras. The vessel contained two prototypic engineered systems: (1) a deliberate hydrogen ignition system and (2) a water spray system. Experiments were conducted under conditions scaled to be nearly prototypic of those expected in Advanced Light Water Reactors (such as the Combustion Engineering (CE) System 80+), with prototypic spray drop diameter, spray mass flux, steam condensation rates, hydrogen injection flow rates, and using the actual proposed plant igniters. The lack of any significant pressure increase during the majority of the burn and condensation events signified that localized, benign hydrogen deflagration(s) occurred with no significant pressure load on the containment vessel. Igniter location did not appear to be a factor in the open geometry. Initially stratified tests with a stoichiometric mixture in the top showed that the water spray effectively mixes the initially stratified atmosphere prior to the deflagration event. All tests demonstrated that thermal glow plugs ignite hydrogen-air-steam mixtures under conditions with water sprays near the flammability limits previously determined for hydrogen-air-steam mixtures under quiescent conditions. This report describes these experiments, gives experimental results, and provides interpretation of the results. 12 refs., 127 figs., 16 tabs

Simulation of first SERENA KROTOS steam explosion experiment

International Nuclear Information System (INIS)

Leskovar, Matjaz; Ursic, Mitja

2009-01-01

A steam explosion may occur when, during a severe reactor accident, the molten core comes into contact with the coolant water. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment. To resolve the open issues in steam explosion understanding and modeling, the OECD program SERENA Phase 2 was launched at the end of year 2007, focusing on nuclear applications. SERENA comprises an experimental program, which is being carried out in the complementary KROTOS and TROI corium facilities, accompanied by a comprehensive analytical program, where also pre- and post-test calculations are foreseen. In the paper the sensitivity post-test calculations of the first SERENA KROTOS experiment KS-1, which were performed with the code MC3D, are presented and discussed. Since the results of the SERENA tests are restricted to SERENA members, only the various calculation results are given, not comparing them to experimental measurements. Various premixing and explosion simulations were performed on a coarse and a fine numerical mesh, applying two different jet breakup models (global, local) and varying the minimum bubble diameter in the explosion simulations (0.5 mm, 5 mm). The simulations revealed that all varied parameters have a significant influence on the calculation results, as was expected since the fuel coolant interaction process is a highly complex phenomenon. The results of the various calculations are presented in comparison and the observed differences are discussed and explained. (author)

High-temperature hydrogen-air-steam detonation experiments in the BNL small-scale development apparatus

International Nuclear Information System (INIS)

Ciccarelli, G.; Ginsberg, T.; Boccio, J.; Economos, C.; Finfrock, C.; Gerlach, L.; Sato, K.

1994-01-01

The Small-Scale Development Apparatus (SSDA) was constructed to provide a preliminary set of experimental data to characterize the effect of temperature on the ability of hydrogen-air-steam-mixtures to undergo detonations and, equally important, to support design of the larger-scale High-Temperature Combustion Facility (HTCF) by providing a test bed for solution of a number of high-temperature design and operational problems. The SSDA, the central element of which is 10-cm inside diameter, 6.1-m long tubular test vessel designed to permit detonation experiments at temperatures up to 700K, was employed to study self-sustained detonations in gaseous mixtures of hydrogen, air, and steam at temperature between 300K and 650K at a fixed pressure of 0.1 MPa. Detonation cell size measurements provide clear evidence that the effect of hydrogen-air gas mixture temperature, in the range 300K to 650K, is to decrease cell size and, hence, to increase the sensitivity of the mixture to undergo detonations. The effect of steam content, at any given temperature, is to increase the cell size and, thereby, to decrease the sensitivity of stoichiometric hydrogen-air mixtures. The one-dimensional ZND model does a very good job at predicting the overall trends in the cell size data over the range of hydrogen-air-steam mixture compositions and temperature studied in the experiments. Experiments were conducted to measure the rate of hydrogen oxidation in the absence of ignition sources at temperatures of 500K and 650K, for hydrogen-air mixtures of 15% and 50%, and for a mixture of equimolar hydrogen-air and 30% steam at 650K. The rate of hydrogen oxidation was found to be significant at 650K. Reduction of hydrogen concentration by chemical reaction from 50 to 44% hydrogen, and from 15 to 11% hydrogen, were observed on a time frame of minutes. The DeSoete rate equation predicts the 50% experiment very well, but greatly underestimates the reaction rate of the lean mixtures

Particles growth by steam nucleation in a containment. The PITEAS experiment

International Nuclear Information System (INIS)

Layly, V.D.

1993-01-01

One of the major issues of the fission products inventory in the containment of a nuclear power plant during the few hours following the initiating phase of a severe accident (involving core degradation and fission products release from the fuel) is the physical behaviour of the aerosols suspended in the containment atmosphere. The aerosol mass concentration versus time is controlled by agglomeration, sedimentation, deposition on walls and steam nucleation phenomena. In order to assess the Nuclear Safety codes dealing with such phenomena, analytical experiments are necessary. PITEAS is an analytical experiment, on the behaviour of soluble aerosols in humid atmosphere. The PITEAS program covers three topics: diffusiophoresis, agglomeration and particle growth by steam nucleation. In the present work, we analyse the results of the tests devoted to the soluble particle growth and how such results are used to assess the IPSN code AEROSOL-B2 code, part of the System of codes ESCADRE. (author)

New Brunswick Power Corporation annual report, 1993/94

International Nuclear Information System (INIS)

1994-01-01

Net income for New Brunswick Power in the year 1993/94 was $23.7 million, up from $1.1 million the previous year. Total revenues decreased by $7.4 million as decreases in export sales were greater than in-province load growth and due to the impacts of a 5% rate increase in 1992 and 2.9% increase in January 1994. In-province energy sales revenues totalled $732.8 million, an increase of $25.1 million from 1992/93. Residential energy sales increased by 2.1% while the those to industrial customers declined by 1.4%, due to continuing economic difficulties in the pulp and paper and mining sector. Out of province revenues decreased by $33.3 million or 19.7%, with total energy exports decreasing by 688.6 GWh to 2453.6 GWh. A significant highlight of the year was integration of the Belledune generating station onto the New Brunswick system in October. A state-of-the-art coal fired facility, Belledune will contribute 450 MW to the system. Its flue gas scrubber will reduce sulfur dioxide emissions by 90%. 6 figs., 9 tabs

New Brunswick Power Corporation annual report, 1993/94

International Nuclear Information System (INIS)

1994-01-01

Net income for New Brunswick Power in the year 1993/94 was $23.7 million, up from $1.1 million the previous year. Total revenues decreased by $7.4 million as decreases in export sales were greater than in-province load growth and due to the impacts of a 5% rate increase in 1992 and 2.9% increase in January 1994. In-province energy sales revenues totalled $732.8 million, an increase of $25.1 million from 1992/93. Residential energy sales increased by 2.1% while the those to industrial customers declined by 1.4%, due to continuing economic difficulties in the pulp and paper and mining sector. Out of province revenues decreased by $33.3 million or 19.7%, with total energy exports decreasing by 688.6 GWh to 2453.6 GWh. A significant highlight of the year was integration of the Belledune generating station onto the New Brunswick sytem in October. A state-of-the-art coal fired facility, Belledune will contribute 450 MW to the system. Its flue gas scrubber will reduce sulfur dioxide emissions by 90%. 6 figs., 9 tabs

Influence of steam generator tube ruptures during semiscale loss-of-coolant experiments

International Nuclear Information System (INIS)

Cozzuol, J.M.; Larson, T.K.

1978-01-01

Examination of the U-tubes in the steam generators of some large commercial pressurized water reactors (PWR) has revealed the existence of leakage and in some cases structural weakening of the tubes. This structural weakening enhances the possibility of tubes rupturing during a hypothesized loss-of-coolant accident (LOCA). Considerable interest has been shown in the analysis of tube ruptures concurrent with a hypothesized LOCA since the presence of tube ruptures has the potential to influence the system thermal-hydraulic response and could foreseeably result in a more severe core thermal behavior than might otherwise occur. To experimentally investigate the influence of steam generator tube ruptures on the thermal-hydraulic response of PWR type system, a series of experiments was conducted in the Semiscale Mod-1 system by EG and G Idaho, Inc., for the U.S. Nuclear Regulatory Commission and the Department of Energy. The primary objective of the experiments was to obtain data which could be used to evaluate the influence of the simulated tube ruptures on the system and core thermal-hydraulic response for a range of tube ruptures that was expected to provide the potential for high cladding temperatures in the Semiscale facility. The experiments were conducted assuming a variety in the number of tubes ruptured during large break loss-of-coolant conditions. The number of experiments conducted permitted determination of the range of tube ruptures for which high peak cladding temperatures could result in the Semiscale Mod-1 system. The paper contains a description of the Semiscale Mod-1 system and a discussion of the steam generator tube rupture tests conducted. The experimental results from the test series and the thermal-hydraulic phenomena found to influence the core thermal response during the experiments are discussed

High-temperature hydrogen-air-steam detonation experiments in the BNL small-scale development apparatus

International Nuclear Information System (INIS)

Ciccarelli, G.; Ginsburg, T.; Boccio, J.; Economos, C.; Finfrock, C.; Gerlach, L.; Sato, K.; Kinoshita, M.

1994-08-01

The Small-Scale Development Apparatus (SSDA) was constructed to provide a preliminary set of experimental data to characterize the effect of temperature on the ability of hydrogen-air-steam mixtures to undergo detonations and, equally important, to support design of the larger scale High-Temperature Combustion Facility (HTCF) by providing a test bed for solution of a number of high-temperature design and operational problems. The SSDA, the central element of which is a 10-cm inside diameter, 6.1-m long tubular test vessel designed to permit detonation experiments at temperatures up to 700K, was employed to study self-sustained detonations in gaseous mixtures of hydrogen, air, and steam at temperatures between 300K and 650K at a fixed initial pressure of 0.1 MPa. Hydrogen-air mixtures with hydrogen composition from 9 to 60 percent by volume and steam fractions up to 35 percent by volume were studied for stoichiometric hydrogen-air-steam mixtures. Detonation cell size measurements provide clear evidence that the effect of hydrogen-air gas mixture temperature, in the range 300K-650K, is to decrease cell size and, hence, to increase the sensitivity of the mixture to undergo detonations. The effect of steam content, at any given temperature, is to increase the cell size and, thereby, to decrease the sensitivity of stoichiometric hydrogen-air mixtures. The hydrogen-air detonability limits for the 10-cm inside diameter SSDA test vessel, based upon the onset of single-head spin, decreased from 15 percent hydrogen at 300K down to between 9 and 10 percent hydrogen at 650K. The one-dimensional ZND model does a very good job at predicting the overall trends in the cell size data over the range of hydrogen-air-steam mixture compositions and temperature studied in the experiments

New Brunswick Market Design Committee : final report

International Nuclear Information System (INIS)

2002-04-01

This report presents a plan for implementing New Brunswick's electricity restructuring. It includes two resolutions and 95 recommendations to help achieve the main policy objectives of the White Paper, the New Brunswick Energy Policy. The most significant policy goal outlined in the White Paper is the restructuring of the electricity sector, with initial competition being only at the wholesale and large industrial retail level. The Board of Commissioners of Public Utilities will regulate many aspects of the new electricity market. In addition, green pricing options will be made available. The Market Design Committee recommends that the government set up a bilateral contract market for wholesale and large industrial customers to contract with alternate providers for electrical power. Power generators would have the freedom to sell by contract to customers both within and outside the province. The report describes the requirements for establishing a bilateral contract market and how it functions. The Committee also recommends designating a Heritage Pool of electricity available from the existing generation assets in the province. Other recommendations include the creation of programs that will help meet environmental protection goals. The programs include net metering, support of embedded generation, renewable portfolio standards, energy efficiency programs, green pricing, broad-based carbon dioxide emissions trading, emission performance standards, and the promotion of cogeneration. 37 refs., 2 figs

Diabetes Case Management in Primary Care: The New Brunswick Experience and Expanding the Practice of the Certified Diabetes Educator Nurse into Primary Care.

Science.gov (United States)

Jones, Shelley L

2015-08-01

The role of the outreach diabetes case manager in New Brunswick, Canada, was first developed in the Moncton Area of Horizon Health Network in response to a physician-identified gap between patients' diagnoses of diabetes and their attendance at the local diabetes education centre. This model of collaborative interprofessional practice increases support for primary care providers and people living with diabetes in that they are being provided the services of certified diabetes educators who can address knowledge gaps with respect to evidence-based guidelines and best practice, promote advancement of diabetes and chronic-disease management therapies and support adherence to treatment plans and self-management practices. This report chronicles a review of the implementation, expansion and evaluation of the outreach diabetes case manager model in the province of New Brunswick, Canada, along with the rationale for development of the role for registered nurses in other jurisdictions. Crown Copyright © 2015. Published by Elsevier Inc. All rights reserved.

An overview of the applications for early warning and mapping of the flood events in New Brunswick

DEFF Research Database (Denmark)

Mioc, Darka; McGillivray, E.; Anton, F.

2014-01-01

This paper gives an overview of the on-line flood warning implementation in the province of New Brunswick, Canada. The on-line flood warning applications are available via the “River Watch” website provided by the New Brunswick Department of Environment. Advanced GIS technology combined with hydr....... The searchable historical database containing reports about the impact of past floods and estimated damages provides a valuable insight into the past of the province of New Brunswick and the motivation for development of the system for flood prediction and management....... with hydrological modelling, provide a mapping and visualization tool that can be used by emergency managers and the general public to predict possible flood zones. The applications developed for “River Watch” support the processing of large amounts of digital terrain and hydrological data, which are then...

Wood supply : what bioenergy resources are available as harvest residue and non-merchantable wood in New Brunswick?

International Nuclear Information System (INIS)

Forgrave, K.

2005-01-01

This paper examines resources available for bioenergy production in New Brunswick. An estimate of New Brunswick harvest levels, based on New Brunswick Forest Products Association web site figures was presented. Private wood lots harvest levels were also provided, with all estimates based on volume per year. Market conditions were discussed, with an emphasis on the dictates of price and availability for private wood lot owners. Market conditions and unused volume statistics were discussed. Various forest management options include the use of veneer, saw log, stud wood and pulp wood. Details of bioenergy percentages and potential were presented and general silviculture issues were discussed. It was concluded that many of the trade-offs of bioenergy involve placing more demands on forests, and that purchase prices will increasingly dictate volumes obtained from private wood lots. tabs, figs

New Coleoptera records from New Brunswick, Canada: Cerambycidae

OpenAIRE

Webster,Reginald; Sweeney,Jon; DeMerchant,Ian; Silk,Peter; Mayo,Peter

2012-01-01

Abstract Five species of Cerambycidae, Acmaeops discoideus (Haldeman), Anelaphus villosus (Fabricius), Phymatodes species (CNC sp. n. #1), Sarosesthes fulminans (Fabricius), and Urgleptus signatus (LeConte) are newly recorded for New Brunswick, Canada. All but Acmaeops villosus are new to the Maritime provinces. Phymatodes testaceus (Linnaeus) is removed from the faunal list of the province as a result of mislabeled specimens, records of Phymatodes maculicollis LeConte are presented confirmin...

Flooding experiments with steam and water in a large diameter vertical tube

International Nuclear Information System (INIS)

Williams, S.N.; Solom, M.; Draznin, O.; Choutapalli, I.; Vierow, K.

2009-01-01

An experimental study on flooding in a large diameter tube is being conducted. In a countercurrent, two-phase flow system, flooding can be defined as the onset of flow reversal of the liquid component which results in cocurrent flow. Flooding can be perceived as a limit to two-phase countercurrent flow, meaning that pairs of liquid and gas flow rates exist that define the envelope for stable countercurrent flow for a given system. Flooding in the AP600 pressurizer surge line can affect the vessel refill rate following a small break LOCA. Analysis of hypothetical severe accidents with current simplified flooding models show that these models represent the largest uncertainty in steam generator tube creep rupture. During a hypothetical station blackout scenario without auxiliary feedwater recovery, should the hot leg become voided, the pressurizer liquid will drain to the hot leg and flooding may occur in the surge line. Experiments have been conducted in a 3-inch (76.2 mm) diameter tube with subcooled water and superheated steam as the working fluids at atmospheric pressure. Water flows down the inside of the tube as an annulus while the steam flows upward in the middle. Water flow rates vary from 3.5 to 12 GPM (0.00022 to 0.00076 m 3 /s) and the water inlet temperature is about 70degC. The steam inlet temperature is about 110degC. It was found that a larger steam flow rate was needed to achieve flooding for a lower water flow rate and for a higher water flow rate. This unique data for flooding in steam-water systems in large diameter tubes will reduce uncertainty in flooding models currently utilized in reactor safety codes. (author)

Circumferential cracking of steam generator tubes

International Nuclear Information System (INIS)

Karwoski, K.J.

1997-04-01

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

Steam generator 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

ALPHA visual data collection. STX005-025: melt drop steam explosion experiments

International Nuclear Information System (INIS)

Moriyama, Kiyofumi; Yamano, Norihiro; Maruyama, Yu; Kudo, Tamotsu; Sugimoto, Jun

1999-03-01

Steam explosion is a phenomenon in which a high temperature liquid gives its internal energy to a low temperature volatile liquid extremely quickly causing rapid evaporation and shock wave generation. In the field of nuclear reactor safety research regarding severe accidents in LWRs, steam explosions involving molten fuel and coolant has been recognized as a potential threat to the integrity of the reactor containment vessel. In the ALPHA (Assessment of Loads and Performance of Containment in Hypothetical Accident) program, experiments were performed to investigate the phenomenology of vapor explosions using iron-alumina thermite melt as a simulant of molten core. This report collects the experimental results especially emphasizing the visual observations by high speed photography. (author)

Carrots and Sticks: New Brunswick and Maine Forest Landowner Perceptions Toward Incentives and Regulations

Science.gov (United States)

Quartuch, Michael R.; Beckley, Thomas M.

2014-01-01

The governments of countries that allow private land ownership have two main tools to motivate landowner behavior: regulations and incentives. This research examines landowner preferences toward these policy tools and asks specifically: Do private forest landowners in New Brunswick and Maine believe that regulations and/or incentives are effective means to motivate responsible stewardship? Can landowners identify explicit regulations and policies that restrict property rights? Also, we were interested to see if any discernible differences existed between these adjacent jurisdictions from different countries, but that share similar forests and a similar settlement history. We identified and interviewed diverse landowners, recorded and transcribed our discussions, and analyzed the results using a grounded theory approach. Findings suggest that both New Brunswick and Maine participants are fairly comfortable with most regulations and many agreed that a combination of incentives and regulations are in fact useful. Furthermore, landowners in New Brunswick discussed non-monetary incentives as a mechanism to reward "good" stewardship; while Maine respondents articulated a degree of responsible stewardship that transcends a need to incentivize landowners. This study demonstrates that diverse landowners may be more comfortable with environmental regulations than previously understood and may be interested in non-monetary incentives.

Steam-water relative permeability

Energy Technology Data Exchange (ETDEWEB)

Ambusso, W.; Satik, C.; Home, R.N. [Stanford Univ., CA (United States)

1997-12-31

A set of relative permeability relations for simultaneous flow of steam and water in porous media have been measured in steady state experiments conducted under the conditions that eliminate most errors associated with saturation and pressure measurements. These relations show that the relative permeabilities for steam-water flow in porous media vary approximately linearly with saturation. This departure from the nitrogen/water behavior indicates that there are fundamental differences between steam/water and nitrogen/water flows. The saturations in these experiments were measured by using a high resolution X-ray computer tomography (CT) scanner. In addition the pressure gradients were obtained from the measurements of liquid phase pressure over the portions with flat saturation profiles. These two aspects constitute a major improvement in the experimental method compared to those used in the past. Comparison of the saturation profiles measured by the X-ray CT scanner during the experiments shows a good agreement with those predicted by numerical simulations. To obtain results that are applicable to general flow of steam and water in porous media similar experiments will be conducted at higher temperature and with porous rocks of different wetting characteristics and porosity distribution.

New Brunswick Laboratory progress report, October 1994--September 1995

International Nuclear Information System (INIS)

1996-03-01

The mission of the New Brunswick Laboratory (NBL) of the A. S. Department of Energy (DOE) is to serve as the National Certifying Authority for nuclear reference materials and to provide an independent Federal technical staff and laboratory resource performing nuclear material measurement, safeguards, and non-proliferation functions in support of multiple program sponsors. This annual report describes accomplishments achieved in carrying out NBL's assigned missions

Groundwater flow in the Brunswick/Glynn County area, Georgia, 2000-04

Science.gov (United States)

Cherry, Gregory S.

2015-01-01

An existing regional steady-state model for coastal Georgia, and parts of South Carolina and Florida, was revised to evaluate the local effects of pumping on the migration of high chloride (saline) water in the Upper Floridan aquifer located in the Brunswick/Glynn County, Georgia (Ga.) area. Revisions were focused on enhancing the horizontal and vertical resolution of the regional model grid in the vicinity of saline water. Modifications to the regional model consisted of (1) limiting grid size to a maximum of 500 feet (ft) per side in the vicinity of chloride contamination; (2) representing the upper and lower Brunswick aquifers with distinct model layers; (3) similarly, representing upper and lower water-bearing zones of the Upper Floridan aquifer with distinct model layers in Glynn and Camden Counties, Ga.; and (4) establishing new hydraulic-property zones in the Upper Floridan aquifer. The revised model simulated steady-state conditions that were assumed to exist during 2000 and 2004.

MELCOR simulation of steam condensation effect on hydrogen behavior in THAI HM-2 experiment

Energy Technology Data Exchange (ETDEWEB)

Lee, Seongnyeon; Lee, Jung-Jae; Cho, Yong-Jin [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Kim, Sung Joong [Hanyang University, Seoul (Korea, Republic of)

2015-10-15

In this study, MELCOR simulation was carried out for THAI HM-2 experiment of OECD. As a results, stratification of hydrogen cloud was reasonably captured in MELCOR simulation. Furthermore, the pressure from simulation results in cases where mass transfer coefficient of MELCOR condensation model was modified was good agreement with the experimental results. Containment Filtered Ventilation System (CFVS) has been introduced as facility to prevent containment failure during severe accident. However, possibility of hydrogen risk has been issued due to inflow of hydrogen, condensation and removal of steam and complicated inner structure in CFVS. Preferentially benchmark work for THAI HM-2 experiment of OECD was decided to validate the methodology before detailed assessment of hydrogen risk in CFVS. The objectives of THAI HM-2 experiment were evaluation of hydrogen behavior, verification of numerical analysis tools and so on. In this paper, therefore, MELCOR simulation was carried out in comparison with the experiment results. Additionally, steam condensation effect was considered for detailed simulation. Hydrogen concentration from MELCOR results was underestimated in comparison to the experimental results.

New Brunswick Market Design Committee : market design issues paper : choice of market model

International Nuclear Information System (INIS)

2001-01-01

A model for the competitive electricity market in New Brunswick was not specified by the White Paper: New Brunswick Energy Policy, published in March 2001. One of the tasks of the Market Design Committee (MDC) is to select a market model for the province. This report was prepared with this objective in mind. It begins by providing a description of the basic functions that must be performed by any electricity system. Different market models will function differently, and a descriptions of how the functions would be performed under each model is presented. Considering the specific size and geographic location of New Brunswick, a number of actual markets that could be of interest are presented. The various electricity markets normally use one of four market models: vertically integrated monopoly utility, a competitive pool market, a bilateral contract market, and a single-buyer market. The first model was not explained as it represents the existing, non-competitive model whereby the government owns and/or regulates price and investment decision. The case where a market operator receives bids and offers from buyers and sellers and matches them to derive a price and schedule for the spot market is what is called the competitive pool market. When electricity trade takes place through a series of contracts between individual buyers and sellers, it is referred to as a bilateral contract market. Finally, the single-buyer market is defined as a monopoly where the buyer purchases from multiple sellers based on competition amongst them. Different examples are provided of applications of the three markets described in the paper. Both New England and New Zealand were chosen to better illustrate the concepts of a fully competitive pool-based market, as they either have close ties to New Brunswick, or share physical similarities. The single-buyer model is illustrated by the case of Northern Ireland where size is similar. The choices made in Quebec were described in the final

Energy efficiency potential study for New Brunswick

International Nuclear Information System (INIS)

1992-05-01

The economic and environmental impacts associated with economically attractive energy savings identified in each of four sectors in New Brunswick are analyzed. The results are derived through a comparison of two potential future scenarios. The frozen efficiency scenario projects what future energy expenditures would be if no new energy efficiency initiatives are introduced. The economic potential scenario projects what those expenditures would be if all economically attractive energy efficiency improvements were gradually implemented over the next 20 years. Energy related emissions are estimated under scenarios with and without fuel switching. The results show, for example, that New Brunswick's energy related CO 2 emissions would be reduced by ca 5 million tonnes in the year 2000 under the economic potential scenario. If fuel switching is adopted, an additional 1 million tonnes of CO 2 emissions could be saved in the year 2000 and 1.6 million tonnes in 2010. The economic impact analysis is restricted to efficiency options only and does not consider fuel switching. Results show the effect of the economic potential scenario on employment, government revenues, and intra-industry distribution of employment gains and losses. The employment impact is estimated as the equivalent of the creation of 2,424 jobs annually over 1991-2010. Government revenues would increase by ca $24 million annually. The industries benefitting most from energy efficiency improvements would be those related to construction, retail trade, finance, real estate, and food/beverages. Industries adversely affected would be the electric power, oil, and coal sectors. 2 figs., 37 tabs

Steam line rupture experiments with the PPOOLEX test facility

International Nuclear Information System (INIS)

Laine, J.; Puustinen, M.

2008-07-01

The results of the steam line rupture experiment series in 2007 with the scaled down PPOOLEX test facility designed and constructed at Lappeenranta University of Technology are reported. The test facility is a closed stainless steel vessel divided into two compartments, dry well and wet well. Air was blown into the dry well compartment and from there through a DN200 blowdown pipe to the condensation pool. Altogether five experiments, each consisting of several blows (tests), were carried out. The main purpose of the experiment series was to study the initial phase of a postulated steam line break accident inside a BWR containment. Specifically, thermal stratification in the dry well compartment and ejection of water plug from the blowdown pipe were of interest. In addition, the effect of counterpressure on bubble dynamics was studied. A temperature difference of approximately 15 deg. C between the upper and lower part of the dry well was measured. In the wet well gas space, a temperature difference of more than 30 deg. C was registered. These were measured during the compression period of the tests. Towards the end of the tests the temperature differences tended to disappear. To get a more detailed picture of temperature distribution in the wet well, especially close to the water level, a dense net of measurements is required in future experiments. In longer experiments, heat conduction to structures and heat losses to surroundings should also be taken into account. Ejection of water plugs from the blowdown pipe did not cause notable loads to the structures due to the suppressing effect of the dry well compartment. The maximum measured pressure pulse at the pool bottom was only 10 kPa and the maximum strain amplitude at the pool bottom rounding was negligible both in axial and circumferential direction. As the counterpressure of the system increased, but the flow rate remained the same, the maximum size of the air bubbles at the blowdown pipe outlet got smaller and

Steam line rupture experiments with the PPOOLEX test facility

Energy Technology Data Exchange (ETDEWEB)

Laine, J.; Puustinen, M. (Lappeenranta Univ. of Technology, Nuclear Safety Research Unit (Finland))

2008-07-15

The results of the steam line rupture experiment series in 2007 with the scaled down PPOOLEX test facility designed and constructed at Lappeenranta University of Technology are reported. The test facility is a closed stainless steel vessel divided into two compartments, dry well and wet well. Air was blown into the dry well compartment and from there through a DN200 blowdown pipe to the condensation pool. Altogether five experiments, each consisting of several blows (tests), were carried out. The main purpose of the experiment series was to study the initial phase of a postulated steam line break accident inside a BWR containment. Specifically, thermal stratification in the dry well compartment and ejection of water plug from the blowdown pipe were of interest. In addition, the effect of counterpressure on bubble dynamics was studied. A temperature difference of approximately 15 deg. C between the upper and lower part of the dry well was measured. In the wet well gas space, a temperature difference of more than 30 deg. C was registered. These were measured during the compression period of the tests. Towards the end of the tests the temperature differences tended to disappear. To get a more detailed picture of temperature distribution in the wet well, especially close to the water level, a dense net of measurements is required in future experiments. In longer experiments, heat conduction to structures and heat losses to surroundings should also be taken into account. Ejection of water plugs from the blowdown pipe did not cause notable loads to the structures due to the suppressing effect of the dry well compartment. The maximum measured pressure pulse at the pool bottom was only 10 kPa and the maximum strain amplitude at the pool bottom rounding was negligible both in axial and circumferential direction. As the counterpressure of the system increased, but the flow rate remained the same, the maximum size of the air bubbles at the blowdown pipe outlet got smaller and

First Report of the Select Committee on Energy: Securing natural gas for New Brunswick

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-11-01

This document reports on the work of the Select Committee on Energy of the Legislative Assembly of New Brunswick, created to examine the proper and expeditious development of natural gas in New Brunswick, in such a way as to make natural gas available in the province to the widest possible extent, and to the maximum benefit of its people. The document outlines the key issues involved, namely the policy on laterals and pipeline access (a fundamental issue), cost of service and incentive regulation, bundled and unbundled services, load balancing, incentives for distribution system expansion, franchises, municipal involvement in gas development, the impact of natural gas industry development on restructuring the electric industry, environmental issues and required changes to legislation. Recommendations with respect to each issue are included.

Chemical operational experience with the water/steam-circuit at KNK II; Presentation at the meeting on Experience exchange on operational experience of fast breeder reactors, Karlsruhe/Bensberg/Kalkar, June 18. - 22. 1990

International Nuclear Information System (INIS)

Grumer, U.

1990-06-01

The availability of sodium cooled reactors depends essentially from the safety and reliability of the sodium heated steam generator. The transition from experimental plants with 12-20 MW electrical power to larger plants with 600 MW (BN-600) or 1200 MW (Superphenix) required the change from modular components to larger and compact steam generators with up to 800 MW. Defects of these large components cause extreme losses in availability of the plant and have to be avoided. In view of this request, a comprehensive test program has been performed at KNK II in addition to the normal control of the water/steam-circuit to compile all operational data on the water and steam side of the sodium heated steam generator. This paper describes the plant and the water/steam-circuit with its mode of operation. The experience with the surveillance and different methods of the conditioning are discussed in detail in this presentation

76 FR 6830 - Biweekly Notice; Applications and Amendments to Facility Operating Licenses Involving No...

Science.gov (United States)

2011-02-08

...- and 6- day supply is based on the diesel generator manufacturer's consumption values for the run time... different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in... Company, Docket Nos. 50-325 and 50-324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County...

Simulation benchmark based on THAI-experiment on dissolution of a steam stratification by natural convection

Energy Technology Data Exchange (ETDEWEB)

Freitag, M., E-mail: freitag@becker-technologies.com; Schmidt, E.; Gupta, S.; Poss, G.

2016-04-01

Highlights: . • We studied the generation and dissolution of steam stratification in natural convection. • We performed a computer code benchmark including blind and open phases. • The dissolution of stratification predicted only qualitatively by LP and CFD models during the blind simulation phase. - Abstract: Locally enriched hydrogen as in stratification may contribute to early containment failure in the course of severe nuclear reactor accidents. During accident sequences steam might accumulate as well to stratifications which can directly influence the distribution and ignitability of hydrogen mixtures in containments. An international code benchmark including Computational Fluid Dynamics (CFD) and Lumped Parameter (LP) codes was conducted in the frame of the German THAI program. Basis for the benchmark was experiment TH24.3 which investigates the dissolution of a steam layer subject to natural convection in the steam-air atmosphere of the THAI vessel. The test provides validation data for the development of CFD and LP models to simulate the atmosphere in the containment of a nuclear reactor installation. In test TH24.3 saturated steam is injected into the upper third of the vessel forming a stratification layer which is then mixed by a superposed thermal convection. In this paper the simulation benchmark will be evaluated in addition to the general discussion about the experimental transient of test TH24.3. Concerning the steam stratification build-up and dilution of the stratification, the numerical programs showed very different results during the blind evaluation phase, but improved noticeable during open simulation phase.

Rod Bundle Heat Transfer: Steady-State Steam Cooling Experiments

International Nuclear Information System (INIS)

Spring, J.P.; McLaughlin, D.M.

2006-01-01

Through the joint efforts of the Pennsylvania State University and the United States Nuclear Regulatory Commission, an experimental rod bundle heat transfer (RBHT) facility was designed and built. The rod bundle consists of a 7 x 7 square pitch array with spacer grids and geometry similar to that found in a modern pressurized water reactor. From this facility, a series of steady-state steam cooling experiments were performed. The bundle inlet Reynolds number was varied from 1 400 to 30 000 over a pressure range from 1.36 to 4 bars (20 to 60 psia). The bundle inlet steam temperature was controlled to be at saturation for the specified pressure and the fluid exit temperature exceeded 550 deg. C in the highest power tests. One important quantity of interest is the local convective heat transfer coefficient defined in terms of the local bulk mean temperature of the flow, local wall temperature, and heat flux. Steam temperatures were measured at the center of selected subchannels along the length of the bundle by traversing miniaturized thermocouples. Using an analogy between momentum and energy transport, a method was developed for relating the local subchannel centerline temperature measurement to the local bulk mean temperature. Wall temperatures were measured using internal thermocouples strategically placed along the length of each rod and the local wall heat flux was obtained from an inverse conduction program. The local heat transfer coefficient was calculated from the data at each rod thermocouple location. The local heat transfer coefficients calculated for locations where the flow was fully developed were compared against several published correlations. The Weisman and El-Genk correlations were found to agree best with the RBHT steam cooling data, especially over the range of turbulent Reynolds numbers. The effect of spacer grids on the heat transfer enhancement was also determined from instrumentation placed downstream of the spacer grid locations. The local

Condensation of the steam in the horizontal steam line during cold water flooding

International Nuclear Information System (INIS)

Strubelj, L.; Tiselj, I.

2006-01-01

Direct contact condensation and condensation induced water-hammer in a horizontal pipe was experimentally investigated at PMK-2 test facility of the Hungarian Atomic Energy Research Institute KFKI. The experiment is preformed in the horizontal section of the steam line of the PMK-2 integral test facility. As liquid water floods the horizontal part of the pipeline, the counter current horizontally stratified flow is being observed. During the flooding of the steam line, the vapour-liquid interface area increases and therefore the vapour condensation rate and the vapour velocity also increase. Similar phenomena can occur in the cold/hot leg of the primary loop of PWR nuclear power plant during loss of coolant accident, when emergency core cooling system is activated. Water level at one cross-section and four local void fraction and temperature at the top of steam line was measured and compared with simulation. Condensed steam increases the water temperature that is why the local temperature measurements are the most important information, from which condensation rate can be estimated, since mass of condensed steam was not measured. Free surface simulation of the experiment with thermal phase change model is presented. Surface renewal concept with small eddies is used for calculation of heat transfer coefficient. With surface renewal theory we did not get results similar to experiment, that is why heat transfer coefficient was increased by factor 20. In simulation with heat transfer coefficient calculated with surface renewal concept bubble entrapment is due to reflection of the wave from the end of the pipe. When heat transfer coefficient is increased, condensation rate and steam velocity are also increased, bubble entrapment is due to Kelvin-Helmholtz instability of the free surface, and the results become similar to the measurements. (author)

Hazardous Waste Cleanup: Bristol-Myers Squibb Company Worldwide Medicines in Brunswick, New Jersey

Science.gov (United States)

The Bristol-Myers Squibb Company (BMS) is located at 1 Squibb Drive in New Brunswick, New Jersey. The site has been an active pharmaceutical manufacturing and research and development facility since it began operation by E.R. Squibb and Sons, Inc. in 1907.

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

International Nuclear Information System (INIS)

Scott Hower; Luiza Vladu; Adrian Nichisov; Mihai Cretu

2006-01-01

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)

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)

Korea-France collaboration for the preparation of joint proposal for the OECD project on the steam explosion experiments

International Nuclear Information System (INIS)

Song, Jin Ho; Kim, H. D.; Kim, J. H.; Hong, S. W.; Min, B. T.; Park, I. K.

2004-07-01

There are some difficulties in design of the new reactor containment and in establishment of accident management strategy for operating reactors, since there exist a few phenomenological uncertainties in a steam explosion which occurs at the time of the interaction between the corium melt and coolant. So, the OECD/NEA recommended a research which could finalize the unsolved issues in a steam explosion and then an international collaborative research involving US, France, Germany, Japan and Korea, so called the SERENA (Steam Explosion Resolution for Nuclear Application) started in 2002. The first phase of this collaborative research is an analytical research and the second phase is planned to be an experimental research which will start in 2005. Korea and France agreed that both countries would cooperate in the second phase of the SERENA program at the specialist meetings and collaborative committee between the KAERI (Korea Atomic Energy Research Institute) and the CEA (Commissariat a l'Energie Atomique). This preparation research is performed in order to carry out the agreement between Korea and France and propose a collaborative research to the SERENA second phase. The contents of this research is to perform a collaborative research between Korea and France to propose an international collaborative research on a steam explosion research to the OECD. The scope includes the technical exchange for the construction of an international collaborative research and the preparation of a proposal for an international collaborative research for the OECD steam explosion program. Two steam explosion specialist meetings were held to discuss the technical issues in the TROI and KROTOS experiments. A formal accession and amendment agreement for the collaboration was contracted between the KAERI/KINS/CEA/IRSN. And the KAERI's capability to perform an international collaborative research was proved by advertizing the TROI test results to the USNRC/CSARP and JAERI. Also, the English

Removal of NAPLs from the unsaturated zone using steam: prevention of downward migration by injecting mixtures of steam and air

DEFF Research Database (Denmark)

Schmidt, R.; Gudbjerg, Jacob; Sonnenborg, Torben Obel

2002-01-01

injection technology is presented, where a mixture of steam and air was injected. In twodimensional experiments with unsaturated porous medium contaminated with nonaqueous phase liquids, it was demonstrated how injection of pure steam lead to severe downward migration. Similar experiments, where steam......Steam injection for remediation of porous media contaminated by nonaqueous phase liquids has been shown to be a potentially efficient technology. There is, however, concern that the technique may lead to downward migration of separate phase contaminant. In this work, a modification of the steam...... and air were injected simultaneously, resulted in practically no downward migration and still rapid cleanup was achieved. The processes responsible for the prevention of downward migration when injecting steam–air mixtures were analyzed using a nonisothermal multiphase flow and transport model. Hereby...

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)

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)

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

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)

Experiment on the Influence Factors of Steam Distillation Rate of Crude Oil in Porous Media

Directory of Open Access Journals (Sweden)

Tian Guoqing

2017-01-01

Full Text Available To explore the influence of complexity of reservoir properties in porous media and the diversity of operating conditions on the steam distillation rate of crude oil in the process of heavy oil exploitation with steam injection, steam distillation simulation devices are used to study steam distillation rate of crude oil in porous media. Then steam distillation ratio is obtained under the condition of different core permeability, oil saturation, steam temperatures, system pressure, steam injection rates and steam distillation rates with different viscosities of crude oil. The results show that the steam distillation rate of crude oil in porous media depends mainly on the nature of the crude oil itself, for temperature and pressure are the key factors compared with the pore structure, the initial oil saturation and steam injection rate. The experimental results help estimate the amount of crude oil and the required steam in the reservoir in the steam drive process, aiming to facilitate the optimization design and operation of steam drive.

Toxicity of sediments and pore water from Brunswick Estuary, Georgia

Science.gov (United States)

Winger, Parley V.; Lasier, Peter J.; Geitner, Harvey

1993-01-01

A chlor-alkali plant in Brunswick, Georgia, USA, discharged >2 kg mercury/d into a tributary of the Turtle River-Brunswick Estuary from 1966 to 1971. Mercury concentrations in sediments collected in 1989 along the tributary near the chlor-alkali plant ranged from 1 to 27 μg/g (dry weight), with the highest concentrations found in surface (0–8 cm) sediments of subtidal zones in the vicinity of the discharge site. Toxicity screening in 1990 using Microtox® bioassays on pore water extracted on site from sediments collected at six stations distributed along the tributary indicated that pore water was highly toxic near the plant discharge. Ten-day toxicity tests on pore water from subsequent sediment samples collected near the plant discharge confirmed high toxicity to Hyalella azteca, and feeding activity was significantly reduced in whole-sediment tests. In addition to mercury in the sediments, other metals (chromium, lead, and zinc) exceeded 50 μg/g, and polychlorobiphenyl (PCB) concentrations ranged from 67 to 95 μg/g. On a molar basis, acid-volatile sulfide concentrations (20–45 μmol/g) in the sediments exceeded the metal concentrations. Because acid-volatile sulfides bind with cationic metals and form metal sulfides, which are generally not bioavailable, toxicities shown by these sediments were attributed to the high concentrations of PCBs and possibly methylmercury.

Experiment Plan of High Temperature Steam and Carbon dioxide Co-electrolysis for Synthetic Gas Production

International Nuclear Information System (INIS)

Yoon, Duk-Joo; Ko, Jae-Hwa

2008-01-01

Currently, Solid oxide fuel cells (SOFC) come into the spotlight in the middle of the energy technologies of the future for highly effective conversion of fossil fuels into electricity without carbon dioxide emission. The SOFC is a reversible cell. By applying electrical power to the cell, which is solid oxide electrolysis cell (SOEC), it is possible to produce synthetic gas (syngas) from high temperature steam and carbon dioxide. The produced syngas (hydrogen and carbon monoxide) can be used for synthetic fuels. This SOEC technology can use high temperature from VHTRs for high efficiency. This paper describes KEPRI's experiment plan of high temperature steam and carbon co-electrolysis for syngas production using SOEC technology

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.

Steam Generator Chemical Cleaning Application: Korean Experience in PWR NPP

International Nuclear Information System (INIS)

Hwang, In-Ho; Varrin-Jr, Robert-D.; Little, Michael-J.; Oh, Yeon-Ok; Choo, Seong-Jib; Park, Jin-Hyeok

2012-09-01

Korea Hydro and Nuclear Power (KHNP) performed an EPRI/SGOG chemical cleaning of the secondary side of the steam generators at Ulchin Unit 3 (UCN3) in March 2011 and at Ulchin Unit 4 (UCN4) in September 2011. The steam generator chemical cleaning (SGCC) was performed with venting at the top-of-tube sheet (TTS) and at tube support plates (TSPs) 4, 5, 6, 7, 8, 9, and 10. A primary objective of this SGCC was to address outer diameter stress corrosion cracking (ODSCC), which has been observed at the TTS and TSPs in the UCN3 SGs. The EPRI/SGOG process has been shown to effectively reduce prevailing ODSCC rates at the TTS and TSPs, particularly when applied with periodic venting in this application. This was the first full-length SGCC campaign with venting performed in Korea. Ulchin Unit 3 commenced commercial operation in August 1998 and Ulchin Unit 4 commenced commercial operation in December 1999. UCN3 and UCN4 are a two-loop pressurized water reactor (PWR) of the Korea Standard Nuclear Plant (KSNP) design. The SGs contain high-temperature mill annealed (HTMA) Alloy 600 tubing and are similar in design to the Combustion Engineering CE-80. The KSNP SGs have been susceptible to outer diameter stress corrosion cracking (ODSCC), which is consistent with operating experience for other SGs containing Alloy 600HTMA tubing material. The UCN3/4 SGs have recently begun to experience ODSCC. Hankook Jungsoo Industries Co., Ltd (HaJI) was selected as the cleaning vendor by KHNP. To date, HaJI has completed five Advanced Scale Conditioning Agent (ASCA) cleaning applications and two EPRI/SGOG Steam Generator Chemical Cleaning (SGCC) campaigns for KHNP. The goal of total deposit removal of the applications were successfully achieved and the amounts are 3,579 kg at UCN3 and 3,786 kg at UCN4 which values were estimated before each cleaning by analysing ECT signal and liquid samples from the SGs. The deposits from the SGs were primarily composed of magnetite. There were no chemical

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

Coastal ground water at risk - Saltwater contamination at Brunswick, Georgia and Hilton Head Island, South Carolina

Science.gov (United States)

Krause, Richard E.; Clarke, John S.

2001-01-01

IntroductionSaltwater contamination is restricting the development of ground-water supply in coastal Georgia and adjacent parts of South Carolina and Florida. The principal source of water in the coastal area is the Upper Floridan aquifer—an extremely permeable and high-yielding aquifer—which was first developed in the late 1800s. Pumping from the aquifer has resulted in substantial ground-water-level decline and subsequent saltwater intrusion of the aquifer from underlying strata containing highly saline water at Brunswick, Georgia, and with encroachment of sea-water into the aquifer at the northern end of Hilton Head Island, South Carolina. The saltwater contamination at these locations has constrained further development of the Upper Floridan aquifer in the coastal area and has created competing demands for the limited supply of freshwater. The Georgia Department of Natural Resources, Georgia Environmental Protection Division (GaEPD) has restricted permitted withdrawal of water from the Upper Floridan aquifer in parts of the coastal area (including the Savannah and Brunswick areas) to 1997 rates, and also has restricted additional permitted pumpage in all 24 coastal area counties to 36 million gallons per day above 1997 rates. These actions have prompted interest in alternative management of the aquifer and in the development of supplemental sources of water supply including those from the shallower surficial and upper and lower Brunswick aquifers and from the deeper Lower Floridan aquifer.

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)

Steam plant for pressurized water reactors

International Nuclear Information System (INIS)

Anon.

1983-01-01

This book discusses the research and development organisations and users to highlight those aspects of the steam turbine and associated plant which are particularly related to the PWR system. The contents include: Characteristics of the steam system (including feed train, dump system and safety aspects); overall design aspects of high and half speed turbines; design aspects of the steam generator and seismic considerations; moisture separators and reheaters; feed pumps and their drives; water treatment; safety related valves; operational experience; availability and performance

Steam 80 steam generator instrumentation

International Nuclear Information System (INIS)

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

1980-01-01

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

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

78 FR 64207 - Application To Export Electric Energy; New Brunswick Energy Marketing Corporation (f/k/a New...

Science.gov (United States)

2013-10-28

... United States to Canada as a power marketer for a five-year term using existing international...-year term. NBEMC, a Canadian corporation, is a power marketer in New Brunswick Province, Canada. The...

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)

Steam blowdown experiments with the condensation pool test rig

International Nuclear Information System (INIS)

Purhonen, H.; Puustinen, M.; Laine, J.; Raesaenen, A.; Kyrki-Rajamaeki, R.; Vihavainen, J.

2005-01-01

During a possible loss-of-coolant accident (Local) a large amount of non-condensable (nitrogen) and condensable (steam) gas is blown from the upper drywell of the containment to the condensation pool through the blowdown pipes at the boiling water reactors (BWRs). The wet well pool serves as the major heat sink for condensation of steam. The blowdown causes both dynamic and structural loads to the condensation pool. There might also be a risk that the gas discharging to the pool could push its way to the emergency core cooling systems (ECCS) and undermine their performance. (author)

Steam explosion triggering and efficiency studies

International Nuclear Information System (INIS)

Buxton, L.D.; Nelson, L.S.; Benedick, W.B.

1979-01-01

A program at Sandia Laboratories to provide relevant data on the interaction of molten LWR core materials with water is described. Two different subtasks were established. The first was the performance of laboratory-scale experiments to investigate the ability to trigger steam explosions for realistic LWR core melt simulants under a wide range of initial conditions. The second was the performance of field-scale experiments to investigate the efficiency of converting the thermal energy of the melt into mechanical work in much larger steam explosions

Flow-induced vibration analysis of a helical coil steam generator experiment using large eddy simulation

Energy Technology Data Exchange (ETDEWEB)

Yuan, Haomin; Solberg, Jerome; Merzari, Elia; Kraus, Adam; Grindeanu, Iulian

2017-10-01

This paper describes a numerical study of flow-induced vibration in a helical coil steam generator experiment conducted at Argonne National Laboratory in the 1980s. In the experiment, a half-scale sector model of a steam generator helical coil tube bank was subjected to still and flowing air and water, and the vibrational characteristics were recorded. The research detailed in this document utilizes the multi-physics simulation toolkit SHARP developed at Argonne National Laboratory, in cooperation with Lawrence Livermore National Laboratory, to simulate the experiment. SHARP uses the spectral element code Nek5000 for fluid dynamics analysis and the finite element code DIABLO for structural analysis. The flow around the coil tubes is modeled in Nek5000 by using a large eddy simulation turbulence model. Transient pressure data on the tube surfaces is sampled and transferred to DIABLO for the structural simulation. The structural response is simulated in DIABLO via an implicit time-marching algorithm and a combination of continuum elements and structural shells. Tube vibration data (acceleration and frequency) are sampled and compared with the experimental data. Currently, only one-way coupling is used, which means that pressure loads from the fluid simulation are transferred to the structural simulation but the resulting structural displacements are not fed back to the fluid simulation

Operating experience with steam generator water chemistry in Japanese PWR plants

International Nuclear Information System (INIS)

Onimura, K.; Hattori, T.

1991-01-01

Since the first PWR plant in Japan started its commercial operation in 1970, seventeen plants are operating as of the end of 1990. First three units initially applied phosphate treatment as secondary water chemistry control and then changed to all volatile treatment (AVT) due to phosphate induced wastage of steam generator tubing. The other fourteen units operate exclusively under AVT. In Japan, several corrosion phenomena of steam generator tubing, resulted from secondary water chemistry, have been experienced, but occurrence of those phenomena has decreased by means of improvement on impurity management, boric acid treatment and high hydrazine operation. Recently secondary water chemistry in Japanese plants are well maintained in every stage of operation. This paper introduces brief summary of the present status of steam generators and secondary water chemistry in Japan and ongoing activities of investigation for future improvement of reliability of steam generator. History and present status of secondary water chemistry in Japanese PWRs were introduced. In order to get improved water chemistry, the integrity of secondary system equipments is essential and the improvement in water chemistry has been achieved with the improvement in equipments and their usage. As a result of those efforts, present status of secondary water is excellent. However, further development for crevice chemistry monitoring technique and an advanced water chemistry data management system is desired for the purpose of future improvement of reliability of steam generator

Influence of steam generator tube ruptures during semiscale loss-of-coolant experiments

International Nuclear Information System (INIS)

Cozzuol, J.M.; Larson, T.K.

1978-01-01

Tests which simulated rupture of steam generator tubes during loss-of-coolant experiments in a PWR type system have been conducted in the Semiscale Mod-1 system. Analysis of test data indicates that high rod cladding temperatures occured only for a band of tube ruptures (between 12 and 20 tubes) and that the peak cladding temperatures attained within this band were strongly dependent on the magnitude of the tube rupture flow rates. Maximum cladding temperature of about 1255 K was observed for tests which simulated tube ruptures within this narrow band. (author)

Current experience and a new modeling on water hammer due to steam condensation in PWR secondary system

International Nuclear Information System (INIS)

Kawanishi, K.; Kasahara, J.; Ueno, T.; Suzuta, T.

1998-01-01

There have been possibilities to occur water hammer in pipelines of turbine system for nuclear or fossil fuel power plants. According to the NUREG report, approximately 150 events have been reported since 1969, we also have an experience recently. Water hammer occurs due to sudden steam condensation with pressure pulse. This kind of pressure pulses has been made by alternative producing and condensing of steam slug in the pipe and its frequency relates subcooling and pipe structures. This paper presents our current experience on water hammer with some experimental studies. The present experiment has been performed to obtain the data base for evaluating the pressure pulses. The test pipe was horizontal tubes with dead end connected to vertical tube which simulating drain line in PWR secondary system. The main results are shown as follows; Magnitude of pressure pulse depends drain velocity and initial subcooling. Pipe structure effects on the frequency and continual time of water hammer phenomenon. A new modeling for quantitative explanation of the phenomena is also presented

An Isothermal Steam Expander for an Industrial Steam Supplying System

Directory of Open Access Journals (Sweden)

Chen-Kuang Lin

2015-01-01

Full Text Available Steam is an essential medium used in the industrial process. To ensure steam quality, small and middle scale boilers are often adopted. However, because a higher steam pressure (compared to the necessary steam pressure is generated, the boiler’s steam pressure will be reduced via a pressure regulator before the steam is directed through the process. Unfortunately, pressure is somewhat wasted during the reducing process. Therefore, in order to promote energy efficiency, a pressure regulator is replaced by a steam expander. With this steam expander, the pressure will be transformed into mechanical energy and extracted during the expansion process. A new type of isothermal steam expander for an industrial steam supplying system will be presented in the paper. The isothermal steam expander will improve the energy efficiency of a traditional steam expander by replacing the isentropic process with an isothermal expansion process. With this, steam condensation will decrease, energy will increase, and steam quality will be improved. Moreover, the mathematical model of the isothermal steam expander will be established by using the Schmidt theory, the same principle used to analyze Stirling engines. Consequently, by verifying the correctness of the theoretical model for the isothermal steam expander using experimental data, a prototype of 100 c.c. isothermal steam expander is constructed.

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)

Steam generator tube performance: experience with water-cooled nuclear power reactors during 1978

International Nuclear Information System (INIS)

Tatone, O.S.; Pathania, R.S.

1980-02-01

The performance of steam generator tubes in water-cooled nuclear power reactors has been reviewed for 1978. Tube failures occurred at 31 of the 86 reactors surveyed. Causes of these failures and procedures designed to deal with them are described. A dramatic decrease in the number of tubes plugged was evident in 1978 compared to the previous year. This is attributed to diligent application of techniques developed from in-plant experience and research and development programs over the past several years. (auth)

Preliminary results of thermal igniter experiments in H2-air-steam environments

International Nuclear Information System (INIS)

Lowry, W.

1981-01-01

Thermal igniters (glow plugs), proposed by the Tennessee Valley Authority for intentional ignition of hydrogen in nuclear reactor containment, have been tested for functionability in mixtures of air, hydrogen, and steam. Test environments included 6% to 16% hydrogen concentrations in air, and 8%, 10%, and 12% hydrogen in mixtures with 30% and 40% steam fractions. All were conducted in a 10.6 ft 3 insulated pressure vessel. For all of these tests the glow plug successfully initiated combustion. Dry air/hydrogen tests exhibited a distinct tendency for complete combustion at hydrogen concentrations between 8% and 9%. Steam suppressed both peak pressures and completeness of combustion. No combustion could be initiated at or above a 50% steam fraction. Circulation of the mixture with a fan increased the completeness of combustion. The glow plug showed no evidence of performance degradation throughout the program

A three-dimensional laboratory steam injection model allowing in situ saturation measurements

Energy Technology Data Exchange (ETDEWEB)

Demiral, B.M.R.; Pettit, P.A.; Castanier, L.M.; Brigham, W.E.

1992-08-01

The CT imaging technique together with temperature and pressure measurements were used to follow the steam propagation during steam and steam foam injection experiments in a three dimensional laboratory steam injection model. The advantages and disadvantages of different geometries were examined to find out which could best represent radial and gravity override flows and also fit the dimensions of the scanning field of the CT scanner. During experiments, steam was injected continuously at a constant rate into the water saturated model and CT scans were taken at six different cross sections of the model. Pressure and temperature data were collected with time at three different levels in the model. During steam injection experiments, the saturations obtained by CT matched well with the temperature data. That is, the steam override as observed by temperature data was also clearly seen on the CT pictures. During the runs where foam was present, the saturation distributions obtained from CT pictures showed a piston like displacement. However, the temperature distributions were different depending on the type of steam foam process used. The results clearly show that the pressure/temperature data alone are not sufficient to study steam foam in the presence of non-condensible gas.

Experience of research, design, capacity, and operation with forced flow steam generators

International Nuclear Information System (INIS)

Bertolotti, G.; McDonald, B.N.; Pocock, F.J.

1975-01-01

The forced flow steam generators in operation in six American nuclear power plants show an excellent operational behaviour. The concept for this type of steam generator has been developed in the USA, and it has been successfully tested over several years regarding its suitability for PWRs of a larger size. The results concerning construction, materials and water chemistry for this steam generator, which will be used for the first time in the FRG in the nuclear power station Muelheim-Kaerlich, have confirmed the high reliability of this high-efficiency component. (orig./LN) [de

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.

Community Geothermal Technology Program: Cloth dyeing by geothermal steam. An experiment in technology transfer from Japan to Hawaii, Final report

Energy Technology Data Exchange (ETDEWEB)

Furumoto, A.S.

1987-12-31

This was an experiment to test whether cloth dyeing using geothermal steam (already proven in Japan) would be feasible in Hawaii. Results: Using a fabricated steam vat, cotton, silk, and synthetic can be dyed; the resulting material received high grades for steadfastness and permanency under dye testing. Techniques that were successful in Matsukawa, were replicated in Puna. However, attempts to embed leaf patterns on cloth using natural leaves and to extract natural dyes from Hawaiian plants were unsuccessful; the color of natural dyes deteriorated in hours. But chemical dyes gave brilliant hues or shades, in contrast to those in Japan where the steam there gave subdued tones. It is concluded that geothermal dyeing can be a viable cottage industry in Puna, Hawaii.

Solid waste processing experience at Susquehanna Steam Electric Station

International Nuclear Information System (INIS)

Phillips, J.W.; Granus, M.W.

1984-01-01

This paper reviews the first year's operation at the Susquehanna Steam Electric Station (SSES) with respect to the Westinghouse Hittman Nuclear Incorporated (Hittman) mobile solidification system and the dry activated waste generation, handling and processing. Experiences pertinent to the mobile solidification system are reviewed with emphasis on the integration of the system into the plant, problems associated with unexpected waste properties and the myriad of operating procedures that had to be prepared. The processing history for 1983 is reviewed in terms of the volume of waste, including solidified wastes, dewatered wastes an DAW. Factors that must be considered in evaluating processing alternatives, i.e., dewatering vs. solidification; steel liners vs. HICs, are discussed. Actions taken by Hittman and SSES to maximize the processing economics are also discussed. Finally, recommendations are provided to the utility considering implementing mobile solification services to ensure a smooth and timely integration of services into the plant

The chemical kinetics of the reactions of lithium with steam-air mixtures

International Nuclear Information System (INIS)

Barnett, D.S.; Kazimi, M.S.

1989-04-01

This work involved the experimental and analytical determination of the consequences of lithium fires in the presence of steam. Experiments were performed to characterize the chemical reactions of lithium with steam-nitrogen and steam-air mixtures. Models were introduced in the LITFIRE code to describe lithium fires in the presence of steam inside the containment building and plasma chamber of a hypothetical fusion reactor. The code was also equipped with the capability to determine the effects of decay heat and lithium fire on the temperature response of the reactor first wall in the event of a coolant disturbance. Forty-two kinetics experiments were performed in which a stream of steam-nitrogen or steam-air was passed over and reacted with approximately three grams of lithium heated to a predetermined temperature. The lithium reaction rates with the constituent gases were measured and characterized for a wide range of lithium temperatures and gas compositions. Experiments were performed with steam molar concentrations of 5, 15 and 30% and lithium temperatures ranging from 400 to 1100 degree C, inclusive. The LITFIRE code was modified to enable it to model the interactions of lithium with steam-air atmospheres. Results of the reaction kinetics experiments were used in the reaction model, and the heat transfer model was expanded to allow it to handle condensible atmospheres. Three groups of accidents were investigated: a spill on the containment building floor, a spill inside the reactor plasma chamber, and a spill inside the plasma chamber with steam injection to the containment building simulating a steam line break. The results were compared to dry air cases under the same conditions. 23 refs., 66 figs., 18 tabs

Molten salt steam generator subsystem research experiment. Volume I. Phase 1 - Final report

Energy Technology Data Exchange (ETDEWEB)

None

1984-10-01

A study was conducted for Phase 1 of a two-phase project whose objectives were to develop a reliable, cost-effective molten salt steam generating subsystem for solar thermal plants, minimize uncertainty in capital, operating, and maintenance costs, and demonstrate the ability of molten salt to generate high-pressure, high-temperature steam. The Phase 1 study involved the conceptual design of molten salt steam generating subsystems for a nominal 100-MWe net stand-alone solar central receiver electric generating plant, and a nominal 100-MWe net hybrid fossil-fueled electric power generating plant that is 50% repowered by a solar central receiver system. As part of Phase 1, a proposal was prepared for Phase 2, which involves the design, construction, testing and evaluation of a Subsystem Research Experiment of sufficient size to ensure successful operation of the full-size subsystem designed in Phase 1. Evaluation of several concepts resulted in the selection of a four-component (preheater, evaporator, superheater, reheater), natural circulation, vertically oriented, shell and tube (straight) heat exchanger arrangement. Thermal hydraulic analysis of the system included full and part load performance, circulation requirements, stability, and critical heat flux analysis. Flow-induced tube vibration, tube buckling, fatigue evaluation of tubesheet junctions, steady-state tubesheet analysis, and a simplified transient analysis were included in the structural analysis of the system. Operating modes and system dynamic response to load changes were identified. Auxiliary equipment, fabrication, erection, and maintenance requirements were also defined. Installed capital costs and a project schedule were prepared for each design.

Dismantling Experiment of Mock-up Tube Bundle of Steam Generator

International Nuclear Information System (INIS)

Kim, Sung Kyun; Lee, Kune Woo

2010-01-01

A SG (steam generator) is one of the biggest decommissioning components in nuclear power plants and one has been replaced 2∼6 times during the whole operation of a nuclear power plant. The old SG should be decommissioned for the purpose of the volume reduction of radioactive waste. Among the components of SG, the tube bundle is one of the most difficult items to be dismantled due to the fact that it is very hard to cut since it is made of Inconel 600 which has high resistance of corrosion and abrasion. Moreover, All cutting process should be performed by remotely since radioactive contamination of the internal surface of SG tubes is very high (about 150,000∼300,000 Bq/cm 2 ). Therefore, it is necessary to choose the appropriate cutting methods by the pros and cons analysis for candidate dismantling technologies and to do experiment study for the validation. In this study, the results of cutting experiment for a mock-up bundle by using band saw cutting method are described herein

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)

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)

Numerical Analysis on Transient of Steam-gas Pressurizer

International Nuclear Information System (INIS)

Kim, Jong-Won; Lee, Yeon-Gun; Park, Goon-Cherl

2008-01-01

In nuclear reactors, various pressurizers are adopted to satisfy their characteristics and uses. The additional active systems such as heater, pressurizer cooler, spray and insulator are essential for a steam or a gas pressurizer. With a steam-gas pressurizer, additional systems are not required due to the use of steam and non-condensable gas as pressure-buffering materials. The steam-gas pressurizer in integrated small reactors experiences very complicated thermal-hydraulic phenomena. To ensure the integrity of this pressurizer type, the analysis on the transient behavior of the steam-gas pressure is indispensable. For this purpose, the steam-gas pressurizer model is introduced to predict the accurate system pressure. The proposed model includes bulk flashing, rainout, inter-region heat and mass transfer and wall condensation with non-condensable gas. However, the ideal gas law is not applied because of significant interaction at high pressure between steam and non-condensable gas. The results obtained from this proposed model agree with those from pressurizer tests. (authors)

Detection of steam leaks into sodium in fast reactor steam generators by acoustic techniques - An overview of Indian programme

International Nuclear Information System (INIS)

Prabhakar, R.; Vyjayanthi, R.K.; Kale, R.D.

1990-01-01

Realising the potential of acoustic leak detection technique, an experimental programme was initiated a few years back at Indira Gandhi Centre for Atomic Research (IGCAR) to develop this technique. The first phase of this programme consists of experiments to measure background noise characteristics on the steam generator modules of the 40 MW (thermal) Fast Breeder Test Reactor (FBTR) at Kalpakkam and experiments to establish leak noise characteristics with the help of a leak simulation set up. By subjecting the measured data from these experiments to signal analysis techniques, a criterion for acoustic leak detection for FBTR steam generator will be evolved. Second phase of this programme will be devoted to developing an acoustic leak detection system suitable for installation in the 500 MWe Prototype Fast Breeder Reactor (PFBR). This paper discusses the first phase of the experimental programme, results obtained from measurements carried out on FBTR steam generators and results obtained from leak simulation experiments. Acoustic leak detection system being considered for PFBR is also briefly described. 4 refs, 8 figs, 1 tab

More than wind: evaluating renewable energy opportunities for First Nations in Nova Scotia and New Brunswick

Energy Technology Data Exchange (ETDEWEB)

Campbell, Diana [MREM Canada (Canada)

2011-04-15

The Nova Scotia and New Brunswick governments have committed to developing the renewable energy sector in the coming years. However, due to a lack of information, First Nations are not familiar with renewable energy technologies and their economic opportunities. The aim of this paper is to provide First Nations with information on the different renewable energies. It has been noticed that First Nations have not seized either the opportunities offered by the Nova Scotia feed-in tariff nor the New Brunswick community energy policy and an overview of these policies is provided. In addition, information on renewable energy technologies is presented along with potential opportunities specific to each First Nation in the 2 provinces. This paper provides First Nations with useful information on renewable energy and with recommendations for immediate and long term action; it is expected that this document will result in a greater involvement of First Nations in the renewable energy sector.

Steam turbine cycle

International Nuclear Information System (INIS)

Okuzumi, Naoaki.

1994-01-01

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

Two-dimensional modeling of water spray cooling in superheated steam

Directory of Open Access Journals (Sweden)

Ebrahimian Vahid

2008-01-01

Full Text Available Spray cooling of the superheated steam occurs with the interaction of many complex physical processes, such as initial droplet formation, collision, coalescence, secondary break up, evaporation, turbulence generation, and modulation, as well as turbulent mixing, heat, mass and momentum transfer in a highly non-uniform two-phase environment. While it is extremely difficult to systematically study particular effects in this complex interaction in a well defined physical experiment, the interaction is well suited for numerical studies based on advanced detailed models of all the processes involved. This paper presents results of such a numerical experiment. Cooling of the superheated steam can be applied in order to decrease the temperature of superheated steam in power plants. By spraying the cooling water into the superheated steam, the temperature of the superheated steam can be controlled. In this work, water spray cooling was modeled to investigate the influences of the droplet size, injected velocity, the pressure and velocity of the superheated steam on the evaporation of the cooling water. The results show that by increasing the diameter of the droplets, the pressure and velocity of the superheated steam, the amount of evaporation of cooling water increases. .

Steam turbines for PWR stations

International Nuclear Information System (INIS)

Muscroft, J.

1989-01-01

The thermodynamic cycle requirements and mechanical design features applying to modern GEC 3000 rev/min steam turbines for pressurised water reactor power stations are reviewed. The most recent developments include machines of 630 MW and 985 MW output which are currently under construction. The importance of service experience with nuclear wet steam turbines associated with a variety of types of water cooled reactor and its relevance to the design of modern 3000 rev/min turbines for pressurised water reactor applications is emphasised. (author)

Online monitoring of steam/water chemistry of a fast breeder test reactor

International Nuclear Information System (INIS)

Subramanian, K.G.; Suriyanarayanan, A.; Thirunavukarasu, N.; Naganathan, V.R.; Panigrahi, B.S.; Jambunathan, D.

2005-01-01

Operating experience with the once-through steam generator of a fast breeder test reactor (FBTR) has shown that an efficient water chemistry control played a major role in minimizing corrosion related failures of steam generator tubes and ensuring steam generator tube integrity. In order to meet the stringent feedwater and steam quality specifications, use of fast and sensitive online monitors to detect impurity levels is highly desirable. Online monitoring techniques have helped in achieving feedwater of an exceptional degree of purity. Experience in operating the online monitors in the steam/water system of a FBTR is discussed in detail in this paper. In addition, the effect of excess hydrazine in the feedwater on the steam generator leak detection system and the need for a hydrazine online meter are also discussed. (orig.)

Steam condenser

International Nuclear Information System (INIS)

Masuda, Fujio

1980-01-01

Purpose: To enable safe steam condensation by providing steam condensation blades at the end of a pipe. Constitution: When high temperature high pressure steam flows into a vent pipe having an opening under water in a pool or an exhaust pipe or the like for a main steam eacape safety valve, non-condensable gas filled beforehand in the steam exhaust pipe is compressed, and discharged into the water in the pool. The non-condensable gas thus discharged from the steam exhaust pipe is introduced into the interior of the hollow steam condensing blades, is then suitably expanded, and thereafter exhausted from a number of exhaust holes into the water in the pool. In this manner, the non-condensable gas thus discharged is not directly introduced into the water in the pool, but is suitable expanded in the space of the steam condensing blades to suppress extreme over-compression and over-expansion of the gas so as to prevent unstable pressure vibration. (Yoshihara, H.)

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

Report: The New Brunswick Indian Arts and Crafts Association Workshop for Board Members. No. 165.

Science.gov (United States)

Goddu, Roland; Nicholas, Darryl

This is a report on a training session conducted for the board members of the New Brunswick Indian Arts and Crafts Association with the aim of outlining purposes of the Association, and assisting in drawing up plans for a formal administrative structure among arts and crafts organizations. An administrative manual containing terms of reference…

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

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

CANDU steam generator aging management: Some perspectives after 20 years in-service experience

International Nuclear Information System (INIS)

Tapping, R.I.; Nickerson, J.; Subash, N.; Roy, S.

2002-01-01

assessment of the future field data. Subtle changes to plant operation and also the unique plant-specific details of materials and design often can have a significant impact on the life capability of this complicated equipment. From the plant-specific assessments, and from the experience from our R and D program and utility interaction, AECL has developed a detailed proactive SG aging management strategy for long and reliable life capability (to 50 years). This proactive strategy has a number of important elements that are recommended to be incorporated into current plant programs. With CANDU plants continuing operation with aging equipment, such as steam generators, AECL is continuing to undertake comprehensive programs to support operations and provide enhanced technologies for performance attainment and improvement in inspection, surveillance and performance/safety analysis methodologies. These technologies are also continuously improved by programs that are targeted to important in-service degradation mechanisms, understanding of the system/equipment tolerance to this degradation, and to measures that can be applied to both new reactors and back fitted to existing plants. Systematic and continuous feedback of experience from operations to the performance programs is another important part of the approach. This paper will outline the results of several CANDU SG condition assessments, and place the recommendations from these assessments in context with effective aging Management programs that need to be followed to achieve economic life. For CANDU 6 utilities, which have had a very good steam generator service record with little or no significant active degradation to date, the paper will outline the approach to proactive aging Management that will see equipment through to long and reliable extended life. Also mentioned will be examples of AECL performance improvement and operations support programs for steam generators. (author)

Thermal-hydraulic experiments for the PCHE type steam generator

International Nuclear Information System (INIS)

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

2015-01-01

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

Steam cracking of hydrocarbons 3. Straight-run naphtha

NARCIS (Netherlands)

Bajus, M.; Vesely, V.; Leclercq, P.A.; Rijks, J.A.

1980-01-01

Steam cracking of straight-run naphtha from Romashkino crude oil was investigated in quartz and stainless steel reactors with a relatively large ratio of Inner surface to volume. The experiments were performed at atmospheric pressure at 780-800 OC for starting ratios of steam to naphtha between 0.5

Reactor physics studies in the steam flooded GCFR-Phase II critical assembly

International Nuclear Information System (INIS)

Bhattacharyya, S.K.

1978-08-01

A possible accident scenario in a Gas-Cooled Fast Reactor (GCFR) is the leakage of secondary steam into the core. Considerable analytical effort has gone into the study of the effects of such an accidental steam entry. The work described represents the first full scale experimental study of the steam-entry phenomenon in GCFRs. The reference GCFR model used for the study was the benchmark GCFR Phase II assembly, and polyethylene foam was used to provide a very homogeneous steam simulation. The reactivity worth of steam entry was measured for three different steam densities. In addition, a set of integral physics parameters were measured in the largest steam density (0.008 g/cm 3 ) configuration. The corresponding parameters were also measured in dry reference GCFR critical assembly for comparison. The experiments were analyzed using ENDF/B-IV data and two-dimensional diffusion theory methods. As in earlier GCFR critical experiments analysis, the Benoist method was used to treat the problem of neutron streaming

The modification of main steam safety valves in Qinshan phase Ⅱ expansion project

International Nuclear Information System (INIS)

Chen Haiqiao

2012-01-01

The main steam safety valves of NPP steam system are second- class nuclear safety component. It used to limit the pressure of SG secondary side and main steam system via emitting steam into the environment. At present, the main steam safety valves have mechanical valves and assisted power valves. According to the experience of power plants at home and abroad, including Qinshan Phase Ⅱ unit 1/2 experience feedback, Qinshan Phase Ⅱ expansion project made modification on valve type, setting value and valve body. This paper introduce the characteristics of different safety valve types, the modification of main steam safety valves and the modification analysis on safety issues.security and impact on the other systems in Qinshan Phase Ⅱ expansion project. (author)

ECO steam explosion experiments on the conversion of thermal into mechanical energy

International Nuclear Information System (INIS)

Cherdron, W.; Kaiser, A.; Schuetz, W.; Will, H.

2001-01-01

In case of a steam explosion, e.g. as a consequence of a severe reactor accident, part of the thermal energy of the melt is transferred into mechanical energy. At Forschungszentrum Karlsruhe, so-called ECO experiments, are being directed to measure the conversion factor under well-defined conditions. In ECO, alumina from a thermite reaction is used as a simulating material instead of corium. Dimensions of the test facility as well as major test conditions, e.g. temperature and release mode of the melt, water inventory and test procedure, are based on the former PREMIX experimental series. In the paper, results of the first test, ECO 01, are given. (orig.)

Steam-generator tube performance: world experience with water-cooled nuclear power reactors during 1978

International Nuclear Information System (INIS)

Tatone, O.S.; Pathania, R.S.

1980-01-01

The performance of steam-generator tubes in water-cooled nuclear power reactors during 1978 is reviewed. Tube failures occurred at 31 of the 86 reactors surveyed. The causes of these failures and the procedures designed to deal with them are described. The number of tubes plugged has decreased dramatically in 1978 compared to the previous year. This is attributed to the diligent application of techniques developed through in-plant experience and research and development programs over the past several years

Herpes simplex virus type 1 is the leading cause of genital herpes in New Brunswick.

Science.gov (United States)

Garceau, Richard; Leblanc, Danielle; Thibault, Louise; Girouard, Gabriel; Mallet, Manon

2012-01-01

Little is known about the role of herpes simplex virus (HSV) type 1 (HSV1) in the epidemiology of genital herpes in Canada. Data on herpes viral cultures for two consecutive years obtained from L'Hôpital Dr GL Dumont, which performs all the viral culture testing in New Brunswick, were reviewed. It was hypothesized that HSV1 was the main cause of genital herpes in New Brunswick. Samples of genital origin sent to the laboratory for HSV culture testing between July 2006 and June 2008 were analyzed. Samples from an unspecified or a nongenital source were excluded from analysis. Multiple positive samples collected from the same patient were pooled into a single sample. HSV was isolated from 764 different patients. HSV1 was isolated in 62.6% of patients (male, 55%; female, 63.8%). HSV1 was isolated in 73.2% of patients 10 to 39 years of age and in 32% of patients ≥40 years of age. The difference in rates of HSV1 infection between the 10 to 39 years of age group and the ≥40 years of age group was statistically significant (Pgenital site. Significant rate differences were demonstrated between the groups 10 to 39 years of age and ≥40 years of age. Little is known about the role of herpes simplex virus (HSV) type 1 (HSV1) in the epidemiology of genital herpes in Canada. Data on herpes viral cultures for two consecutive years obtained from L’Hôpital Dr GL Dumont, which performs all the viral culture testing in New Brunswick, were reviewed. It was hypothesized that HSV1 was the main cause of genital herpes in New Brunswick. Samples of genital origin sent to the laboratory for HSV culture testing between July 2006 and June 2008 were analyzed. Samples from an unspecified or a nongenital source were excluded from analysis. Multiple positive samples collected from the same patient were pooled into a single sample. HSV was isolated from 764 different patients. HSV1 was isolated in 62.6% of patients (male, 55%; female, 63.8%). HSV1 was isolated in 73.2% of patients 10 to

Experience and modeling of radioactivity transport following steam generator tube rupture

International Nuclear Information System (INIS)

Hopenfeld, J.

1985-01-01

A review of the capabilities of the CITADEL computer code as well as plant experience to project radioactivity releases following a steam generator tube rupture in pressurized-water reactors shows that certain experimental data are needed for reliable offsite dose predictions. This article defines five parameters that are the key for such predictions and discusses the functional dependence of these parameters on various operational variables. The above parameters can be used in conjuction with CITADEL or they can be inserted in the appropriate equations, which then can be programmed conveniently as a subroutine in thermal-hydraulic system codes. A joint Westinghouse Electric Corporation, Electric Power Research Institute, and Nuclear Regulatory Commission program aimed at obtaining the five parameters empirically is described

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

International Nuclear Information System (INIS)

Cao Yuhua; Xiao Bo; He Liu; Huang Min

2014-01-01

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

An Improved Steam Injection Model with the Consideration of Steam Override

OpenAIRE

He , Congge; Mu , Longxin; Fan , Zifei; Xu , Anzhu; Zeng , Baoquan; Ji , Zhongyuan; Han , Haishui

2017-01-01

International audience; The great difference in density between steam and liquid during wet steam injection always results in steam override, that is, steam gathers on the top of the pay zone. In this article, the equation for steam override coefficient was firstly established based on van Lookeren’s steam override theory and then radius of steam zone and hot fluid zone were derived according to a more realistic temperature distribution and an energy balance in the pay zone. On this basis, th...

Investigation of brush seals for application in steam turbines

International Nuclear Information System (INIS)

Zorn, Peter

2012-01-01

Brush seals have high potential for efficiency increase compared to conventional labyrinth seals in steam turbines. Due to less experience in operation today there is a lot of scepticism with customers of steam turbine manufacturers. Therefore this thesis is investigating characteristics of this type of seal. Experiments and numerical models will be presented, which lead to better knowledge about leakages and influence of flow through seal onto dynamics of rotor in comparison to labyrinth seals. This thesis is increasing area of experience and one more positive reference.

Proceedings of the specialists' meeting on steam generator failure and failure propagation experience, held in Aix-en Provence, France, 26-28 September 1990

International Nuclear Information System (INIS)

Maupre, J.P.

1990-09-01

The 35 participants, representing 7 Member States and one International Organization discussed recent investigations on leak development and propagation in LMFBR steam generators. The meeting was divided into three technical sessions: review of national status on studies of failure propagation (8 papers); propagation experience on reactor steam generators (4 papers); studies of failure propagation: codes, hydrogen detection, tests (11 papers). A separate abstract was prepared for each of these papers

Steam turbine installations

International Nuclear Information System (INIS)

Bainbridge, A.

1976-01-01

The object of the arrangement described is to enable raising steam for driving steam turbines in a way suited to operating with liquid metals, such as Na, as heat transfer medium. A preheated water feed, in heat transfer relationship with the liquid metals, is passed through evaporator and superheater stages, and the superheated steam is supplied to the highest pressure stage of the steam turbine arrangement. Steam extracted intermediate the evaporator and superheater stages is employed to provide reheat for the lower pressure stage of the steam turbine. Only a major portion of the preheated water feed may be evaporated and this portion separated and supplied to the superheater stage. The feature of 'steam to steam' reheat avoids a second liquid metal heat transfer and hence represents a simplification. It also reduces the hazard associated with possible steam-liquid metal contact. (U.K.)

Technology of turbine plant operating with wet steam

International Nuclear Information System (INIS)

1989-01-01

The technology of turbine plant operating with wet steam is a subject of continuing interest and importance, notably in view of the widespread use of wet steam cycles in nuclear power plants and the recent developments of advanced low pressure blading for both conventional and wet steam turbines. The nature of water formation in expanding steam has an important influence on the efficiency of turbine blading and on the integrity and safe operating life of blading and associated turbine and plant components. The subjects covered in this book include research, flow analysis and measurement, development and design of turbines and ancillary plant, selection of materials of construction, manufacturing methods and operating experience. (author)

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

experiments conducted with SGTF steam generator and the performance is compared with the theoretically predicted results obtained from numerical study. (authors)

New Brunswick air quality monitoring results for the year 2007 : executive summary

International Nuclear Information System (INIS)

2009-01-01

Air quality has been monitored in New Brunswick since the 1960s. This report summarized air quality results for general public information, with emphasis on air quality assessment in relation to existing air quality standards and objectives. Long-term trend data were also presented for representative sites. Air contaminants currently covered by provincial objectives were measured at 59 sites across the province during 2007. Acid rain was measured at 13 additional sites. Some locations were monitored for volatile organic compounds and mercury in air. Quality assurance procedures used in the provincial air quality system were also described. The report revealed that there were no exceedances of New Brunswick air quality objectives for nitrogen dioxide or carbon monoxide at any of the provincial monitoring sites in 2007. In many instances, exceedances for ozone, total reduced sulphur, fine particulate matter and total volatile organic compound concentrations were lower in 2007 than in 2006. Air quality trends indicate that since the late 1970s and 1980s, air quality has improved for all pollutants currently being measured, with the possible exception of ground level ozone. Annual average levels of sulphur dioxide have decreased significantly over the past 15-20 years. The long term levels of carbon monoxide and nitrogen dioxide have also decreased. Acid deposition has declined since the early 1990s, but its effects continue to be of concern in the province. In 2007, sulphate in precipitation was moderately lower than in 2006.

Specialists' meeting on maintenance and repair of LMFBR steam generators. Summary report

International Nuclear Information System (INIS)

2002-01-01

The purpose of the meeting was to review and discuss the experience accumulated in various countries on the general design philosophy of LMFBR steam generators from the view point of maintenance and repair, in-service inspection of steam generator tube bundles, identification and inspection of failed tubes and the cleaning and repairing of failed steam generators. The following main topic areas were discussed by participants: National review presentations on maintenance and repair of LMFBR steam generators - design philosophy for maintenance and repair; Research and Development work on maintenance and repair; Experience on steam generator maintenance and repair. During the meeting papers were presented by the participants on behalf of their countries and organizations. A final discussion session was held and summaries, general conclusions and recommendations were approved by consensus

Steam as turbine blade coolant: Experimental data generation

Energy Technology Data Exchange (ETDEWEB)

Wilmsen, B.; Engeda, A.; Lloyd, J.R. [Michigan State Univ., East Lansing, MI (United States)

1995-10-01

Steam as a coolant is a possible option to cool blades in high temperature gas turbines. However, to quantify steam as a coolant, there exists practically no experimental data. This work deals with an attempt to generate such data and with the design of an experimental setup used for the purpose. Initially, in order to guide the direction of experiments, a preliminary theoretical and empirical prediction of the expected experimental data is performed and is presented here. This initial analysis also compares the coolant properties of steam and air.

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-28-2 (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

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

Investigating mental fitness and school connectedness in Prince Edward Island and New Brunswick, Canada.

Science.gov (United States)

Murnaghan, Donna; Morrison, William; Laurence, Courtney; Bell, Brandi

2014-07-01

As youth struggle with anxiety and depression, promoting positive mental fitness is a primary concern. Canadian school-based mental health programs that focus on positive psychology and positive mental health initiatives emphasize safe and supportive environments, student engagement, resilience, and self-determination. This study examined predictors of mental fitness and its 3 components (autonomy, competence, and relatedness). School Health Action Planning and Evaluation System-Prince Edward Island (SHAPES-PEI) and the New Brunswick Student Wellness Survey (NB SWS) are data collection and feedback systems that survey youth about 4 health behaviors. Grade 7-12 students in Prince Edward Island (N = 3318) and New Brunswick (N = 7314) completed a mental fitness questionnaire in 2008-2009 (PEI) and 2006-2007 (NB). Four linear regression models were conducted to examine student characteristics associated with mental fitness, autonomy, competence, and relatedness. Positive associations were found between school connectedness (p fitness, as well as autonomy, competence, and relatedness. There were also significant relationships between affect, pro-social and antisocial behaviors, tried smoking, and mental fitness. A better understanding of adolescent health and its predictors is needed. By identifying core parameters for mental fitness, we can inform how to address students' needs through appropriate programs and policies supporting healthy school environments. © 2014, American School Health Association.

Evaluation of acoustic resonance at branch section in main steam line. Part 1. Effects of steam wetness on acoustic resonance

International Nuclear Information System (INIS)

Uchiyama, Yuta; Morita, Ryo

2011-01-01

The power uprating of the nuclear power plant (NPP) is conducted in United States, EU countries and so on, and also is planned in Japan. However, the degradation phenomena such as flow-induced vibration and wall thinning may increase or expose in the power uprate condition. In U.S. NPP, the dryer had been damaged by high cycle fatigue due to acoustic-induced vibration under a 17% extended power uprating (EPU) condition. This is caused by acoustic resonance at the stub pipes of safety relief valves (SRVs) in the main steam lines (MSL). Increased velocity by uprating excites the pressure fluctuations and makes large amplitude resonance. To evaluate the acoustic resonance at the stub pipes of SRVs in actual BWR, it is necessary to clarify the acoustic characteristics in steam flow. Although there are several previous studies about acoustic resonance, most of them are not steam flow but air flow. Therefore in this study, to investigate the acoustic characteristics in steam flow, we conducted steam flow experiments in each dry and wet steam conditions, and also nearly saturated condition. We measured pressure fluctuation at the top of the single stub pipe and in main steam piping. As a result, acoustic resonance in dry steam flow could be evaluated as same as that in air flow. It is clarified that resonance amplitude of fluctuating pressure at the top of the stub pipe in wet steam was reduced to one-tenth compared with that in dry. (author)

New Brunswick Market Design Committee : Congestion management issues

International Nuclear Information System (INIS)

2001-01-01

The restructuring of the New Brunswick wholesale power market comprises a number of issues that need to be resolved concerning transmission system related policy decisions and detailed design issues. The wholesale market structure, ownership structure, and means of preventing market power abuses all have an impact on the resolution of many of those issues. Some transmission related decisions regarding congestion management must be made, and they are examined in this document. The report includes a discussion of the issues related to congestion on the transmission system, a review of the decisions that remain to be made while proposing a number of alternatives, reviews decisions that other jurisdictions have made in somewhat similar circumstances. Finally, the advantages and disadvantages of each alternative are identified. Several high level transmission tariff design issues requiring to be addressed later in greater detail are listed in this document. 1 tab

Effectiveness of thermal ignition devices in lean hydrogen-air-steam mixtures

International Nuclear Information System (INIS)

Tamm, H.; McFarlane, R.; Liu, D.D.S.

1985-03-01

Deliberate ignition of hydrogen at low concentrations in reactor containment systems is one method of controlling hydrogen during degraded core accidents. Since many postulated accident conditions have substantial amounts of steam present, experiments have been performed to determine the hydrogen-air-steam concentration regimes in which ignitors would be effective. In these experiments, both a GM AC 7G thermal flow plug and a Tayco Model 3442 ignitor have been used. These ignitors have been installed in PWR containments with ice condensers and in BWR Mark III containments. This report presents the results of these ignitor effectiveness experiments, and gives the ignition limits and the effect of steam on the ignitor surface temperatures required for ignition

Wet steam turbines for CANDU-Reactors

International Nuclear Information System (INIS)

Westmacott, C.H.L.

1977-01-01

The technical characteristics of 4 wet steam turbine aggregates used in the Pickering nuclear power station are reported on along with operational experience. So far, the general experience was positive. Furthermore, plans are mentioned to use this type of turbines in other CANDU reactors. (UA) [de

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)

Characteristics of steam jet impingement on annulus

International Nuclear Information System (INIS)

Yoon, Sang H.; Kim, Won J.; Suh, Kune Y.; Song, Chul H.

2004-01-01

The steam jet impingement occurs when the steam through the cold leg from the steam generator strikes the inner reactor barrel during the reflood phase of a loss-of-coolant accident (LOCA), which is a characteristic behavior for the APR1400 (Advanced Power Reactor 1400 MWe). In the cold leg break LOCA, the steam and water flows in the downcomer are truly multidimensional. The azimuthal velocity distribution of the steam flow has an important bearing on the thermal hydraulic phenomena such as the emergency coolant water direct bypass, sweepout, steam condensation, and so forth. The investigation of jet flow is required to determine the steam path and momentum reduction rate after the impingement. For the observation of the steam behavior near the break, the computational fluid dynamic (CFD) analysis has been carried out using CFX5.6. The flow visualization and analysis demonstrate the velocity profiles of the steam flow in the annulus region for the same boundary conditions. Pursuant to the CFD results, the micro-Pitot tubes were positioned at varying angles, and corrected for their sensitivity. The experiments were carried out to directly measure the pressure differential and to visualize the flow utilizing a smoke injection method. Results from this study are slated to be applied to MARS, which is a thermal hydraulic system code for the best-estimate analysis. The current one- or two-dimensional analysis in MARS was known to distort the local flow behavior. To enhance prediction capability of MARS, it is necessary to inspect the steam path in the break flow and mechanically simulate the momentum variation. The present experimental and analytical results can locally be applied to developing the engineering models of specific and essential phenomena. (author)

SNR-steam generator design with respect to large sodium water reactions

International Nuclear Information System (INIS)

Jong, J.J. de; Kellner, A.; Florie, C.J.L.

1984-01-01

This paper deals with the experiences gained during the licensing procedure for the steam generators for the SNR 300 LMFBR regarding large sodium-water reactions. A description is given of the different calculations executed to investigate the effects of large leaks on the 85 MW helical coiled and straight tube steam generators. The investigations on the helical coiled steam generators are divided in the formulations of fluid behaviour, dynamic force calculations, dynamic response calculation and finally stress analyses. Several results are shown. The investigations on the straight tube steam generators are performed using models describing fluid-structure interaction, coupled with stress analyses. Several results are presented. A description is given of the problems and necessary construction changes during the licensing process. Advises are given for future analyses and design concepts for second generation commercial size LMFBR steam generators with respect to large leaks; based on the experience, gained with SNR 300, and using some new calculations for SNR 2. (author)

Natural gas for New Brunswick: First report

International Nuclear Information System (INIS)

1998-01-01

The development of the gas field off Sable Island and the imminent construction of a gas pipeline which will deliver natural gas to New Brunswick has prompted a thorough examination of energy-related issues in the province. This report presents the findings of the provincial energy committee which examined the implications of the arrival of natural gas to the province. The committee held a series of public hearings and consultations, and also received written submissions. After a historical perspective on natural gas as an energy source in the province and a review of the gas industry participants and their interests, the report discusses such issues as gas pipeline economics, local distribution company operations, infrastructure development, the regulatory framework, energy market competition, regional price equity, development of in-province gas sources, pipeline access, pipeline laterals and expansions, establishment of gas distribution franchises, municipal involvement in gas development, the impact of gas industry development on electric utility restructuring, and the environmental benefits of natural gas. Finally, recommendations are made regarding how natural gas should be regulated and distributed

Steam Digest 2002

Energy Technology Data Exchange (ETDEWEB)

2003-11-01

Steam Digest 2002 is a collection of articles published in the last year on steam system efficiency. DOE directly or indirectly facilitated the publication of the articles through it's BestPractices Steam effort. Steam Digest 2002 provides a variety of operational, design, marketing, and program and program assessment observations. Plant managers, engineers, and other plant operations personnel can refer to the information to improve industrial steam system management, efficiency, and performance.

Improvement of steam separator in boiling water reactors

Energy Technology Data Exchange (ETDEWEB)

Schaefer, Jan Peter; Cremer, Ingo; Lorenz, Maik [AREVA GmbH, Erlangen (Germany)

2013-07-01

The potential to improve the function of the steam separator is identified and explored by scaled air-water tests and validated CFD calculations. It can be outlined that requirements on a modern steam separator for BWR plants will be fulfilled, combined with very good operational experience of the existing separator designs (e.g. material, layout). With the new steam separator design, modern high performance fuel assembly designs can be used for various core loading strategies (e.g. low leakage). This allows an increased thermal power of up to +50 % for the fuel element clusters in the center of the core with high radial peaking factors. In addition, any problems with unallowable high moisture at the turbine are solved with the new design, which have been identified for running BWR plants with the old steam separator design after applying new core loading patterns (e.g. after power uprates). A compatible steam separator design for all running BWRs is ready to launch. (orig.)

The role of CFD combustion modelling in hydrogen safety management – VI: Validation for slow deflagration in homogeneous hydrogen-air-steam experiments

Energy Technology Data Exchange (ETDEWEB)

Cutrono Rakhimov, A., E-mail: cutrono@nrg.eu [Nuclear Research and Consultancy Group (NRG), Westerduinweg 3, 1755 ZG Petten (Netherlands); Visser, D.C., E-mail: visser@nrg.eu [Nuclear Research and Consultancy Group (NRG), Westerduinweg 3, 1755 ZG Petten (Netherlands); Holler, T., E-mail: tadej.holler@ijs.si [Jožef Stefan Institute (JSI), Jamova cesta 39, 1000 Ljubljana (Slovenia); Komen, E.M.J., E-mail: komen@nrg.eu [Nuclear Research and Consultancy Group (NRG), Westerduinweg 3, 1755 ZG Petten (Netherlands)

2017-01-15

Highlights: • Deflagration of hydrogen-air-steam homogeneous mixtures is modeled in a medium-scale containment. • Adaptive mesh refinement is applied on flame front positions. • Steam effect influence on combustion modeling capabilities is investigated. • Mean pressure rise is predicted with 18% under-prediction when steam is involved. • Peak pressure is evaluated with 5% accuracy when steam is involved. - Abstract: Large quantities of hydrogen can be generated during a severe accident in a water-cooled nuclear reactor. When released in the containment, the hydrogen can create a potential deflagration risk. The dynamic pressure loads resulting from hydrogen combustion can be detrimental to the structural integrity of the reactor. Therefore, accurate prediction of these pressure loads is an important safety issue. In previous papers, we validated a Computational Fluid Dynamics (CFD) based method to determine the pressure loads from a fast deflagration. The combustion model applied in the CFD method is based on the Turbulent Flame Speed Closure (TFC). In our last paper, we presented the extension of this combustion model, Extended Turbulent Flame Speed Closure (ETFC), and its validation against hydrogen deflagration experiments in the slow deflagration regime. During a severe accident, cooling water will enter the containment as steam. Therefore, the effect of steam on hydrogen deflagration is important to capture in a CFD model. The primary objectives of the present paper are to further validate the TFC and ETFC combustion models, and investigate their capability to predict the effect of steam. The peak pressures, the trends of the flame velocity, and the pressure rise with an increase in the initial steam dilution are captured reasonably well by both combustion models. In addition, the ETFC model appeared to be more robust to mesh resolution changes. The mean pressure rise is evaluated with 18% under-prediction and the peak pressure is evaluated with 5

Steam-Generator Integrity Program/Steam-Generator Group Project

International Nuclear Information System (INIS)

1982-10-01

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

Steam drums

International Nuclear Information System (INIS)

Crowder, R.

1978-01-01

Steam drums are described that are suitable for use in steam generating heavy water reactor power stations. They receive a steam/water mixture via riser headers from the reactor core and provide by means of separators and driers steam with typically 0.5% moisture content for driving turbines. The drums are constructed as prestressed concrete pressure vessels in which the failure of one or a few of the prestressing elements does not significantly affect the overall strength of the structure. The concrete also acts as a radiation shield. (U.K.)

Conceptual design study of Cu bonded steam generator

International Nuclear Information System (INIS)

Chikazawa, Yoshitaka; Konomura, Mamoru

2004-05-01

In phase II of feasibility study of commercialized fast reactor cycle systems of JNC, we make a concept of a sodium cooled reactor without secondary sodium circuits. And a sodium cooled reactor with Cu bonded steam generators is one of promising concept. As the result of FY 2001 study, the construction cost of reactor cooling system with rectangular tube Cu bonded steam generators is 0.71 to 1.23 times as much as that of an ordinary sodium cooled reactor with secondary sodium circuits. In the FY 2003 study, plastic and creep analysis to evaluate life distortion are carried out and inelastic strains and creep fatigue damage are checked for full code compliance. The NNC's crack growth experiments show that there are few possibility to penetrate a crack from the steam tube side to the sodium tube side at the operating temperature. But penetration is observed in a four point bend test at the room temperature, because the notch opens widely in the bend test. In the FY 2004 study, a gas pressurized crack growth experiment is planed to confirm that there is no crack penetration in the condition of operating steam generators. (author)

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

Gas Flow Validation with Panda Tests from the OECD SETH Benchmark Covering Steam/Air and Steam/Helium/Air Mixtures

International Nuclear Information System (INIS)

Royl, P.; Travis, J.R.; Breitung, W.; Kim, J.; Kim, S.B.

2009-01-01

The CFD code GASFLOW solves the time-dependent compressible Navier-Stokes Equations with multiple gas species. GASFLOW was developed for nonnuclear and nuclear applications. The major nuclear applications of GASFLOW are 3D analyses of steam/hydrogen distributions in complex PWR containment buildings to simulate scenarios of beyond design basis accidents. Validation of GASFLOW has been a continuously ongoing process together with the development of this code. This contribution reports the results from the open posttest GASFLOW calculations that have been performed for new experiments from the OECD SETH Benchmark. Discussed are the steam distribution tests 9 and 9 bis, 21 and 21 bis involving comparable sequences with and without steam condensation and the last SETH test 25 with steam/helium release and condensation. The latter one involves lighter gas mixture sources like they can result in real accidents. The helium is taken as simulant for hydrogen

Volume reduction of dry active waste by use of a waste sorting table at the Brunswick nuclear power plant

International Nuclear Information System (INIS)

Snead, P.B.

1988-01-01

Carolina Power and Light Company's Brunswick nuclear power plant has been using a National Nuclear Corporation Model WST-18 Waste Sorting Table to monitor and sort dry active waste for segregating uncontaminated material as a means of low-level waste volume reduction. The WST-18 features 18 large-area, solid scintillation detectors arranged in a 3 x 6 array underneath a sorting/monitoring surface that is shielded from background radiation. An 11-week study at Brunswick showed that the use of the waste sorting table resulted in dramatic improvements in both productivity (man-hours expended per cubic foot of waste processed) and monitoring quality over the previous hand-probe frisking method. Use of the sorting table since the study has confirmed its effectiveness in volume reduction. The waste sorting table paid for its operation in volume reduction savings alone, without accounting for the additional savings from recovering reusable items

Steam explosion triggering and efficiency studies

International Nuclear Information System (INIS)

Buxton, L.D.; Nelson, L.S.; Benedick, W.B.

1979-01-01

Laboratory experiments on the thermal interaction of simulated light water reactor (LWR) fuel melts and water are summarized. Their purpose was to investigate the possibility of steam explosions occurring for a range of hypothetical accident conditions. Pressure, temperature, hot liquid motion and cold liquid motion were monitored during the experiments

New Brunswick Power Corporation: Business plan, 1994-1999

International Nuclear Information System (INIS)

1994-02-01

The first publicly available business plan for the New Brunswick Power Corp. is presented. The five-year plan provides an overview of the Corporation's performance and directions, including possible future rate increases. A review of the corporate history of the utility is followed by a description of the strategic framework under which the Corporation conducts its business operations. The information presented includes customer requirements, power generation and transmission, system operations, personnel management, environmental protection, and external factors affecting operations. This overview demonstrates the complex issues facing the utility, the choices made in the past, and the matters that will have to be faced in the future. The business overview is followed by focused business plans in six key functional areas (facilities, operations, personnel, technology, environment, and finances) and comprehensive financial forecasts that outline a commitment to maintain competitive rates for customers. The reasoning behind the forecasts is explained and a glossary is included. 21 figs., 16 tabs

[Steam and air co-injection in removing TCE in 2D-sand box].

Science.gov (United States)

Wang, Ning; Peng, Sheng; Chen, Jia-Jun

2014-07-01

Steam and air co-injection is a newly developed and promising soil remediation technique for non-aqueous phase liquids (NAPLs) in vadose zone. In this study, in order to investigate the mechanism of the remediation process, trichloroethylene (TCE) removal using steam and air co-injection was carried out in a 2-dimensional sandbox with different layered sand structures. The results showed that co-injection perfectly improved the "tailing" effect compared to soil vapor extraction (SVE), and the remediation process of steam and air co-injection could be divided into SVE stage, steam strengthening stage and heat penetration stage. Removal ratio of the experiment with scattered contaminant area was higher and removal speed was faster. The removal ratios from the two experiments were 93.5% and 88.2%, and the removal periods were 83.9 min and 90.6 min, respectively. Steam strengthened the heat penetration stage. The temperature transition region was wider in the scattered NAPLs distribution experiment, which reduced the accumulation of TCE. Slight downward movement of TCE was observed in the experiment with TCE initially distributed in a fine sand zone. And such downward movement of TCE reduced the TCE removal ratio.

Point Lepreau nuclear project unlocks power future of New Brunswick

Energy Technology Data Exchange (ETDEWEB)

Burgess, E J [New Brunswick Electric Power Commission, Fredericton (Canada)

1976-08-01

Projects under development will increase the generating capacity of New Brunswick Electric Power Commission from 1.47 to 3.14 GW. Of these projects, the most significant is the 0.63 GW Point Lepreau CANDU reactor. Progress in the construction of the power reactor is summarized. The concrete reactor building was slipformed in April 1976. A second nuclear unit at Point Lepreau is being considered. Information is also provided on the new oil-fired station at Coleson Cove, a new oil- or coal-fired unit at Dalhousie and further hydro units at Mactaquac. A 0.16 GW pumped storage hydro unit at Green River is being considered. Information on transmission (Including the HVDC system), substations, and connection with Hydro Quebec is included.

An Improved Steam Injection Model with the Consideration of Steam Override

Directory of Open Access Journals (Sweden)

He Congge

2017-01-01

Full Text Available The great difference in density between steam and liquid during wet steam injection always results in steam override, that is, steam gathers on the top of the pay zone. In this article, the equation for steam override coefficient was firstly established based on van Lookeren’s steam override theory and then radius of steam zone and hot fluid zone were derived according to a more realistic temperature distribution and an energy balance in the pay zone. On this basis, the equation for the reservoir heat efficiency with the consideration of steam override was developed. Next, predicted results of the new model were compared with these of another analytical model and CMG STARS (a mature commercial reservoir numerical simulator to verify the accuracy of the new mathematical model. Finally, based on the validated model, we analyzed the effects of injection rate, steam quality and reservoir thickness on the reservoir heat efficiency. The results show that the new model can be simplified to the classic model (Marx-Langenheim model under the condition of the steam override being not taken into account, which means the Marx-Langenheim model is corresponding to a special case of this new model. The new model is much closer to the actual situation compared to the Marx-Langenheim model because of considering steam override. Moreover, with the help of the new model, it is found that the reservoir heat efficiency is not much affected by injection rate and steam quality but significantly influenced by reservoir thickness, and to ensure that the reservoir can be heated effectively, the reservoir thickness should not be too small.

Function analysis of steam isolation valves

International Nuclear Information System (INIS)

Persson, R.; Sjoeberg, A.; Aakesson, H.; Kilpi, K.; Noro, H.; Siikonen, T.; Wallen, G.

1981-01-01

Function analysis of system-medium-operated steam isolation valves has been the objective of the Swedish-Finnish IVLS project, the results of which are presented in this report. Theoretical models were to be verified against available experimental data, to some extent from the HDR blowdown experiments, which are part of a German reactor safety program. Finnish hydraulic measurements on a valve model (scale 1:2.15) have been performed to give complementary data. The analysis work has covered the thermal-hydraulic behaviour of steam isolation valves as well as phenomena related to structural mechanics. (Auth.)

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

Role of the statistician in the decommissioning of the New Brunswick Laboratory and other nuclear facilities

International Nuclear Information System (INIS)

Gilbert, R.O.

1980-12-01

This report examines what the statistician can contribute to decommissioning operations, with particular emphasis on the New Brunswick Laboratory (NBL) currently scheduled for decommissioning beginning in FY81. In the opinion of the author, a professional statistician should be a full member of the planning team directing decommissioning operations at the New Brunswick Laboratory. This opinion is based in part on the familiarity with the valuable contributions made by statisticians toward the cleanup of transuranics in soil on the Enewetak Atoll. More generally, however, the professional statistician can help plan the decommissioning effort to help ensure that representative data are obtained, analyzed and, interpreted in appropriate ways so that RA decisions can be made with the required confidence. The statistician's contributions at the NBL could include providing guidance on the number and location of samples and in-situ measurements, analyzing and interpreting these data, designing a data management and documentation system, interfacing with the certification contractor's statistician, and assisting in writing documentation and final reports. In all cases, the statistician should work closely with the professional health physicist and others on the planning team in a closely coordinated effort of planning and data analysis

Analysis of experimental characteristics of multistage steam-jet electors of steam turbines

Science.gov (United States)

Aronson, K. E.; Ryabchikov, A. Yu.; Brodov, Yu. M.; Brezgin, D. V.; Zhelonkin, N. V.; Murmanskii, I. B.

2017-02-01

A series of questions for specification of physical gas dynamics model in flow range of steam-jet unit and ejector computation methodology, as well as functioning peculiarities of intercoolers, was formulated based on analysis of experimental characteristics of multistage team-jet steam turbines. It was established that coefficient defining position of critical cross-section of injected flow depends on characteristics of the "sound tube" zone. Speed of injected flow within this tube may exceed that of sound, and pressure jumps in work-steam decrease at the same time. Characteristics of the "sound tube" define optimal axial sizes of the ejector. According to measurement results, the part of steam condensing in the first-stage coolant constitutes 70-80% of steam amount supplied into coolant and is almost independent of air content in steam. Coolant efficiency depends on steam pressure defined by operation of steam-jet unit of ejector of the next stage after coolant of steam-jet stage, temperature, and condensing water flow. As a rule, steam entering content of steam-air mixture supplied to coolant is overheated with respect to saturation temperature of steam in the mixture. This should be taken into account during coolant computation. Long-term operation causes changes in roughness of walls of the ejector's mixing chamber. The influence of change of wall roughness on ejector characteristic is similar to the influence of reverse pressure of the steam-jet stage. Until some roughness value, injection coefficient of the ejector stage operating in superlimiting regime hardly changed. After reaching critical roughness, the ejector switches to prelimiting operating regime.

THTR steam generator licensing experience as seen by the manufacturer

International Nuclear Information System (INIS)

Fricker, H.W.

1981-01-01

This paper describes the licensing procedures of the manufacture of the 300 MWe THTR steam generator. The following problems are discussed: operating data, design, materials used, manufacture and installation of the generator, and also quality control

Strategic elements of steam cycle chemistry control practices at TXU's Comanche Peak steam electric station

International Nuclear Information System (INIS)

Fellers, B.; Stevens, J.; Nichols, G.

2002-01-01

Early industry experience defined the critical importance of Chemistry Control Practices to maintaining long-term performance of PWR steam generators. These lessons provided the impetus for a number of innovations and alternate practices at Comanche Peak. For example, advanced amine investigations and implementation of results provided record low iron transport and deposition. The benefits of the surface-active properties of dimethyl-amine exceeded initial expectations. Operation of pre-coat polishers and steam generator blowdown demineralizers in the amine cycle enabled optimization of amine concentrations and stable pH control. The strategy for coordinated control of oxygen and hydrazine dosing complemented the advanced amine program for protective oxide stabilization. Additionally, a proactive chemical cleaning was performed on Unit 1 to prevent degradations from general fouling of steam generator tube-tube support plate (TSP) and top-of-tubesheet (TTS) crevices. This paper shares the results of these innovations and practices. Also, the bases, theory, and philosophy supporting the strategic elements of program will be presented. (authors)

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

Corrosion fatigue in LP steam turbine blading - experiences, causes and appropriate measures; Korrosionsutmattning i aangturbinskovlar - Erfarenheter, inverkande faktorer och moejliga aatgaerder

Energy Technology Data Exchange (ETDEWEB)

Tavast, J [ABB STAL AB, Finspaang (Sweden)

1996-12-01

Corrosion fatigue in LP steam turbine blading was reviewed together with result of tests performed in order to find blade materials with improved resistance against this. According to international experience, corrosion fatigue of 12Cr steam turbine blades in the transition zone between dry and wet steam, is one of the major causes, if not the major cause, for unavailability of steam turbines. Corrosion fatigue in LP blading is a frequent problem also in Swedish and Finnish nuclear power plants, especially in turbines of type D54 in BWR-plants. Corrosion fatigue has also been discovered in at least one type of nuclear turbine. Initiation times have been very long and the varying experiences in different types of turbines may simply reflect differing initiation times. Corrosion fatigue may therefore become more frequent in other types of turbines in the future. The type of water treatment (BWR/PWR) and possibly temperature after reheating seem to influence the risk for corrosion fatigue. Influence of inleakage of cooling water is less clear for these nuclear plants. The long initiation times together with the fact that very few of the cracked blades have actually failed, indicate that the cracks initiate and/or propagate during transients. Extensive laboratory tests show that there are alternative blade materials available with improved resistance against corrosion fatigue, with the most promising being 15/5 PH and A905, together with Ti6Al4V. The Ti alloy shows the best resistance against corrosion fatigue in most environments and is already used in some turbines. Disadvantage is a higher cost and possible need for redesign of the blades. The alternative materials are recommended for use for blades in the transition zone between dry and wet steam in LP turbines. The main disadvantage is a lack of references, even if 15%5 PH has been used to a very limited extent. 40 refs, 24 figs, 12 tabs, 9 appendices

New ferritic steels for advanced steam plants

Energy Technology Data Exchange (ETDEWEB)

Mayer, K H; Koenig, H [GEC ALSTHOM Energie GmbH, Nuremberg (Germany)

1999-12-31

During the last 15-20 years ferritic-martensitic 9-12 % chromium steels have been developed under international research programmes which permit inlet steam temperatures up to approx. 625 deg C and pressures up to about 300 bars, thus leading to improvements in thermal efficiency of around 8 % and a CO{sub 2} reduction of about 20 % versus conventional steam parameters. These new steels are already being applied in 13 European and 34 Japanese power stations with inlet steam temperature up to 610 deg C. This presentation will give an account of the content, scope and results of the research programmes and of the experience gained during the production of components which have been manufactured from the new steels. (orig.) 13 refs.

New ferritic steels for advanced steam plants

Energy Technology Data Exchange (ETDEWEB)

Mayer, K.H; Koenig, H. [GEC ALSTHOM Energie GmbH, Nuremberg (Germany)

1998-12-31

During the last 15-20 years ferritic-martensitic 9-12 % chromium steels have been developed under international research programmes which permit inlet steam temperatures up to approx. 625 deg C and pressures up to about 300 bars, thus leading to improvements in thermal efficiency of around 8 % and a CO{sub 2} reduction of about 20 % versus conventional steam parameters. These new steels are already being applied in 13 European and 34 Japanese power stations with inlet steam temperature up to 610 deg C. This presentation will give an account of the content, scope and results of the research programmes and of the experience gained during the production of components which have been manufactured from the new steels. (orig.) 13 refs.

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

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)

Ontario Hydro's operating experience with steam generators with specifics on Bruce A and Bruce B problems

Energy Technology Data Exchange (ETDEWEB)

Eatock, J W; Patterson, R W [Ontario Hydro, Toronto, ON (Canada); Dyck, R W [Ontario Hydro, Central Production Services Division, Toronto, ON (Canada)

1991-04-01

The performance of the steam generators in Ontario Hydro nuclear power stations is reviewed. This performance has generally been outstanding compared to world averages, with very low tube failure and plugging rates. Steam generator problems have made only minor contributions to Ontario Hydro nuclear station incapability factors. The mechanisms responsible for the the observed tube degradation and failures are described. The majority of the leaks have been due fatigue in the U-bend of the Bruce 'A' steam generators. There have been very few failures attributed to corrosion of the three tube materials used in Ontario Hydro steam generators. Recent performance has been deteriorating primarily due to deposit accumulation in the steam generators. Plugging of the broached holes in the upper support plates at Bruce 'A' has caused some derating of two units. Increases have been observed in the primary heat transport system reactor inlet temperature of several units. These increases may be attributed to steam generator tube surface fouling. In addition, several units have accumulated deep, hard sludge piles on the tube sheet, although little damage been observed. Recently some fretting of tubes has been observed at BNGSB in the U-bend support region. Remedial measures are being taken to address the current problems. Solutions are being evaluated to reduce the generation of corrosion products in the feedtrain and their subsequent transport to the steam generators. (author)

Ontario Hydro's operating experience with steam generators with specifics on Bruce A and Bruce B problems

International Nuclear Information System (INIS)

Eatock, J.W.; Patterson, R.W.; Dyck, R.W.

1991-01-01

The performance of the steam generators in Ontario Hydro nuclear power stations is reviewed. This performance has generally been outstanding compared to world averages, with very low tube failure and plugging rates. Steam generator problems have made only minor contributions to Ontario Hydro nuclear station incapability factors. The mechanisms responsible for the the observed tube degradation and failures are described. The majority of the leaks have been due fatigue in the U-bend of the Bruce 'A' steam generators. There have been very few failures attributed to corrosion of the three tube materials used in Ontario Hydro steam generators. Recent performance has been deteriorating primarily due to deposit accumulation in the steam generators. Plugging of the broached holes in the upper support plates at Bruce 'A' has caused some derating of two units. Increases have been observed in the primary heat transport system reactor inlet temperature of several units. These increases may be attributed to steam generator tube surface fouling. In addition, several units have accumulated deep, hard sludge piles on the tube sheet, although little damage been observed. Recently some fretting of tubes has been observed at BNGSB in the U-bend support region. Remedial measures are being taken to address the current problems. Solutions are being evaluated to reduce the generation of corrosion products in the feedtrain and their subsequent transport to the steam generators. (author)

Droplet solidification and the potential for steam explosions

International Nuclear Information System (INIS)

Epstein, M.; Fauske, H.K.; Luangdilok, W.

2009-01-01

It is well known that under certain circumstances a mixture of coarse-hot (molten) drops in water formed from pouring a hot melt into water explodes. This so-called 'steam explosion' is generally believed to involve steam-bubble-collapse-induced fine fragmentation of the melt drops and concomitant water vaporization on a timescale that is short compared with the steam pressure relief time. Motivated by the idea put forth by Okkonen and Sehgal that rapid solidification would render UO 2 -containing (Corium) melt drops stiff and resistant to the steam-bubble-collapse-induced fragmentation required to support an explosion, here we combine solidification theory with an available theory of the stability of thin, submerged crusts subject to acceleration to predict the 'cutoff time' beyond which melt-drop fragmentation is suppressed by crust cover rigidity. Illustration calculations show that the cutoff time for Corium melt drops in water is a fraction of a second and probably shorter than the time it takes to form the explosion-prerequisite-coarse-premixture configuration of melt drops in water, while the opposite is true for the molten aluminum oxide/water system for which the window of opportunity for an explosion is predicted to be several seconds. These theoretical findings are consistent with early experiments that revealed molten uranium oxide or Corium pours into water to be non-explosive and that produced steam explosions upon pouring molten aluminum oxide into water. Also in this paper, the recent TROI Corium/water interaction experiments are examined and it is concluded that they do not contravene the earlier experimental observations that the pouring of prototypical Corium mixtures into water does not result in steam explosions with destructive potential. (author)

Five-component propagation model for steam explosion analysis

International Nuclear Information System (INIS)

Yang, Y.; Moriyama, Kiyofumi; Park, H.S.; Maruyama, Yu; Sugimoto, Jun

1999-01-01

A five-field simulation code JASMINE-pro has been developed at JAERI for the calculation of the propagation and explosion phase of steam explosions. The basic equations and the constitutive relationships specifically utilized in the propagation models in the code are introduced in this paper. Some calculations simulating the KROTOS 1D and 2D steam explosion experiments are also stated in the paper to show the present capability of the code. (author)

LABORATORY SCALE STEAM INJECTION TREATABILITY STUDIES

Science.gov (United States)

Laboratory scale steam injection treatability studies were first developed at The University of California-Berkeley. A comparable testing facility has been developed at USEPA's Robert S. Kerr Environmental Research Center. Experience has already shown that many volatile organic...

Targeted steam injection using horizontal wells with limited entry perforations

Energy Technology Data Exchange (ETDEWEB)

Boone, T. J.; Youck, D. G.; Sun, S. [Imperial Oil Resources, Calgary, AB (Canada)

1998-12-31

An experimental horizontal well using limited-entry perforations as a method for distributing steam to different zones was used to replace ten vertical injection wells. The well was located between rows of vertical wells in a reservoir that has been subjected to more than ten years of operation under cyclic steam stimulation. The limited-entry perforations enabled steam to be targeted at the cold regions of the reservoir. This paper presents an assessment of the well based on theoretical calculations, measured injection pressures and rates and 3-D seismic imaging. All the data collected during the experiment support the conclusion that effective steam distribution along the well has been achieved. It was also concluded that this technology has significant potential for SAGD applications as a mechanism for achieving improved steam distribution at a much reduced cost. 5 refs., 8 figs.

Experience in adjusting of the level regulation system of steam generators of the Rovno NPP

International Nuclear Information System (INIS)

Patselyuk, S.N.; Sokolov, A.G.; Kazakov, V.I.; Dorosh, Yu.A.

1984-01-01

A system of feed water level control in steam generators at the Rovno NPP with WWER-440 reactors which comprises start-up as well as main regulators is described. The start-up regulator (single-pulsed with a signal by the level) keeps the level in the steam generator at loadings up to 30% of the nominal reactor power Nsub(nom.) The main regulator is connected in the three-pulsed circuit and it receives signals by steam and water flow rate and by the level in the steam generator. The main regulator has been started only at loadings above 40% Nsub(nom.). After reconstruction it was used in the 15-100% Nsub(nom.) range. Characteristics of the level control system in the steam generator at perturbations intoduced by the main circulating pump (MCP) and turbine disconnection as well as change in feed water flow rate have been studied. The studies have revealed that the system ensures necessary quality of control in stationary modes. The system operates stably at perturbations of feed water flow rate up to 50% Nsub(nom.). Perturbations by MCP connections and disconnections is most difficult for control system

Avoiding steam-bubble-collapse-induced water hammers in piping systems

International Nuclear Information System (INIS)

Chou, Y.; Griffith, P.

1989-10-01

In terms of the frequency of occurrence, steam bubble collapse in subcooled water is the dominant initiating mechanism for water hammer events in nuclear power plants. Water hammer due to steam bubble collapse occurs when water slug forms in stratified horizontal flow, or when steam bubble is trapped at the end of the pipe. These types of water hammer events have been studied experimentally and analytically in order to develop stability maps showing those combinations of filling velocities and liquid subcooling that cause water hammer and those which don't. In developing the stability maps, experiments with different piping orientations were performed in a low pressure laboratory apparatus. Details of these experiments are described, including piping arrangement, test procedures, and test results. Visual tests using a transparent Lexan pipe are also performed to study the flow regimes accompanying the water hammer events. All analytical models were tested by comparison with the corresponding experimental results. Based on these models, and step-by-step approach for each flow geometry is presented for plant designers and engineers to follow in avoiding water hammer induced by steam bubble collapse when admitting cold water into pipes filled with steam. 37 refs., 54 figs., 2 tabs

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)

Field experience of cleanliness inspection for secondary-side in PWR steam generators

International Nuclear Information System (INIS)

Ding Xunshen

1997-05-01

The mechanical cleaning and TV inspection technology for secondary-side in steam generators of Daya Bay Nuclear Power Plant has been used for preventing the heat transfer tubes from damage caused by residues in steam generator and foreign objects. A lancing has been used for steam generators. The high-pressure jet sent from the central lane to the inside of the bundle has two objectives: (a) the mechanical energy in the jet breaks up the deposit, and (b) the particles are then carried to the periphery of the generator, where they are collected by a water circulation system. The TV inspection consists of the inspection before lancing and after lancing. The former includes the inspection of outer tube lane and central tube lane, and the extraction of foreign objects; the latter includes the inspection of inter tube area, outer tube lane and central tube lane, and the extraction of foreign objects. Video cassette are visualized by a specialist who is qualified to realize a map representative of the cleanliness state of the tubesheet, and to judge if it is or not acceptable. The TV inspection obtained a cleanliness conclusion according to acceptance criteria. So, it is important to work out cleanliness acceptance criteria suit for every operation stages. The site practices shown that after lancing, although hard deposit existed and height of hard deposit in some local place exceeded 5 mm, but the sludge remaining on tubesheet was less. If we can conduct periodic lancing to steam generator, the increase rate of hard deposit will be small. (7 refs., 12 figs.)

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

Some possible causes and probability of leakages in LMFBR steam generators

International Nuclear Information System (INIS)

Bolt, P.R.

1984-01-01

Relevant operational experience with steam generators for process and conventional plant and thermal and fast reactors is reviewed. Possible causes of water/steam leakages into sodium/gas are identified and data is given on the conditions necessary for failure, leakage probability and type of leakage path. (author)

Diverse tetrapod trackways in the Lower Pennsylvanian Tynemouth Creek Formation, near St. Martins, southern New Brunswick, Canada

DEFF Research Database (Denmark)

Falcon-Lang, Howard J.; Gibling, Martin R.; Benton, Michael J.

2010-01-01

Newly discovered tetrapod trackways are reported from eight sites in the Lower Pennsylvanian Tynemouth Creek Formation of southern New Brunswick, Canada. By far the most abundant and well-preserved tracks comprise pentadactyl footprints of medium size (32–53 mm long) with slender digits and a nar......Newly discovered tetrapod trackways are reported from eight sites in the Lower Pennsylvanian Tynemouth Creek Formation of southern New Brunswick, Canada. By far the most abundant and well-preserved tracks comprise pentadactyl footprints of medium size (32–53 mm long) with slender digits...... and a narrow splay (mostly b55°). Digit lengths typically approximate a phalangeal formula of 23453 (manus) and 23454 (pes), but this may vary due to extramorphology. These tracks are referred to Pseudobradypus and they are attributed to early amniotes. A second type of track (rare) comprises very small (5......–8 mm long) tetradactyl manus, and incompletely preserved pedes. Referred to Batrachichnus, these are attributed to temnospondyl amphibians. A third type (also rare) comprises small pentadactyl pedes (20–25 mm long) showing stubby, widely splayed (152°) digits with a terminal bulge. Manus are probably...

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

Worldwide assessment of steam-generator problems in pressurized-water-reactor nuclear power plants

International Nuclear Information System (INIS)

Woo, H.H.; Lu, S.C.

1981-01-01

Objective is to assess the reliability of steam generators of pressurized water reactor (PWR) power plants in the United States and abroad. The assessment is based on operation experience of both domestic and foreign PWR plants. The approach taken is to collect and review papers and reports available from the literature as well as information obtained by contacting research institutes both here and abroad. This report presents the results of the assessment. It contains a general background of PWR plant operations, plant types, and materials used in PWR plants. A review of the worldwide distribution of PWR plants is also given. The report describes in detail the degradation problems discovered in PWR steam generators: their causes, their impacts on the performance of steam generators, and the actions to mitigate and avoid them. One chapter is devoted to operating experience of PWR steam generators in foreign countries. Another discusses the improvements in future steam generator design

High pressure oxidation of sponge-Zr in steam/hydrogen mixtures

International Nuclear Information System (INIS)

Kim, Y.S.

1997-01-01

A thermogravimetric apparatus for operation in 1 and 70 atm steam-hydrogen or steam-helium mixtures was used to investigate the oxidation kinetics of sponge-Zr containing 215 ppm Fe. Weight-gain rates, reflecting both oxygen and hydrogen uptake, were measured in the temperature range 350-400 C. The specimens consisted of thin sponge-Zr layers metallurgically bonded to a Zircaloy disk. The edges of the disk specimens were coated with a thin layer of pure gold to avoid the deleterious effect of corners. Following each experiment, the specimens were examined metallographically to reveal the morphology of the oxide and/or hydride formed. Two types of oxide, one black and uniform and the other white and nodular, were observed on sponge-Zr surfaces oxidized in steam environments at 70 atm. The oxidation rate when white-nodular oxide formed was a factor of two higher than that of black-uniform oxide at 400 C for steam contents above 1 mol%. The oxidation rate was independent of total pressure, the carrier gas (H 2 or He) and steam content above ∝1 mol%. The oxidation kinetics of sponge-Zr follows a linear law for maximum reaction times up to ∝6 days. The oxidation rate in steam-hydrogen mixtures at 70 atm total pressure decreases when the steam content approaches the steam-starved region (∝0.5 mol% steam at 400 C and ∝0.02 mol% steam at 350 C). Lower steam concentrations cause massive hydriding of the specimens. Even at steam concentrations above the critical value, direct hydrogen absorption from the gas was manifest by hydrogen pickup fractions greater than unity. (orig.)

Steam--water mixing in nuclear reactor safety loss-of-coolant experiments

International Nuclear Information System (INIS)

Naff, S.A.; Schwarz, W.F.

1978-01-01

Computer models used to predict the response of reactors to hypothesized accidents necessarily incorporate approximating assumptions. To verify the models by comparing predicted and measured responses in test facilities, these assumptions must be confirmed to be realistic. Recent experiments in facilities capable of repeatedly duplicating the transient behavior of a pressurized water reactor undergoing a pipe rupture show that the assumption of complete water-steam mixing during the transient results in the predicted decompression being faster than that observed. Water reactor safety studies currently in progress include programs aimed at the verification of computer models or ''codes'' used to predict reactor system responses to various hypothesized accidents. The approach is to compare code predictions of transients with the actual test transients in experimental facilities. The purpose of this paper is to explain an important instance in which predictions and data are not in complete agreement and to indicate the significance to water reactor safety studies

The effect of steam separataor efficiency on transient following a steam line break

International Nuclear Information System (INIS)

Choi, J.H.; Ohn, M.Y.; Lee, N.H.; Hwang, S.T.; Lee, S.K.

1996-01-01

Detailed thermalhydraulic simulations for CANDU 6 steam line break inside containment are performed to predict the response of the primary and secondary circuits. The analysis is performed using the thermalhydraulic computer code, CATHENA, with a coupled primary and secondary circuit model. A two-loop representation of the primary and secondary circuits is modelled. The secondary circuit model includes the feedwater line from the deaerator storage tank, multi-node steam generators and the steam line up to the turbine. Two cases were carried out using different assumptions for the efficiency of the steam separators. Case 1 assumes the efficiency of the steam separators becomes zero when the water level in the steam drum increases to the elevation of primary cyclones, or the outlet flow from the steam generator becomes higher than 150 % of normal flow. Case 2 assumes the efficiency becomes zero only when the water level in the steam drum reaches the elevation of primary cyclones. The simulation results show that system responses are sensitive to the assumption for the efficiency of the steam separators and case 1 gives higher discharge energy. Fuel cooling is assured, since primary circuit is cooled down sufficiently by the steam generators for both cases. (author)

Steam separator-superheater with drawing of a fraction of the dried steam

International Nuclear Information System (INIS)

Bessouat, Roger; Marjollet, Jacques.

1976-01-01

This invention concerns a vertical separator-superheater of the steam from a high pressure expansion turbine before it is admitted to an expansion turbine at a lower pressure, by heat exchange with steam under a greater pressure, and drawing of a fraction of the dried steam before it is superheated. Such drawing off is necessary in the heat exchange systems of light water nuclear reactors. Its purpose is to provide a separator-superheater that provides an even flow of non superheated steam and a regular distribution of the steam to be superheated to the various superheating bundles, with a significantly uniform temperature of the casing, thereby preventing thermal stresses and ensuring a minimal pressure drop. The vertical separator-superheater of the invention is divided into several vertical sections comprising as from the central area, a separation area of the steam entrained water and a superheater area and at least one other vertical section with only a separation area of the steam entrained water [fr

Simulation studies of steam-propane injection for the Hamaca heavy oil field

Energy Technology Data Exchange (ETDEWEB)

Venturini, G.J.; Mamora, D.D. [Texas A and M Univ., Austin, TX (United States)

2003-07-01

Laboratory experiments have been conducted at Texas A and M University to examine the use of steam additives such as propane, methane and nitrogen to improve the production of heavy oils and increase steam recovery efficiency. In particular, the use of steam-propane injection for heavy Hamaca crude oil with API gravity of 9.3 and viscosity of 25,000 cp at 50 degrees C was examined. Experimental runs involved the injection of steam or propane into injection cells at a constant rate, temperature and cell outlet pressure. The experimental results suggest that the use of steam-propane injection may translate to reduction of fuel costs for field injections. Initially, propane-steam injection resulted in a two-month oil production acceleration compared to pure steam injection. A significant gain in discounted revenue and savings in steam injection costs could be realized. The study also showed the oil product rate peak with steam-propane injection was much higher than that with pure steam injection. The oil production acceleration increases with increasing propane content. Oil recovery at the end of a five-year forecast period increases by 6.7 per cent of original oil in place (OOIP) compared to 2.3 per cent OOIP with pure steam injection. 12 refs., 6 tabs., 28 figs.

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

Methane Steam Reforming Kinetics for a Rhodium-Based Catalyst

DEFF Research Database (Denmark)

Jakobsen, Jon Geest; Jakobsen, M.; Chorkendorff, Ib

2010-01-01

Methane steam reforming is the key reaction to produce synthesis gas and hydrogen at the industrial scale. Here the kinetics of methane steam reforming over a rhodium-based catalyst is investigated in the temperature range 500-800 A degrees C and as a function of CH4, H2O and H-2 partial pressures....... The methane steam reforming reaction cannot be modeled without taking CO and H coverages into account. This is especially important at low temperatures and higher partial pressures of CO and H-2. For methane CO2 reforming experiments, it is also necessary to consider the repulsive interaction of CO...

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

Steam turbine generators for Sizewell 'B' nuclear power station

International Nuclear Information System (INIS)

Hesketh, J.A.; Muscroft, J.

1990-01-01

The thermodynamic cycle of the modern 3000 r/min steam turbine as applied at Sizewell 'B' is presented. Review is made of the factors affecting thermal efficiency including the special nature of the wet steam cycle and the use of moisture separation and steam reheating. Consideration is given to the optimization of the machine and cycle parameters, including particular attention to reheating and to the provision of feedheating, in order to achieve a high overall level of performance. A modular design approach has made available a family of machines suitable for the output range 600-1300 MW. The constructional features of the 630 MW Sizewell 'B' turbine generators from this range are described in detail. The importance of service experience with wet steam turbines and its influence on the design of modern turbines for pressurized water reactor (PWR) applications is discussed. (author)

Small Engines as Bottoming Cycle Steam Expanders for Internal Combustion Engines

Directory of Open Access Journals (Sweden)

Rohitha Weerasinghe

2017-01-01

Full Text Available Heat recovery bottoming cycles for internal combustion engines have opened new avenues for research into small steam expanders (Stobart and Weerasinghe, 2006. Dependable data for small steam expanders will allow us to predict their suitability as bottoming cycle engines and the fuel economy achieved by using them as bottoming cycles. Present paper is based on results of experiments carried out on small scale Wankel and two-stroke reciprocating engines as air expanders and as steam expanders. A test facility developed at Sussex used for measurements is comprised of a torque, power and speed measurements, electronic actuation of valves, synchronized data acquisition of pressure, and temperatures of steam and inside of the engines for steam and internal combustion cycles. Results are presented for four engine modes, namely, reciprocating engine in uniflow steam expansion mode and air expansion mode and rotary Wankel engine in steam expansion mode and air expansion mode. The air tests will provide base data for friction and motoring effects whereas steam tests will tell how effective the engines will be in this mode. Results for power, torque, and p-V diagrams are compared to determine the change in performance from air expansion mode to steam expansion mode.

Stage 1: Expression of interest and consultation document for natural gas distribution in New Brunswick

International Nuclear Information System (INIS)

1998-01-01

The New Brunswick government intends to award a franchise to establish natural gas distribution in the province. To this end, the province wishes to invite bids from qualified entities to establish gas distribution facilities. The province will select the preferred bidder(s) through a two-stage competitive bidding process. This document details the province's policy objectives, questions and issues to be addressed in stage 1 of the process, and the schedule for the process. Appendices include copies of relevant provincial statutes and regulations

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

Function analysis of steam isolation valves

International Nuclear Information System (INIS)

Persson, R.; Kilpi, K.; Noro, H.; Siikonen, T.; Sjoeberg, A.; Wallen, G.; Aakesson, H.

1981-01-01

Function analysis of system-medium-operated steam isolation valves has been the objective of the Swedish-Finnish IVLS project, the results of which are presented in this report. Theoretical models were to be verified against available experimental data, to some extent from the HDR blowdown experiments, which are part of a German reactor safety program. Finnish hydraulic measurements on a valve model (scale 1:2.15) have been performed to give complementary data. The analysis work has covered the thermal-hydraulic behaviour of steam isolation valves as well as phenomena related to structural mechanics. Work performed under contract with the Swedish Nuclear Power Inspectorate. (Author)

Changing the simualtor's steam generator

International Nuclear Information System (INIS)

Ruiz Martin, J.A.; Ortega Pascual, F.

2006-01-01

Two Spanish nuclear power plants (two PWR units each one) have planned to change their Westinghouse D-3 steam generators (SGo henceforth) for a new model, 61W/D3 from Siemens/KWU (SGn henceforth), during 1995/1997. This is the reason why TECNATOM has developed during 1994's last term, a new software for the full scope simulator that incorporates the modifications related to the steam generator substiution programme. This allows an anticipated training on the procedures, not only for normal, but for emergency procedures. As it is a component which has not yet been included in these plants, there are not real references or operative experience data. Therefore, the design of the validation strategy was one of the key points in this work. (author)

33 CFR 334.450 - Cape Fear River and tributaries at Sunny Point Army Terminal, Brunswick County, N.C.; restricted...

Science.gov (United States)

2010-07-01

... 33 Navigation and Navigable Waters 3 2010-07-01 2010-07-01 false Cape Fear River and tributaries... AND RESTRICTED AREA REGULATIONS § 334.450 Cape Fear River and tributaries at Sunny Point Army Terminal, Brunswick County, N.C.; restricted area. (a) The area. That portion of Cape Fear River due west of the main...

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%

Strategies for steam

International Nuclear Information System (INIS)

Hennagir, T.

1996-01-01

This article is a review of worldwide developments in the steam turbine and heat recovery steam generator markets. The Far East is driving the market in HRSGs, while China is driving the market in orders placed for steam turbine prime movers. The efforts of several major suppliers are discussed, with brief technical details being provided for several projects

Flow-induced vibration in LMFBR steam generators: a state-of-the-art review

International Nuclear Information System (INIS)

Shin, Y.S.; Wambsganss, M.W.

1975-05-01

This state-of-the-art review identifies and discusses existing methods of flow-induced vibration analysis applicable to steam generators, their limitations, and base-technology needs. Also included are discussions of five different LMFBR steam-generator configurations and important design considerations, failure experiences, possible flow-induced excitation mechanisms, vibration testing, and available methods of vibration analysis. The objectives are to aid LMFBR steam-generator designers in making the best possible evaluation of potential vibration in steam-generator internals, and to provide the basis for development of design guidelines to avoid detrimental flow-induced vibration

Condensation of steam

International Nuclear Information System (INIS)

Prisyazhniuk, V.A.

2002-01-01

An equation for nucleation kinetics in steam condensation has been derived, the equation taking into account the concurrent and independent functioning of two nucleation mechanisms: the homogeneous one and the heterogeneous one. The equation is a most general-purpose one and includes all the previously known condensation models as special cases. It is shown how the equation can be used in analyzing the process of steam condensation in the condenser of an industrial steam-turbine plant, and in working out new ways of raising the efficiency of the condenser, as well as of the steam-turbine plant as a whole. (orig.)

Study on premixing phase of steam explosion at JAERI

Energy Technology Data Exchange (ETDEWEB)

Yamano, Norihiro; Moriyama, Kiyofumi; Maruyama, Yu; Park, H.; Yang, Y.; Sugimoto, Jun [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-01-01

Melt jet breakup (MJB) and fragmentation has been studied in the frame of ALPHA program. In the first two experiments of MJB series, jet of molten lead-bismuth eutectic alloy was released into a deep pool of saturated water. Steam generation rate was measured and correlated with the jet behavior observed by a high-speed camera. The jet breakup length and debris size distribution were also evaluated. In parallel with the experimental study, JASMINE code has been developed for the simulation of steam explosion. The melt jet breakup model and the particle breakup model in the code were tested by analyzing FARO-L14 and ALPHA MJB experiments. (author)

Zirconium metal-water oxidation kinetics. V. Oxidation of Zircaloy in high pressure steam

International Nuclear Information System (INIS)

Pawel, R.E.; Cathcart, J.V.; Campbell, J.J.; Jury, S.H.

1977-12-01

A series of scoping tests to determine the influence of steam pressure on the isothermal oxidation kinetics of Zircaloy-4 PWR tubing was undertaken. The oxidation experiments were conducted in flowing steam at 3.45, 6.90, and 10.34 MPa (500, 1000, and 1500 psi) at 905 0 C (1661 0 F), and at 3.45 and 6.90 MPa at 1101 0 C (2014 0 F). A comparison of the results of these experiments with those obtained for oxidation in steam at atmospheric pressure under similar conditions indicated that measurable enhancement of the oxidation rate occurred with increasing pressure at 905 0 C, but not at 1100 0 C

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

International Nuclear Information System (INIS)

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

2006-01-01

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

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

Steam Generator Owners Group PWR secondary water chemistry guidelines

International Nuclear Information System (INIS)

Welty, C.S. Jr.; Green, S.J.

1985-01-01

In 1981 the Steam Generator Owners Group (SGOG), a group of domestic and foreign pressurized water reactor (PWR) owners, developed and issued the PWR secondary water chemistry guidelines. The guidelines were prepared in response to the growing recognition that a majority of the problems causing reduced steam generator reliability (e.g., denting, wasteage, pitting, etc.) were related to secondary (steam) side water purity. The guidelines were subsequently issued as an Electric Power Research Institute (EPRI) report. In 1984 they were revised to reflect industry experience in adopting the original issuance and to incorporate new information on causes of corrosion damage. The guidelines have been endorsed and their adoption recommended by the SGOG

A reflux capsule steam generator for sodium cooled reactors

International Nuclear Information System (INIS)

Lantz, E.

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

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

Interfacial heat transfer in countercurrent flows of steam and water

International Nuclear Information System (INIS)

Megahed, M.M.

1987-04-01

A study was conducted to examine the departure from equilibrium conditions with respect to direct contact condensation. A simple analytical model, which used an equilibrium factor, K, was derived. The model was structured to represent the physical dimensions of a nuclear reactor downcomer annulus, water subcooling, wall temperature, and water flow rate. In a two step process the model was first used to isolate the average interfacial heat transfer coefficient from vertical countercurrent steam/water data of Cook et al., with the aid of a Stanton number correlation. In the second step the model was assessed by regeneration of measured steam flow rates in the experiments by Cook et al., and an additional experiment of Kim. This report documents the analytical model, the derived Stanton number correlation, and the comparison of the calculated and measured steam flow rates by which the accuracy of the model was assessed

Technical diagnostics of steam turbines

International Nuclear Information System (INIS)

Vlckova, B.; Drahy, J.

1987-01-01

This paper deals with practical experience in application of technical diagnostics methods to steam turbines, in particular using pedestal and shaft vibration measurements as well as estimation of bearing metal temperature and ultrasound emission signals. An estimation of effectiveness of the diagnostics methods used is given on the basis of experimental investigations made on a 30-MW turbine. (author)

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

Study on antioxidant experiment on forged steel tube sheet and tube hole for steam generator

International Nuclear Information System (INIS)

Zong Hai; Wang Detai; Ding Yang

2012-01-01

Antioxidant experiment on forged steel tube sheet and tube hole for steam generator was studied and the influence of different simulated heat treatments on the antioxidant performance of tube sheet and tube hole was made. The influence of different antioxidant methods on the size of tube hole was drawn. Furthermore, the change of size and weight of 18MnD5 forged steel tube sheet on the condition of different simulated heat treatments was also studied. The analytical results have proved reference information for the use of 18MnD5 material and for key processes of processing tube hole and wearing and expanding U-style tube. (authors)

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

How to compute the power of a steam turbine with condensation, knowing the steam quality of saturated steam in the turbine discharge

Energy Technology Data Exchange (ETDEWEB)

Hernandez Albarran, Manuel Jaime; Krever, Marcos Paulo Souza [Braskem, Sao Paulo, SP (Brazil)

2009-07-01

To compute the power and the thermodynamic performance in a steam turbine with condensation, it is necessary to know the quality of the steam in the turbine discharge and, information of process variables that permit to identifying with high precision the enthalpy of saturated steam. This paper proposes to install an operational device that will expand the steam from high pressure point on the shell turbine to atmosphere, both points with measures of pressure and temperature. Arranging these values on the Mollier chart, it can be know the steam quality value and with this data one can compute the enthalpy value of saturated steam. With the support of this small instrument and using the ASME correlations to determine the equilibrium temperature and knowing the discharge pressure in the inlet of surface condenser, the absolute enthalpy of the steam discharge can be computed with high precision and used to determine the power and thermodynamic efficiency of the turbine. (author)

Proceedings of the 2007 Annual Meeting of the Canadian Mathematics Education Study Group = Actes de la Rencontre Annuelle 2007 du Groupe Canadien d'Etude en Didactique des Mathematiques (31st, Fredricton, New Brunswick, Canada, Jun 8-12, 2007)

Science.gov (United States)

Liljedahl, Peter, Ed.

2008-01-01

This submission contains the Proceedings of the 2007 Annual Meeting of the Canadian Mathematics Education Study Group (CMESG), held at the University of New Brunswick in Fredricton, New Brunswick. The CMESG is a group of mathematicians and mathematics educators who meet annually to discuss mathematics education issues at all levels of learning.…

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.

Steam Digest: Volume IV

Energy Technology Data Exchange (ETDEWEB)

2004-07-01

This edition of the Steam Digest is a compendium of 2003 articles on the technical and financial benefits of steam efficiency, presented by the stakeholders of the U.S. Department of Energy's BestPractices Steam effort.

Steam Digest Volume IV

Energy Technology Data Exchange (ETDEWEB)

None

2004-07-01

This edition of the Steam Digest is a compendium of 2003 articles on the technical and financial benefits of steam efficiency, presented by the stakeholders of the U.S. Department of Energy's BestPractices Steam effort.

Preliminary investigation of aluminum combustion in air and steam.

OpenAIRE

Hallenbeck, Amos Edward.

1983-01-01

Approved for public release; distribution in unlimited. The goal of the experiment is to understand the role of metal-steam combustion in the explosion of underwater shaped cnarges. An apparatus was constructed to investigate combustion of aluminum in stes.m. For background information, aluminum wires (1 mm diameter, 50 mm length) were ignited in air by high current (480 amperes) . Tests in air and steam were photographed using 35 mm color slides and 16 mm movies (4300 fr...

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)

Experience with vacuum distillation cleaning of a full-size steam generator

Energy Technology Data Exchange (ETDEWEB)

Smit, C C

1975-07-01

In the 50 MW Sodium Component Test Facility at Hengelo tests are conducted on several types of full size prototype steam generators and an intermediate heat exchanger. The necessary post-test examination of these prototype components requires a complete removal of all sodium. Since in some cases the endurance test has to be continued after internal inspection, the cleaning-method should be such that no damage occurs to the component. After partial disassembly and internal inspection the component will be reassembled and must be acceptable for further use. The qualification tests of the Neratoom straight tube steam generator were concluded in June 1974. The evaporator module was decided to be partially disassembled in order to meet the requirement of a thorough examination before fabrication of the SNR-generators is started. In preparation for the most suitable cleaning procedure, several methods of sodium removal were considered.

Experience with vacuum distillation cleaning of a full-size steam generator

International Nuclear Information System (INIS)

Smit, C.C.

1975-01-01

In the 50 MW Sodium Component Test Facility at Hengelo tests are conducted on several types of full size prototype steam generators and an intermediate heat exchanger. The necessary post-test examination of these prototype components requires a complete removal of all sodium. Since in some cases the endurance test has to be continued after internal inspection, the cleaning-method should be such that no damage occurs to the component. After partial disassembly and internal inspection the component will be reassembled and must be acceptable for further use. The qualification tests of the Neratoom straight tube steam generator were concluded in June 1974. The evaporator module was decided to be partially disassembled in order to meet the requirement of a thorough examination before fabrication of the SNR-generators is started. In preparation for the most suitable cleaning procedure, several methods of sodium removal were considered

KNK steam generator leakage, evaluations and improvements

International Nuclear Information System (INIS)

Dumm, K.; Ratzel, W.

1975-01-01

On September 23, 1972 the KNK reactor was shut down due to a sodium-water reaction. Detection, localisation and possible leak development are described and discussed. Improvements on the KNK steam generator system due to this leak experience are explained. (author)

KNK steam generator leakage, evaluations and improvements

Energy Technology Data Exchange (ETDEWEB)

Dumm, K; Ratzel, W

1975-07-01

On September 23, 1972 the KNK reactor was shut down due to a sodium-water reaction. Detection, localisation and possible leak development are described and discussed. Improvements on the KNK steam generator system due to this leak experience are explained. (author)

Steam-chemical reactivity for irradiated beryllium

Energy Technology Data Exchange (ETDEWEB)

Anderl, R.A.; McCarthy, K.A.; Oates, M.A.; Petti, D.A.; Pawelko, R.J.; Smolik, G.R. [Idaho National Engineering and Environmental Lab., Idaho Falls, ID (United States)

1998-01-01

This paper reports the results of an experimental investigation to determine the influence of neutron irradiation effects and annealing on the chemical reactivity of beryllium exposed to steam. The work entailed measurements of the H{sub 2} generation rates for unirradiated and irradiated Be and for irradiated Be that had been previously annealed at different temperatures ranging from 450degC to 1200degC. H{sub 2} generation rates were similar for irradiated and unirradiated Be in steam-chemical reactivity experiments at temperatures between 450degC and 600degC. For irradiated Be exposed to steam at 700degC, the chemical reactivity accelerated rapidly and the specimen experienced a temperature excursion. Enhanced chemical reactivity at temperatures between 400degC and 600degC was observed for irradiated Be annealed at temperatures of 700degC and higher. This reactivity enhancement could be accounted for by the increased specific surface area resulting from development of a surface-connected porosity in the irradiated-annealed Be. (author)

Structure of steam water mixture spray

International Nuclear Information System (INIS)

Mitsuhashi, Yuki; Mizutani, Hiroya; Sanada, Toshiyuki; Saito, Takayuki

2008-01-01

The flow structure of steam and water mixture spray is studied both numerically and experimentally. The velocity and pressure profiles of the single phase flow are calculated using numerical methods. Using calculated flow fields, the droplet behavior is predicted by the one-way interaction model. This numerical analysis clarifies that the droplets are still accelerated after they are sprayed from the nozzle. In the experiments, the spray of the mixture is observed by using ultra high-speed video camera, and the velocity field is measured by using PIV technique. Along with this PIV velocity field measurement, the velocities and diameters of droplets are measured by phase Doppler anemometry. Furthermore, mixing process of steam and water, and atomization process of liquid film are observed through the transparent nozzle. The high-speed photography observation reveals that the flow inside the nozzle forms the annular flow and the most of the liquid film is atomized at the nozzle outlet. Finally, the optimum method of processing mixture of steam and water is proposed. (author)

Selling steam

International Nuclear Information System (INIS)

Zimmer, M.J.; Goodwin, L.M.

1991-01-01

This article addresses the importance of steam sales contract is in financing cogeneration facilities. The topics of the article include the Public Utility Regulatory Policies Act provisions and how they affect the marketing of steam from qualifying facilities, the independent power producers market shift, and qualifying facility's benefits

Vaporization of tungsten-metal in steam at high temperatures

International Nuclear Information System (INIS)

Greene, G.A.; Finfrock, C.C.

2000-01-01

The vaporization of tungsten from the APT spallation target dominates the radiological source term for unmitigated target overheating accidents. Chemical reactions of tungsten with steam which persist to tungsten temperatures as low as 800 C result in the formation of a hydrated tungsten-oxide which has a high vapor pressure and is readily convected in a flowing atmosphere. This low-temperature vaporization reaction essentially removes the oxide film that forms on the tungsten-metal surface as soon as it forms, leaving behind a fresh metallic surface for continued oxidation and vaporization. Experiments were conducted to measure the oxidative vaporization rates of tungsten in steam as part of the effort to quantify the MT radiological source term for severe target accidents. Tests were conducted with tungsten rods (1/8 inch diameter, six inches long) heated to temperatures from approximately 700 C to 1350 C in flowing steam which was superheated to 140 C. A total of 19 experiments was conducted. Fifteen tests were conducted by RF induction heating of single tungsten rods held vertical in a quartz glass retort. Four tests were conducted in a vertically-mounted tube furnace for the low temperature range of the test series. The aerosol which was generated and transported downstream from the tungsten rods was collected by passing the discharged steam through a condenser. This procedure insured total collection of the steam along with the aerosol from the vaporization of the rods. The results of these experiments revealed a threshold temperature for tungsten vaporization in steam. For the two tests at the lowest temperatures which were tested, approximately 700 C, the tungsten rods were observed to oxidize without vaporization. The remainder of the tests was conducted over the temperature range of 800 C to 1350 C. In these tests, the rods were found to have lost weight due to vaporization of the tungsten and the missing weight was collected in the downstream condensate

Wet steam wetness measurement in a 10 MW steam turbine

Directory of Open Access Journals (Sweden)

Kolovratník Michal

2014-03-01

Full Text Available The aim of this paper is to introduce a new design of the extinction probes developed for wet steam wetness measurement in steam turbines. This new generation of small sized extinction probes was developed at CTU in Prague. A data processing technique is presented together with yielded examples of the wetness distribution along the last blade of a 10MW steam turbine. The experimental measurement was done in cooperation with Doosan Škoda Power s.r.o.

The New Brunswick Laboratory Safeguards Measurement Evaluation Program

International Nuclear Information System (INIS)

Cacic, C.G.; Trahey, N.M.; Zook, A.C.

1987-01-01

The New Brunswick Laboratory (NBL) has been tasked by the U.S. Department of Energy (DOE) Office of Safeguards and Security (OSS) to assess and evaluate the adequacy of measurement technology as applied to materials accounting in DOE nuclear facilities. The Safeguards Measurement Evaluation (SME) Program was developed as a means to monitor and evaluate the quality and effectiveness of accounting measurements by site, material balance area (MBA), or unit process. Phase I of the SME Program, initiated during 1985, involved evaluation of the primary accountability measurement methods at six DOE Defense Programs facilities: Savannah River Plant, Portsmouth Gaseous Diffusion Plant, Y-12 Plant, Rocky Flats Plant, Rockwell Hanford Operations, and NBL. Samples of uranyl nitrate solution, dried plutonium nitrates, and plutonium oxides were shipped to the participants for assay and isotopic abundance measurements. Resulting data are presented and evaluated as indicators of current state-of-the-practice accountability measurement methodology, deficiencies in materials accounting practices, and areas for possible assistance in upgrading measurement capabilities. Continuing expansion of the SME Program to include materials which are representative of specific accountability measurement points within the DOE complex is discussed

Pressure drop, steam content and turbulent cross exchange in water/steam flows

International Nuclear Information System (INIS)

Teichel, H.

1978-01-01

For describing the behaviour of two-phase flows of water and steam with the help of calculating patterns, a number of empirical correlations are required. - In this article, correlations for the friction pressure drop in water/steam flows are compared, as well as for the steam mass and the volumetric steam content with each other and with the test results on simple geometries. As the mutual effect between cooling chanels plays an important part at the longitudinal flow through bar bundles, the appertaining equations are evaluated, in addition. (orig.) 891 HP [de

Primary separator replacement for Bruce Unit 8 steam generators

International Nuclear Information System (INIS)

Roy, S.B.; Mewdell, C.G.; Schneider, W.G.

2000-01-01

During a scheduled maintenance outage of Bruce Unit 8 in 1998, it was discovered that the majority of the original primary steam separators were damaged in two steam generators. The Bruce B steam generators are equipped with GXP type primary cyclone separators of B and W supply. There were localized perforations in the upper part of the separators and large areas of generalized wall thinning. The degradation was indicative of a flow related erosion corrosion mechanism. Although the unit- restart was justified, it was obvious that corrective actions would be necessary because of the number of separators affected and the extent of the degradation. Repair was not considered to be a practical option and it was decided to replace the separators, as required, in Unit 8 steam generators during an advanced scheduled outage. GXP separators were selected for replacement to minimize the impact on steam generator operating characteristics and analysis. The material of construction was upgraded from the original carbon steel to stainless steel to maximize the assurance of full life. The replacement of the separators was a first of a kind operation not only for Ontario Power Generation and B and W but also for all CANDU plants. The paper describes the degradations observed and the assessments, the operating experience, manufacture and installation of the replacement separators. During routine inspection in 1998, many of the primary steam separators in two of steam generators at Bruce Nuclear Division B Unit 8 were observed to have through wall perforations. This paper describes assessment of this condition. It also discusses the manufacture and testing of replacement primary steam separators and the development and execution of the replacement separator installation program. (author)

CANDU steam generator tubing material service experience and allied development

International Nuclear Information System (INIS)

Hart, A.E.; Lesurf, J.E.

1976-01-01

This paper covers the following aspects for the tube materials in CANDU-PHW steam generators: inservice performance with respect to tube leaks and coolant activity attributable to boiler tube corrosion, selection of tube materials for use with non-boiling and boiling primary coolants, supporting development on corrosion, vibration, fretting wear, tube inspection, leak detection and plugging of defective tubes. (author)

Application results of a prototype ultrasonic liquid film sensor to a 7 MPa steam-water two-phase flow experiment

International Nuclear Information System (INIS)

Aoyama, Goro; Fujimoto, Kiyoshi; Katono, Kenichi; Nagayoshi, Takuji; Baba, Atsushi; Yasuda, Kenichi

2016-01-01

A prototype ultrasonic liquid film sensor was applied to a high-temperature steam-water two-phase flow experiment. The liquid film sensor was vertically installed in a loop which was connected to HUSTLE, a multi-purpose steam source test facility. The hydraulic diameter of the measurement section was 9.4 mm. The output waveforms of the sensor were acquired with a digital oscilloscope. The fluid temperature and system pressure were kept at 288°C and 7.2 MPa, respectively, during the experiment. The pulse-echo method was used to calculate the liquid film thickness. The cross-correlation calculation was utilized to determine the time difference between the pulse reflected at the sensor surface and the pulse reflected at the liquid film surface. The time-averaged liquid film thicknesses were less than 0.055 mm in the annular flow condition. The increase of the time-averaged thickness was small with the change of the gas momentum flux. The film thicknesses measured with the sensor were compared with the past experimental results; the former were smaller than one-fourth of the thickness estimated as the mean film thickness. The comparison results suggested that the continuous liquid sublayer thickness was measured with the liquid film sensor. (author)

What is geothermal steam worth?

International Nuclear Information System (INIS)

Thorhallsson, S.; Ragnarsson, A.

1992-01-01

Geothermal steam is obtained from high-temperature boreholes, either directly from the reservoir or by flashing. The value of geothermal steam is similar to that of steam produced in boilers and lies in its ability to do work in heat engines such as turbines and to supply heat for a wide range of uses. In isolated cases the steam can be used as a source of chemicals, for example the production of carbon dioxide. Once the saturated steam has been separated from the water, it can be transported without further treatment to the end user. There are several constraints on its use set by the temperature of the reservoir and the chemical composition of the reservoir fluid. These constraints are described (temperature of steam, scaling in water phase, gas content of steam, well output) as are the methods that have been adopted to utilize this source of energy successfully. Steam can only be transported over relatively short distances (a few km) and thus has to be used close to the source. Examples are given of the pressure drop and sizing of steam mains for pipelines. The path of the steam from the reservoir to the end user is traced and typical cost figures given for each part of the system. The production cost of geothermal steam is estimated and its sensitivity to site-specific conditions discussed. Optimum energy recovery and efficiency is important as is optimizing costs. The paper will treat the steam supply system as a whole, from the reservoir to the end user, and give examples of how the site-specific conditions and system design have an influence on what geothermal steam is worth from the technical and economic points of view

Update of operations with Westinghouse steam generators

International Nuclear Information System (INIS)

Malinowski, D.D.; Fletcher, W.D.

1978-01-01

Westinghouse commercial steam generators in operation now number 112, of which 98 are tubed with Inconel 600, the remainder with stainless steel. The implementation of all volatile treatment (AVT) was reported. It was noted that several plants had exhibited some tube corrosion during their initial periods using AVT; this observation indicated that the transition from phosphate chemistry control to AVT may have been subject to certain residual effects due to incomplete removal of phosphated deposits. As inspection results from steam generators operated on AVT became more generally available with the passage of time, a pattern of results emerged that seemed to correlate with the operating experience with phosphate chemistry control. Specifically, all the plants that experienced corrosion problems had from 1 to 8 yr of operational history using phosphates, while those with less than a year's experience using phosphates tended to be less affected by corrosion problems

Steam Digest 2001

Energy Technology Data Exchange (ETDEWEB)

2002-01-01

Steam Digest 2001 chronicles BestPractices Program's contributions to the industrial trade press for 2001, and presents articles that cover technical, financial and managerial aspects of steam optimization.

Simulation of steam explosion in stratified melt-coolant configuration

International Nuclear Information System (INIS)

Leskovar, Matjaž; Centrih, Vasilij; Uršič, Mitja

2016-01-01

Highlights: • Strong steam explosions may develop spontaneously in stratified configurations. • Considerable melt-coolant premixed layer formed in subcooled water with hot melts. • Analysis with MC3D code provided insight into stratified steam explosion phenomenon. • Up to 25% of poured melt was mixed with water and available for steam explosion. • Better instrumented experiments needed to determine dominant mixing process. - Abstract: A steam explosion is an energetic fuel coolant interaction process, which may occur during a severe reactor accident when the molten core comes into contact with the coolant water. In nuclear reactor safety analyses steam explosions are primarily considered in melt jet-coolant pool configurations where sufficiently deep coolant pool conditions provide complete jet breakup and efficient premixture formation. Stratified melt-coolant configurations, i.e. a molten melt layer below a coolant layer, were up to now believed as being unable to generate strong explosive interactions. Based on the hypothesis that there are no interfacial instabilities in a stratified configuration it was assumed that the amount of melt in the premixture is insufficient to produce strong explosions. However, the recently performed experiments in the PULiMS and SES (KTH, Sweden) facilities with oxidic corium simulants revealed that strong steam explosions may develop spontaneously also in stratified melt-coolant configurations, where with high temperature melts and subcooled water conditions a considerable melt-coolant premixed layer is formed. In the article, the performed study of steam explosions in a stratified melt-coolant configuration in PULiMS like conditions is presented. The goal of this analytical work is to supplement the experimental activities within the PULiMS research program by addressing the key questions, especially regarding the explosivity of the formed premixed layer and the mechanisms responsible for the melt-water mixing. To

LMFBR steam generators in the United Kingdom

International Nuclear Information System (INIS)

Anderson, R.; Hayden, O.

2002-01-01

Experience has been gained in the UK on the operation of LMFBR Steam Generator Units (SGU) over a period of 20 years from the Dounreay Fast Reactor (DFR) and the Prototype Fast Reactor (PFR). The DFR steam generator featured a double barrier and therefore did not represent a commercial design. PFR, however, faced the challenge of a single wall design and it is experience from this which is most valuable. The PFR reactor went critical in March 1974 and the plant operating history since then has been dominated by experience with leaks in the tube to tube plate welds of the high performance U-tubes SGU's. Operation at high power using the full complement of three secondary sodium circuits was delayed until July 1976 by the occurrence of leaks in the tube to tube plate welds of the superheater and reheater units which are fabricated in stainless steel. Repairs were carried out to the two superheaters and they were returned to service. The reheater tube bundle was removed from circuit after sodium was found to have entered the steam side. When the sodium had been removed and inspection carried out it was decided not to recover the unit. Since 1976 the remaining five stainless steel units have operated satisfactorily. This year a replacement reheater unit has been installed. This is of a new design in 9-Cr-Mo ferritic steel using a sleeve through which the steam tube passes to eliminate the tube to tube plate weld. Despite a few early leaks in evaporator tube to tube plate welds up to 1979, these failures did not initially present a major problem. However, in 1980 the rate of evaporator weld failures increased and despite the successful application of a shot peening process to eliminate stress corrosion failures from the water side of the weld, failures traced to the sodium side continued. A sleeving process was developed for application to complete evaporator units on a production basis with the objective of bypassing the welds at each end of the 500 tubes. The decision

Combustion of hydrogen/air/steam mixtures in a repeated obstacle field

International Nuclear Information System (INIS)

Kumar, R.K.; Bowles, E.M.; Koroll, G.W.

1994-01-01

Combustion experiments with hydrogen/air/steam mixtures were performed in a cylindrical vessel of 1.5-m internal diameter and 5.7-m height in a repeated obstacle field. The investigations included hydrogen concentrations in the range of 10 to 20% and steam concentrations of up to 30%. For the mixtures investigated, the flame accelerated very rapidly in the vessel, reached a peak value, and decelerated equally rapidly For hydrogen/air mixtures with hydrogen concentrations above 15%, the flame speeds reached values well in excess of the sonic velocity in the mixture. Addition of steam reduced the flame speed and the peak pressure, however, the reduction was significant only for steam concentrations >20%. Experiments performed with different obstacle spacings and flow blockages indicated that flame speed decreased with increased spacing and increased with increased blockage. The effect of initial pressure on flame speed was found to be small. For a given mixture, the peak flame speed was found to be independent of the igniter location. Simple empirical correlations have been proposed to calculate the flame speeds and peak pressures in a closed vessel with closely spaces repeated obstacles. (author)

Super titanium blades for advanced steam turbines

International Nuclear Information System (INIS)

Coulon, P.A.

1990-01-01

In 1986, the Alsthom Steam Turbines Department launched the manufacture of large titanium alloy blades: airfoil length of 1360 mm and overall length of 1520 mm. These blades are designed for the last-stage low pressure blading of advanced steam turbines operating at full speed (3000 rpm) and rating between 300 and 800 MW. Using titanium alloys for steam turbine exhaust stages as substitutes for chrome steels, due to their high strength/density ratio and their almost complete resistance to corrosion, makes it possible to increase the length of blades significantly and correspondingly that steam passage section (by up to 50%) with a still conservative stresses level in the rotor. Alsthom relies on 8 years of experience in the field of titanium, since as early as 1979 large titanium blades (airfoil length of 1240 mm, overall length of 1430 mm) were erected for experimental purposes on the last stage of a 900 MW unit of the Dampierre-sur-Loire power plant and now totals 45,000 operating hours without problems. The paper summarizes the main properties (chemical, mechanical and structural) recorded on very large blades and is based in particular on numerous fatigue corrosion test results to justify the use of the Ti 6 Al 4 V alloy in a specific context of micrographic structure

Research on axial total pressure distributions of sonic steam jet in subcooled water

International Nuclear Information System (INIS)

Wu Xinzhuang; Li Wenjun; Yan Junjie

2012-01-01

The axial total pressure distributions of sonic steam jet in subcooled water were experimentally investigated for three different nozzle diameters (6.0 mm, 8.0 mm and 10.0 mm). The inlet steam pressure, and pool subcooling subcooled water temperature were in the range of 0.2-0.6 MPa and 420-860 ℃, respectively. The effect of steam pressure, subcooling water temperature and nozzle size on the axial pressure distributions were obtained, and also the characteristics of the maximum pressure and its position were studied. The results indicated that the characteristics of the maximum pressure were influenced by the nozzle size for low steam pressure, but the influence could be ignored for high steam pressure. Moreover, a correlation was given to correlate the position of the maximum pressure based on steam pressure and subcooling water temperature, and the discrepancies of predictions and experiments are within ±15%. (authors)

Development of axial tomography technique for the study of steam explosion

Energy Technology Data Exchange (ETDEWEB)

Lee, Jae Young; Seo, S. W.; You, S. [Handong Golbal Univ., Pohang (Korea, Republic of)

2006-05-15

In this report, axial tomography applying to steam explosion is implemented. When steam explosion experiment is performed, we have seen the difficulty with physical modeling due to the complex phenomena of generated steam, propagation of shock wave and bubble breakup and coalescence. Hence, the uncertainty due to these phenomena is occurred. The fast and global measurement of the steam distribution is imperative to understand the complex phenomena performed during the steam explosion, KAERI have developed the fast and global measuring instrument to monitor such phenomena of axial steam distribution. Generally, X-ray is used as measuring method, but this method is very expensive and has limited measurement area. So we need new method that can substitute X-ray method and in this research, ECT method is replaced. The research is performed dividing within two parts: Software and Hardware. In the software part, the electric field analysis code and algorithm for inverse projection were developed. And, in the hardware part, capacitance measurement circuit is developed to measure up to fF level. Operable axial tomography was analyzed with concept design of axial tomography appropriate to steam explosion accident and analysis code for axial electric field analysis and inverse algorithm were developed, moreover, designing signal analysis system for axial tomography was performed.

Decision in the matter of a generic hearing in respect to market issues and conduct related to the sale of gas and customer services in the natural gas industry in New Brunswick

International Nuclear Information System (INIS)

2003-01-01

In response to changes in the Gas Distribution Act and the Gas Distributor Marketing Regulation, the New Brunswick Board of Commissioners of Public Utilities (Board), issued an order in June 2003 for a generic hearing regarding the sale of gas and customer services in the natural gas industry in New Brunswick. The purpose of the hearing was to have an open discussion among intervenors regarding several issues, such as whether the Code of Conduct provides fair and adequate rules for marketers; should Enbridge Gas New Brunswick (EGNB) be required to follow rules similar to the Code of Conduct; the requirements surrounding letters of credit; the Board's decision on Rules and Regulations regarding the billing conduct of gas distributors and marketers; the information that should be provided by a gas marketer or a distributor to the customer before the customer enters into an agreement for the supply of gas; the need for information on price volatility; the need for EGNB to notify potential customers of all possible suppliers of gas; the removal of automatic renewal clauses from gas retailers' contracts; how the Board should determine if prices are reasonably and sufficiently competitive; financial reporting requirements; the role of the Board in response to customer complaints regarding the sale of gas by a distributor; and, the Board's action regarding Enbridge Atlantic's plan to exit the New Brunswick market. This report provided the response to these issues by 10 intervenors

Effects of non-condensable gas on the condensation of steam

International Nuclear Information System (INIS)

Jackson, J.D.; An, P.; Reinert, A.; Ahmadinejad, M.

2000-01-01

The experimental work reported here was undertaken with the aim of extending the database currently available on the condensation of steam in the presence of non-condensable gases and thereby improving the empirical input to thermal-hydraulic codes which might be used for design and safety assessment of advanced water-cooled nuclear reactors. Heat was removed from flowing mixtures of steam and air in a test section by means of a water-cooled condensing plate. The test facility constructed for the study incorporates a degassing unit which supplies water to a boiler. This delivers steam steadily to a mixing chamber where it joins with a flow of preheated air. The mixture of steam and air is supplied to the bottom of a cylindrical test section in which it flows upwards over a double sided condensing plate which can be vertical, inclined or horizontal, The rate at which heat is removed by cooling water flowing through internal passages in the plate can de determined calorimetrically knowing the flow rate of the water and its temperature rise. After commissioning experiments had shown that reliable measurements of condensation heat transfer rate could be made using the test facility, a programme of development work followed in the course of which three different designs of condensing plate were evaluated in turn. The version eventually used in the main programme of experiments which followed was made from copper. However, its surfaces were coated with a thin layer of nickel and then with one of chromium. It was found that such a surface consistently promoted dropwise condensation and showed no signs of deterioration after lengthy periods of use. The rate of heat removal from pure steam and from mixtures of steam and air in varying proportions was measured as a function of plate sub-cooling for a variety of plate orientations. (author)

Steam-generator tube failures: world experience in water-cooled nuclear power reactors during 1972

International Nuclear Information System (INIS)

Stevens-Guille, P.D.

1975-01-01

During 1972, approximately one in three operating reactors with steam generators incurred tube failures, predominantly near the tube sheet and in the bend region. Various forms of corrosion were the most frequent cause of failure. Eddy-current inspection was the preferred method for locating and investigating the cause of failure. Extensive use was made of both mechanical and explosive plugs for repair. As a class, steam generators with Monel 400 tubes had the lowest failure rates, and those with Inconel 600 tubes had the highest. (U.S.)

Analysis of experiments for vertical out-tube steam condensation in presence of non-condensable gases

International Nuclear Information System (INIS)

Su Jiqiang; Sun Zhongning; Fan Guangming; Guo Zixuan

2014-01-01

In order to investigate the influence of various parameters in the steam condensation heat transfer process with non-condensable gas, and to get a more suitable empirical correlation, the wall under-cooling, pressure and the content of non-condensable gas were studied outside a vertical tube by experiments. The results showed that: at the same pressure, the relationship between wall sub-cooling and HTC is exponential, and helium stratification does not happen within the experimental range. Based on the analysis of various experimental variables, combined with a large number of experimental data, a wider scope of application of the empirical correlation associated is obtained with the experimental value of the error within ±20%. (authors)

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)

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

DEMONSTRATION BULLETIN STEAM ENHANCED REMEDIATION STEAM TECH ENVIRONMENTAL SERVICES, INC.

Science.gov (United States)

Steam Enhanced Remediation is a process in which steam is injected into the subsurface to recover volatile and semivolatile organic contaminants. It has been applied successfully to recover contaminants from soil and aquifers and at a fractured granite site. This SITE demonstra...

North Anna Power Station - Unit 1: Overview of steam generator replacement project activities

International Nuclear Information System (INIS)

Gettler, M.W.; Bayer, R.K.; Lippard, D.W.

1993-01-01

The original steam generators at Virginia Electric and Power Company's (Virginia Power) North Anna Power Station (NAPS) Unit 1 have experienced corrosion-related degradation that require periodic inspection and plugging of steam generator tubes to ensure their continued safe and reliable operation. Despite improvements in secondary water chemistry, continued tube degradation in the steam generators necessitated the removal from service of approximately 20.3 percent of the tubes by plugging, (18.6, 17.3, and 25.1 for steam generators A, B, and C, respectively). Additionally, the unit power was limited to 95 % during, its last cycle of operation. Projections of industry and Virginia Power experience indicated the possibility of mid-cycle inspections and reductions in unit power. Therefore, economic considerations led to the decision to repair the steam generators (i.e., replace the steam generator lower assemblies). Three new Model 51F Steam Generator lower assembly units were ordered from Westinghouse. Virginia Power contracted Bechtel Power Corporation to provide the engineering and construction support to repair the Unit 1 steam generators. On January 4, 1993, after an extended coastdown period, North Anna Unit 1 was brought off-line and the 110 day (breaker-to-breaker) Steam Generator Replacement Project (SGRP) outage began. As of this paper, the outage is still in progress

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.

Design of large steam turbines for PWR power stations

International Nuclear Information System (INIS)

Hobson, G.; Muscroft, J.

1983-01-01

The thermodynamic cycle requirements for use with pressurized water reactors are reviewed and the manner in which thermal efficiency is maximised is outlined. The special nature of the wet steam cycle associated with turbines for this type of reactor is discussed. Machine and cycle parameters are optimised to achieve high thermal efficiency, particular attention being given to arrangements for water separation and steam reheating and to provisions for feedwater heating. Principles and details of mechanical design are considered for a range of both full-speed turbines running at 3000 rpm on 50 Hz systems and half-speed turbines running at 1800 rpm on 60 Hz systems. The importance of service experience with nuclear wet steam turbines and its relevance to the design of modern turbines for pressurized water reactor applications is discussed. (author)

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

Experimental studies of PWR vertical steam generator performance

International Nuclear Information System (INIS)

Ding Xunshen

1998-06-01

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

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)

Fretting wear damage of steam generator tubes and its prediction modeling

International Nuclear Information System (INIS)

Che Honglong; Lei Mingkai

2013-01-01

The steam generator is the key equipment used for the energy transition in nuclear power plant. Since the high-temperature and high-pressure fluid flows with high speed, the steam generator tubes will be excited and vibrate, leading to the tremendous fretting wear problem on the tubes, sometimes even leading to tube cracking. This paper introduces typical fretting wear cases, the result of corresponding simulation wear experiment and damage mechanism which combining mechanical wear and erosion-corrosion. Work rate model could give a reasonable life prediction about the steam generator tube, and this predictive model has been used in nuclear power plant safety assessment. (authors)

Experience with and techniques of diagnosing power plant steam turbines without dismantling

International Nuclear Information System (INIS)

Drapal, A.; Kopecek, K.

1987-01-01

Within the framework of vibration diagnostics of steam turbines at the Dukovany nuclear power plant the following factors were monitored: the summation signal of vibrations (usually the path of vibration movement), the time course of the vibration and the phase angle. In non-steady states also run-in and run-out curves, the absolute vibration of bearing stands and the relative vibration of the rotor are monitored. The method has so far not allowed to diagnose failures of antifriction bearings, loose parts, some gear box defects, the development of cracks in vanes, radial cracks in the disk, etc. Briefly characterized is the portable equipment which is available at the Dukovany nuclear power plant for vibration diagnostics of steam turbines. Suggestions are made for completing the system for monitoring service life, operation economics, the diagnosis of control circuits, etc. (Z.M.)

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.

Enhancement of enzymatic saccharification of Eucalyptus globulus: steam explosion versus steam treatment.

Science.gov (United States)

Martin-Sampedro, Raquel; Revilla, Esteban; Villar, Juan C; Eugenio, Maria E

2014-09-01

Steam explosion and steam pre-treatment have proved capable of enhancing enzymatic saccharification of lignocellulosic materials. However, until now, these methods had not been compared under the same operational conditions and using the same raw material. Both pre-treatments lead to increased yields in the saccharification of Eucalyptus globulus; but results have been better with steam pre-treatments, despite the more accessible surface of exploded samples. The reason for this finding could be enzymatic inhibition: steam explosion causes a more extensive extraction of hemicelluloses and releases a greater amount of degradation products which can inhibit enzymatic action. Enzymatic inhibition is also dependent on the amount and chemical structure of lignin, which was also a contributing factor to the lower enzymatic yields obtained with the most severe pre-treatment. Thus, the highest yields (46.7% glucose and 73.4% xylose yields) were obtained after two cycle of steam treatment, of 5 and 3 min, at 183°C. Copyright © 2014 Elsevier Ltd. All rights reserved.

Mechanical efficiency of the energy release during a steam explosion

International Nuclear Information System (INIS)

Krieg, R.

1997-01-01

The mechanical processes during the expansion phase of a steam explosion with intimately fragmented liquid particles is investigated based on elementary principles and analytical solutions. During a short load pulse, the different densities of the water and the melted particles lead to different velocities. After the load pulse, viscosity effects lead to a slow down of the higher velocities and to a corresponding reconversion of the kinetic energy of the mixture into thermal energy. It is shown that both effects are proportional to each other. The ratio between the residual and the applied mechanical energy is defined as the mechanical efficiency of the steam explosion. Using data typical for a steam explosion in a pressurized water reactor, mechanical efficiencies of <50% are estimated. Considering that the thermodynamic efficiencies are quite limited, the very low conversion rates from thermal energy into mechanical energy observed during steam explosion experiments can be more easily understood

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

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

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)

Steam explosion - physical foundations and relation to nuclear reactor safety

International Nuclear Information System (INIS)

Schumann, U.

1982-08-01

'Steam explosion' means the sudden evaporation of a fluid by heat exchange with a hotter material. Other terms are 'vapour explosion', 'thermal explosion', and 'energetic fuel-coolant interaction (FCI)'. In such an event a large fraction of the thermal energy initially stored in the hot material may possibly be converted into mechanical work. For pressurized water reactors one discusses (e.g. in risk analysis studies) a core melt-down accident during which molten fuel comes into contact with water. In the analysis of the consequences one has to investigate steam explosions. In this report an overview over the state of the knowledge is given. The overview is based on an extensive literature review. The objective of the report is to provide the basic knowledge which is required for understanding of the most important theories on the process of steam explosions. Following topics are treated: overview on steam explosion incidents, work potential, spontaneous nucleation, concept of detonation, results of some typical experiments, hydrodynamic fragmentation of drops, bubbles and jets, coarse mixtures, film-boiling, scenario of a core melt-down accident with possible steam-explosion in a pressurized water reactor. (orig.) [de

Steam explosions of single drops of pure and alloyed molten aluminum

International Nuclear Information System (INIS)

Nelson, L.S.

1995-01-01

Studies of steam explosion phenomena have been performed related to the hypothetical meltdown of the core and other components of aluminum alloy-fueled production reactors. Our objectives were to characterise the triggers, if any, required to initiate these explosions and to determine the energetics and chemical processes associated with these events. Three basic studies have been carried out with 1-10 g single drops of molten aluminum or aluminum-based alloys: untriggered experiments in which drops of melt were released into water; triggered experiments in which thermal-type steam explosions occurred; and one triggered experiment in which an ignition-type steam explosion occurred. In untriggered experiments, spontaneous steam explosions never occurred during the free fall through water of single drops of pure Al or of the alloys studied here. Moreover, spontaneous explosions never occurred upon or during contact of the globules with several underwater surfaces. When Li was present in the alloy, H 2 was generated as a stream of bubbles as the globules fell through the water, and also as they froze on the bottom surface of the chamber. The triggered experiments were performed with pure Al and the 6061 alloy. Bare bridgewire discharges and those focused with cylindrical reflectors produced a small first bubble that collapsed and was followed by a larger second bubble. When the bridgewire was discharged at one focus of an ellipsoidal reflector, a melt drop at the other focus triggered only very mildly in spite of a 30-fold increase in peak pressure above that of the bridgewire discharge without the reflector. Experiments were also performed with globules of high purity Al in which the melt release temperature was progressively increased. Moderate thermal-type explosions were produced over the temperature range 1273-1673 K. At about 1773 K, however, one experiment produced a brilliant flash of light and bubble growth about an order of magnitude faster than normal; it

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)

Operating experience of main steam isolation valves at Fessenheim and Bugey

International Nuclear Information System (INIS)

Dredemis, G.; Fourest, B.; Giroux, C.

1985-07-01

The paper presents the experience of Hopkinson MSIVs over about 40 reactor-years (1977 to 1984) of operation at Fessenheim and Bugey units (900 MWe PWR). The various problems encountered including ageing effects on auxiliary equipments and increases in closure time are discussed. The corrective actions undertaken by the utility and the safety assessment of these events performed by the french safety authorities are also described. This study is the synthesis of an in-depth analysis of Main Steam Isolation Valves (MSIV) and their auxiliary circuits equipping the Bugey and Fessenheim 900 MWe PWR nuclear power plants. These valves are different from those installed in the other French 900 MWe PWR reactors. The evaluation of the operation of these valves was made on the basis of incidents which occured during operation of the units or during the periodic tests, as well as anomalies discovered during maintenance operations. This analysis proved that the anomalies related to the design of the valves, as well as to their manufacture and installation, had been correctly dealt with. Furthermore, it should have also revealed potential anomalies due to ageing of the equipment

Steam power plant

International Nuclear Information System (INIS)

Campbell, J.W.E.

1981-01-01

This invention relates to power plant forced flow boilers operating with water letdown. The letdown water is arranged to deliver heat to partly expanded steam passing through a steam reheater connected between two stages of the prime mover. (U.K.)

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

Steam turbines for the future

International Nuclear Information System (INIS)

Trassl, W.

1988-01-01

Approximately 75% of the electrical energy produced in the world is generated in power plants with steam turbines (fossil and nuclear). Although gas turbines are increasingly applied in combined cycle power plants, not much will change in this matter in the future. As far as the steam parameters and the maximum unit output are concerned, a certain consolidation was noted during the past decades. The standard of development and mathematical penetration of the various steam turbine components is very high today and is applied in the entire field: For saturated steam turbines in nuclear power plants and for steam turbines without reheat, with reheat and with double reheat in fossil-fired power plants and for steam turbines with and without reheat in combined cycle power plants. (orig.) [de

The Invisibility of Steam

Science.gov (United States)

Greenslade, Thomas B., Jr.

2014-01-01

Almost everyone "knows" that steam is visible. After all, one can see the cloud of white issuing from the spout of a boiling tea kettle. In reality, steam is the gaseous phase of water and is invisible. What you see is light scattered from the tiny droplets of water that are the result of the condensation of the steam as its temperature…

Material properties influence on steam explosion efficiency. Prototypic versus simulant melts, eutectic versus non-eutectic melts

International Nuclear Information System (INIS)

Leskovar, M.; Mavko, B.

2006-01-01

A steam explosion may occur during a severe nuclear reactor accident if the molten core comes into contact with the coolant water. A strong enough steam explosion in a nuclear power plant could jeopardize the containment integrity and so lead to a direct release of radioactive material to the environment. Details of processes taking place prior and during the steam explosion have been experimentally studied for a number of years with adjunct efforts in modelling these processes to address the scaling of these experiments. Steam explosion experiments have shown that there are important differences of behaviour between simulant and prototypical melts, and that also at prototypical melts the fuel coolant interactions depend on the composition of the corium. In experiments with prototypic materials no spontaneous steam explosions occurred (except with an eutectic composition), whereas with simulant materials the steam explosions were triggered spontaneously. The energy conversion ratio of steam explosions with prototypic melts is at least one order of magnitude lower than the energy conversion ratio of steam explosions with simulant melts. Although the different behaviour of prototypic and simulant melts has been known for a number of years, there is no reliable explanation for these differences. Consequently it is not possible to reliably estimate whether corium would behave so non-explosive also in reactor conditions, where the mass of poured melt is nearly three orders of magnitude larger than in experimental conditions. An even more fascinating material effect was observed recently at corium experiments with eutectic and non-eutectic compositions. It turned out that eutectic corium always exploded spontaneously, whereas non-eutectic corium never exploded spontaneously. In the paper, a possible explanation of both material effects (prototypic/simulant melts, eutectic/non-eutectic corium) on the steam explosion is provided. A model for the calculation of the

A strategy for the application of steam explosion codes to reactor analysis

International Nuclear Information System (INIS)

Moriyama, Kiyofumi; Nakamura, Hideo

2006-01-01

A technical view on the strategy for the application of steam explosion codes for plant scale analysis is described. It includes assumption of triggering at the time of peak premixed melt mass, tuning of the explosion model on typical alumina steam explosion data, consideration of void and solidification effects as primary mechanism to limit the premixed mass and explosion energetics, choice of simple heat partition models affecting evaporation. The view was developed through experiences in development, verification and application of a steam explosion simulation code, JASMINE, at Japan Atomic Energy Agency (JAEA), as well as participation in OECD SERENA Phase-1 program. (author)

The accelerated oxidation of zircaloy-4 at 700∼900 .deg. C in high pressure steam

International Nuclear Information System (INIS)

Kim, K. P.; Park, K. H.

1999-01-01

To find the effect of pressure on the high temperature oxidation of zircaloy-4, an autoclave capable of measuring the degree of oxidation at high temperatures and high pressure was manufactured. The specimens used in experiments are commercially available Zircaloy-4 used in Kori nuclear power plants. All the measurements were done at 700∼900 .deg. C in steam. Pressure effects were noticed. The oxide thickness was much thicker in high pressure steam, comparing to that in the 1 atm steam. And, the higher is the steam pressure, the thicker becomes the oxide. The enhancement of oxide growth rate at 700∼900 .deg. C in high pressure steam is approximately propotion to the power of 1.0∼1.6 of the ratio of experimental steam pressure to critical steam pressure. There is a critical steam pressure above that the oxidation rate enhances. The critical steam pressure was measured as 30∼40 bar. The enhanced oxidation seems from the oxide cracking due to the tetragonal to monoclinic phase transformation at high pressure steam

Decision in the matter of a generic hearing in respect to market issues and conduct related to the sale of gas and customer services in the natural gas industry in New Brunswick; Decision en l'affaire concernant une audience generale en ce qui touche les preoccupations du marche et la conduite relative a la vente de gaz et de services a la clientele dans l'industrie du gaz naturel au Nouveau-Brunswick

Energy Technology Data Exchange (ETDEWEB)

NONE

2003-09-02

In response to changes in the Gas Distribution Act and the Gas Distributor Marketing Regulation, the New Brunswick Board of Commissioners of Public Utilities (Board), issued an order in June 2003 for a generic hearing regarding the sale of gas and customer services in the natural gas industry in New Brunswick. The purpose of the hearing was to have an open discussion among intervenors regarding several issues, such as whether the Code of Conduct provides fair and adequate rules for marketers; should Enbridge Gas New Brunswick (EGNB) be required to follow rules similar to the Code of Conduct; the requirements surrounding letters of credit; the Board's decision on Rules and Regulations regarding the billing conduct of gas distributors and marketers; the information that should be provided by a gas marketer or a distributor to the customer before the customer enters into an agreement for the supply of gas; the need for information on price volatility; the need for EGNB to notify potential customers of all possible suppliers of gas; the removal of automatic renewal clauses from gas retailers' contracts; how the Board should determine if prices are reasonably and sufficiently competitive; financial reporting requirements; the role of the Board in response to customer complaints regarding the sale of gas by a distributor; and, the Board's action regarding Enbridge Atlantic's plan to exit the New Brunswick market. This report provided the response to these issues by 10 intervenors.

Optimal design of marine steam turbine

International Nuclear Information System (INIS)

Liu Chengyang; Yan Changqi; Wang Jianjun

2012-01-01

The marine steam turbine is one of the key equipment in marine power plant, and it tends to using high power steam turbine, which makes the steam turbine to be heavier and larger, it causes difficulties to the design and arrangement of the steam turbine, and the marine maneuverability is seriously influenced. Therefore, it is necessary to apply optimization techniques to the design of the steam turbine in order to achieve the minimum weight or volume by means of finding the optimum combination of design parameters. The math model of the marine steam turbine design calculation was established. The sensitivities of condenser pressure, power ratio of HP turbine with LP turbine, and the ratio of diameter with height at the end stage of LP turbine, which influence the weight of the marine steam turbine, were analyzed. The optimal design of the marine steam turbine, aiming at the weight minimization while satisfying the structure and performance constraints, was carried out with the hybrid particle swarm optimization algorithm. The results show that, steam turbine weight is reduced by 3.13% with the optimization scheme. Finally, the optimization results were analyzed, and the steam turbine optimization design direction was indicated. (authors)

Steam chugging in a boiling water reactor pressure-suppression system

International Nuclear Information System (INIS)

Pitts, J.H.

1980-01-01

Results of a transient analysis predicting the general characteristics of steam chugging compare well with the results of two large scale experiments: GKM II, test 21 and GKSS, test 16. Predicted fundamental periods of chugging are within 5 and 16 per cent of the respective experimental values. The results of the analysis include effects of air in the drywell, momentum loss and heat transfer in the condensation pipe, direct contact condensation heat transfer at the gas-water interface and momentum and heat transfer in the wetwell water pool. Bubble shape is calculated in two-dimensional cylindrical coordinates. Required inputs to the analysis include the geometry, initial conditions and constants to determine both the steam inlet mass flowrate to the drywell as a function of time and conduction heat transfer through the wall of the condensation pipe. There are no arbitrary free parameters which must be specified to predict specific experiments. Rather, the analysis is based on fundamental physical phenomena, experimental coefficients documented for general heat transfer and fluid mechanics characteristics and standard analytical techniques. The random nature of steam chugging observed in some experiments is partially explained by predicted regimes of chugging and changes in the maximum extent of a bubble below the condensation pipe exit during each regime. (orig.)

Thermo-hydraulic stability study of a steam generator

International Nuclear Information System (INIS)

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

2012-01-01

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

Steam-water separator

International Nuclear Information System (INIS)

Modrak, T.M.; Curtis, R.W.

1978-01-01

A two-stage steam-water separating device is introduced, where the second stage is made as a cyclone separator. The water separated here is collected in the first stage of the inner tube and is returned to the steam raising unit. (TK) [de

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

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)

Fast reactor steam generators with sodium on the tube side. Design and operational parameters

International Nuclear Information System (INIS)

1994-01-01

A comparison of design and operational characteristics as well as analysis of experience gained during the long terms operation of the Micro Module Inverse Steam Generator and Module Inverse Steam Generator at BOR 60 reactor are main aims of this technical report. 20 refs, 47 figs, 14 tabs

Disposal of Steam Generators from Decommissioning of PWR Nuclear Power Plants

International Nuclear Information System (INIS)

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

2008-01-01

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

Restoration ecology: aiding and abetting secondary succession on abandoned peat mines in Nova Scotia and New Brunswick, Canada

OpenAIRE

P.J. Vander Kloet; T.S. Avery; S.P. Vander Kloet; G.R. Milton

2012-01-01

The efficacy of using vegetative clumps derived from seeds with a variety of origins to establish nuclei for regeneration of bog vegetation on abandoned peat mines in Nova Scotia and New Brunswick (Canada) was tested using seeds within scats (excrement) and seeds from berries, various techniques for creating clumps, and different clump sizes. Direct placement of scat pieces on peat in the field did not produce successful colonisation. Vegetative clumps begun in a greenhouse, whether from seed...

Investigation of Steam Flow Behavior During Horizontal Injection into Vertical Annulus

International Nuclear Information System (INIS)

Yoon, Sang H.; Kim, Won J.; Ku, Ja H.; Suh, Kune Y.; Song, Chul H.

2004-01-01

Qualification of uncertainty margins for accidents, which are classified as the design basis accidents, requires thermal hydraulic codes and related code models with an enhanced level of sophistication. In a cold leg break accident, the flow in downcomer is multidimensional and the velocity distribution of the steam flow in downcomer serves as a good example. For observation of the flow behavior near the break, experiments are performed to measure the velocity of the steam flow in a vessel scaled down from the APR1400 (Advanced Power Reactor 1400 MWe). In this case, the steam has a quality approaching unity and thus is dealt with as a single-phase gas. The velocity of the steam flow is measured by micro-Pitot tubes arranged horizontally and vertically around the outer shell of the 1/20 scaled-down test vessel OMEGA (Optimized Multidimensional Experiment Geometric Apparatus). A commercial computational fluid dynamics code yields analytic results of multidimensional flow motion in the complex annular passage with flow obstacles. CFX is run with well-defined boundary conditions to obtain velocity profiles of the steam flow in the annular downcomer. Results of CFX shed light on the experimental setup as to fixing the location and angle of the micro-Pitot tubes, and correcting the sensitivity of the micro- Pitot tubes, for instance. This study aims to improve the multidimensional capability of the MARS code, which is based on RELAP5 and COBRA-IV, in predicting the multiphase flow behavior in the reactor downcomer. MARS is currently based on one- and two-dimensional flow analyses, which tends to distort total flow due to misrepresentation of the local phenomena. It is thus necessary to scrutinize the steam flow path and mechanistically model the momentum variation. These experimental and analytical results can locally be applied to developing the models of specific forms and essential phenomena treated in MARS. (authors)

STEAM DALAM PEMBUATAN PAKAN UNTUK KOMODITAS AKUAKULTUR

Directory of Open Access Journals (Sweden)

Sukarman Sukarman

2010-12-01

Full Text Available Kualitas fisik pakan (pelet untuk hewan akuakultur sangat penting, karena akan dimasukkan ke dalam air dan diharapkan tidak banyak mencemari lingkungan. Salah satu faktor yang berpengaruh dalam menjaga kualitas fisik pakan adalah penambahan dan pengaturan steam pada saat proses pembuatan pelet. Steam adalah aliran gas yang dihasilkan oleh air pada saat mendidih. Steam dibagi menjadi 3 jenis yaitu steam basah, saturated steam, dan superheated steam. Steam yang digunakan dalam proses pembuatan pelet adalah saturated steam. Pengaruh penambahan steam pada kualitas pelet bisa mencapai 20%. Penambahan steam dengan jumlah dan kualitas yang tepat akan menghasilkan pelet berkualitas. Sedangkan jika pengaturan dan penambahannya tidak tepat, maka kualitas fisik pelet akan rendah dan kemungkinan bisa merusak kandungan nutrisi seperti vitamin dan protein. Penambahan steam yang benar bisa dilakukan di dalam kondisioner dengan mengatur retention time, sudut kemiringan paddle conditioner, kecepatan putaran bearing dan menjaga kualitas steam dari mesin boiler sampai dengan kondisioner.

Small Engines as Bottoming Cycle Steam Expanders for Internal Combustion Engines

OpenAIRE

Weerasinghe, Rohitha; Hounsham, Sandra

2017-01-01

Heat recovery bottoming cycles for internal combustion engines have opened new avenues for research into small steam expanders [1]. Dependable data for small steam expanders will allow us to predict on their suitability as bottoming cycle engines and the fuel economy achieved by using them as bottoming cycles. Wankel Engines, with its lower resistance properties at small scale provide excellent contenders for bottoming cycle expanders. Present paper is based on results of experiments carried ...

Synthesis and optimization of steam system networks. 2. Multiple steam levels

CSIR Research Space (South Africa)

Price, T

2010-08-01

Full Text Available The use of steam in heat exchanger networks (HENs) can be reduced by the application of heat integration with the intention of debottlenecking the steam boiler and indirectly reducing the water requirement [Coetzee and Majozi. Ind. Eng. Chem. Res...

Detection and localisation of leaks in steam generators

Energy Technology Data Exchange (ETDEWEB)

Westenbrugge, J.K. van; Visbach, D M [B.V. NERATOM, The Hague (Netherlands)

1978-10-01

Steam generator experience in Netherlands is concentrated in NERATOM, the company that takes part n the construction of LMFBRs: SNR-300 in Kalkar in the framework of the international company INB (Internationale Natrium Brutreaktor Bau) together with the German company INTERATOM and the Belgian company BelgoNucleaire. Experience consists of designing constructing and testing of prototypes of the SNR-300 plant components, constructing and licensing of the components for SNR-300 proper and in conceptual design of components for larger power plants. Supporting research and testing of the prototypes is carried out by the organization for Industrial Research TNO. The two types studied are the straight tube type and the helical tube type steam generator. Descriptions of both types were published earlier. It should be pointed out here that the construction material is the ferritic 2% Cr 1 Mo 1NiNb-steel (DIN Wnr. 1.6770)

Status of Siemens steam generator design and measures to assure continuous long-term reliable operation

International Nuclear Information System (INIS)

Hoch, G.

1999-01-01

Operating pressurized water reactors with U-tube steam generators have encountered difficulties with either one or a combination of inadequate material selection, poor design or manufacturing and an insufficient water chemistry control which resulted in excessive tube degradation. In contrast to the above mentioned problems, steam generators from Siemens/KWU are proving by operating experience that all measures undertaken at the design stage as well as during the operating and maintenance phase were effective enough to counteract any tube corrosion phenomena or other steam generator related problem. An Integrated Service Concept has been developed, applied and wherever necessary improved in order to ensure reliable steam generator operation. The performance of the steam generators is updated continuously, evaluated and implemented in lifetime databases. The main indicator for steam generator integrity are the results of the eddy current testing of the steam generator tubes. Tubes with indications are rated with lifetime threshold values and if necessary plugged, based on individual assessment criteria.(author)

Environmental approvals in New Brunswick : economic considerations

Energy Technology Data Exchange (ETDEWEB)

Marrack, C.; Maitland, R. [Suez Renewable Energy North America, Houston, TX (United States)

2008-07-01

This presentation provided a timeline of economic considerations related to New Brunswick's regulatory approval process for wind power developments. The economics of wind power projects require careful consideration during the initial planning phases, as it is not yet known if projects are viable. Spending in the early stages of a project should therefore be limited to items that focus on components of the environmental impact assessment (EIA) process that include establishing a relationship with First Nations groups in the area, and conducting seasonal bird studies. Economic considerations change when project viability is confirmed, and developers can then progress to conducting traditional knowledge studies and further seasonal bird studies. Baseline information studies should be reviewed, and biophysical surveys should involve the identification of any wetlands, sensitive areas, and rare plants. Archaeology studies are also required by the provincial government, as well as bat studies to determine if the site has a resident population of bats. Public and stakeholder consultations and open houses should then be held with an adequate time-frame for the submission of questions and concerns and the development of mitigation strategies. Project viability should be confirmed before power purchase agreements (PPAs) are signed. After PPAs are signed, the largest economic consideration is the schedule-related risk associated with legal and financial problems. It was concluded that noise impact, visual impact, and socioeconomic assessments and studies can be conducted after the PPA is secured. tabs., figs.

Steam cleaning device

International Nuclear Information System (INIS)

Karaki, Mikio; Muraoka, Shoichi.

1985-01-01

Purpose: To clean complicated and long objects to be cleaned having a structure like that of nuclear reactor fuel assembly. Constitution: Steams are blown from the bottom of a fuel assembly and soon condensated initially at the bottom of a vertical water tank due to water filled therein. Then, since water in the tank is warmed nearly to the saturation temperature, purified water is supplied from a injection device below to the injection device above the water tank on every device. In this way, since purified water is sprayed successively from below to above and steams are condensated in each of the places, the entire fuel assembly elongated in the vertical direction can be cleaned completely. Water in the reservoir goes upward like the steam flow and is drained together with the eliminated contaminations through an overflow pipe. After the cleaning has been completed, a main steam valve is closed and the drain valve is opened to drain water. (Kawakami, Y.)

A balanced strategy in managing steam generator thermal performance

International Nuclear Information System (INIS)

Hu, M. H.; Nelson, P. R.

2009-01-01

This paper presents a balanced strategy in managing thermal performance of steam generator designed to deliver rated megawatt thermal (MWt) and megawatt electric (MWe) power without loss with some amount of thermal margin. A steam generator (SG) is a boiling heat exchanger whose thermal performance may degrade because of steam pressure loss. In other words, steam pressure loss is an indicator of thermal performance degradation. Steam pressure loss is mainly a result of either 1) tube scale induced poor boiling or 2) tube plugging historically resulting from tubing corrosion, wear due to flow induced tube vibration or loose parts impact. Thermal performance degradation was historically due to tube plugging but more recently it is due to poor boiling caused by more bad than good constituents of feedwater impurities. The whole SG industry still concentrates solely on maintenance programs towards preventing causes for tube plugging and yet almost no programs on maintaining adequate boiling of fouled tubes. There can be an acceptable amount of tube scale that provides excellent boiling capacity without tubing corrosion, as operational experience has repeatedly demonstrated. Therefore, future maintenance has to come up balanced programs for allocating limited resources in both maintaining good boiling capacity and preventing tube plugging. This paper discusses also thermal performance degradation due to feedwater impurity induced blockage of tube support plate and thus subsequent water level oscillations, and how to mitigate them. This paper provides a predictive management of tube scale for maintaining adequate steam pressure and stable water level without loss in MWt/MWe or recovering from steam pressure loss or water level oscillations. This paper offers a balanced strategy in managing SG thermal performance to fulfill its mission. Such a strategy is even more important in view of the industry trend in pursuing extended power uprate as high as 20 percent

Steam purity in PWRs

International Nuclear Information System (INIS)

Hopkinson, J.

1982-01-01

Impurities enter the secondary loop of the PWR through both makeup water from lake or well and cooling-water leaks in the condenser. These impurities can be carried to the steam generator, where they cause corrosion deposits to form. Corrosion products in steam are swept further through the system and become concentrated at the point in the low-pressure turbine where steam begins to condense. Several plants have effectively reduced impurities, and therefore corrosion, by installing a demineralizer for the makeup water, a resin-bed system to clean condensed steam from the condenser, and a deaerator to remove oxygen from the water and so lower the risk of system metal oxidation. 5 references, 1 figure

Wet-steam erosion of steam turbine disks and shafts

International Nuclear Information System (INIS)

Averkina, N. V.; Zheleznyak, I. V.; Kachuriner, Yu. Ya.; Nosovitskii, I. A.; Orlik, V. G.; Shishkin, V. I.

2011-01-01

A study of wet-steam erosion of the disks and the rotor bosses or housings of turbines in thermal and nuclear power plants shows that the rate of wear does not depend on the diagrammed degree of moisture, but is determined by moisture condensing on the surfaces of the diaphragms and steam inlet components. Renovating the diaphragm seals as an assembly with condensate removal provides a manifold reduction in the erosion.

Alternatives to electrical cogeneration: The direct application of steam engines

International Nuclear Information System (INIS)

Phillips, W.C.

1993-01-01

Although small to medium sized industrial facilities are aware of electrical cogeneration, often they are too small for it to be economically justifiable. The direct application of steam turbine power to equipment formerly powered by electric motors, can allow them to use steam capacity to reduce electrical demand and consumption, bypassing cogeneration. Cogeneration converts the heat energy of steam into circular mechanical motion and then converts the circular mechanical motion into electricity. Each conversion entails a loss of energy due to friction and other conversion losses. A substantial amount of the generated electricity is then converted back into circular motion with electric motors, again incurring energy losses. Directly applying the mechanical motion of turbines eliminates both the motion-to-electricity (generator) and the electricity-to-motion (motor) conversion losses. Excess steam capacity during the summer is not unusual for facilities that use steam to provide winter heating. Similarly, most of these facilities experience a large electrical demand peak during the cooling season due to the electricity needed to operate centrifugal chillers. Steam capacity via a turbine to power the chillers can allow the boilers to operate at a higher loading while reducing electrical consumption and demand precisely those periods when demand reduction is most needed. In facilities where the steam generating capacity is sufficient, air compressors provide an appropriate year-round application for turbine power. This paper is the result of an on-going project by the Energy Division, State of North Carolina, Department of Economic and Community Development, in conjunction with the University of North Carolina at Charlotte. The objective of this project is to educate the operating engineers and managers of small to medium sized manufacturing facilities on the technical application and economic justification of steam turbine power

Mechanical design of the hot steam headers of the THTR-300 steam generators

International Nuclear Information System (INIS)

Blumer, U.; Stumpf, M.

1988-01-01

The high pressure steam headers of the THTR steam generators have been subject to special attention during the design phase due to the following reasons: these components are the pressure retaining parts with the heaviest wall thickness in the region of the steam generators; they therefore are sensitive to thermal transient conditions; they are operated in the elevated temperature regime, where creep effects cannot be neglected; there is almost no service experience from fossil steam generators with this type of material (Alloy 800). Safety consideration therefore have been rather extensive and have focussed on two main areas which will be treated in this paper: 1. Analytical investigations on the cyclic material behaviour under all specified operating conditions, taking into account the non-elastic response of the material. 2. Limitation of the consequences of a header rupture by installation of heavy whip restraints. Elastic-plastic-creep analyses: The analyses were performed in different stages and are explained in the corresponding order: Evaluation of the critical location on the header and establishment of a simplified model of a nozzle region for further analysis. Preliminary thermal analyses of all specified transient conditions on simplified procedures, in order to establish a severity ranking of the conditions. Establishment of representative loading blocks. Evaluation of the material properties for thermal and structural, especially non-elastic behaviour. Detailed thermal analyses. Detailed structural analyses of the non-elastic cyclic response. Extrapolation for all cycles and assessment of the results by design codes. Discussion of the results. Header whip restraint design: In addition to the above analysis efforts, heavy whip restraints were provided to assure limitation of the effects of a header failure. This pager shows the measures that were taken to restrain the movement in case of longitudinal and transverse breaks: The anti-whip designs are

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

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.

Numerical simulations of pressure fluctuations at branch piping in BWR main steam line

International Nuclear Information System (INIS)

Morita, Ryo; Inada, Fumio; Yoshikawa, Kazuhiro; Takahashi, Shiro

2009-01-01

The power uprating of a nuclear power plant may increase/accelerate degradation phenomena such as flow-induced vibration and wall thinking. A steam dryer was damaged by a high cycle fatigue due to an acoustic-induced vibration at the branch piping of safety relief valves (SRVs) in main steam lines. In this study, we conducted the numerical simulations of steam/air flow around a simplified branch piping to clarify the basic characteristics of resonance. LES simulations were conducted in ordinary pressure/temperature air and steam under BWR plant conditions. In both cases, the excitation of the pressure fluctuations at the branch was observed under some inlet velocity conditions. These fluctuations and inlet conditions were normalized and the obtained results were compared. The normalized results showed that the range and maximum amplitude of pressure fluctuations were almost the same in low-pressure/temperature air and high-pressure/temperature steam. We found that ordinary pressure/temperature air experiments and simulations can possibly clarify the characteristics of the resonance in high-pressure/temperature steam. (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)

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Manpower development for safe operation of nuclear power plant. China. Main steam bypass system operation and maintenance. Task: 6.1.6. Technical report

International Nuclear Information System (INIS)

Stubley, P.H.

1994-01-01

This mission concentrated on the Steam Bypass system of Qinshan Nuclear Power Plant. The system had experienced spurious opening of the bypass valves, disrupting the steam pressure control and the steam generator level control system. A series of commissioning type tests were defined which should allow the operators to revise the setpoints used in the control of the bypass system, and thus prevent spurious opening while maintaining the desired steam pressure control during power maneuvering. Training also included giving experience from other operating plants on aspects of steam and feedwater systems and components, especially as this experience affected maintenance or gave rise to problems. Steam generated maintenance experience is especially applicable, and a future mission is planned for an expert in this field. In addition other aspects of the Chinese nuclear program was assessed to guide future missions. This included assessment of operating procedures from an availability point of view

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.)

Steam generator replacement at Kansai Electric Power Co., Inc

International Nuclear Information System (INIS)

Kimura, S.; Dodo, Takashi; Negishi, Kazuo

1995-01-01

Eleven nuclear units are in operation at the Kansai Electric Power Co., Inc.. In seven of them, Mihama-1·2·3, Takahama-1·2, and Ohi-1·2, comparatively long duration for tube inspection and repair have been required during late annual outages. KEPCO decided to replace all steam generators in these 7 units with the latest model which was improved upon the past degradation experiences, as a result of comprehensive considerations including public confidence in nuclear power generation, maintenability, and economic efficiency. This report presents the design improvements in new steam generators, replacement techniques, and so on. (author)

STEAM by Design

Science.gov (United States)

Keane, Linda; Keane, Mark

2016-01-01

We live in a designed world. STEAM by Design presents a transdisciplinary approach to learning that challenges young minds with the task of making a better world. Learning today, like life, is dynamic, connected and engaging. STEAM (Science, Technology, Environment, Engineering, Art, and Math) teaching and learning integrates information in…

Steam generator replacement from ALARA aspects

International Nuclear Information System (INIS)

Terry, I.; Breznik, B.

2003-01-01

This paper is going to consider radiological related parameters important for steam generator replacement (SGR) implementation. These parameters are identified as ALARA related parameters, owner-contractor relationship, planning, health physics with logistic services, and time required for the replacement. ALARA related parameters such as source or initial dose rate and plant system configuration define the initial conditions for the planning. There is room to optimise work planning. managerial procedures and also the staff during the implementation phase. The overview of these general considerations is based on the following background: using internationally available data and the experience of one of the vendors, i.e. Siemens-Framatome, and management experience of SG replacement which took place at Krsko NPP in the spring of 2000. Generally plant decisions on maintenance or repair procedures under radiation conditions take into account ALARA considerations. But in the main it is difficult to adjudge the results of an ALARA study, usually in the form of a collective dose estimate, because a comparison standard is missing. That is, very often the planned work is of a one-off nature so comparisons are not possible or the scopes are not the same. In such a case the collective doses for other types of work are looked at and a qualitative evaluation is made. In the case of steam generator replacement this is not the case. Over years of steam generator replacements world-wide a standard has been developed gradually. The first part of the following displays an overview of SGR and sets the Krsko SGR in perspective by applying dose analysis. The second part concentrates on the Krsko SGR itself and its ALARA aspects. (authors)

Three steam generator replacement projects in 1995: Consortium Siemens Framatome is well prepared to contribute its experience to the SGR at the Krsko NPP

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 generators replacements and formed a consortium in 1991, the following projects were performed in 1995: Ringhals 3, Tihange 3 and Asco 2. 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)

The EBR-II steam generating system - operation, maintenance, and inspection

International Nuclear Information System (INIS)

Buschman, H.W.; Penney, W.H.; Longua, K.J.

2002-01-01

The Experimental Breeder Reactor II (EBR-II) has operated for 20 years at the Idaho National Engineering Laboratory near Idaho Falls. EBR-II is a Liquid Metal Fast Breeder Reactor (LMFBR) with integrated power producing capability. EBR-II has operated at a capacity factor over 70% in the past few years. Superheated steam is produced by eight natural circulation evaporators, two superheaters, and a conventional steam drum. Steam throttle conditions are 438 C and 8.62 MPa. The designs of the evaporators and superheaters are essentially identical; both are counterflow units with low pressure nonradioactive sodium on the shell side. During the 20 years of operation, components of the steam generator have been subjected to a variety of inspections including visual, dimensional, and ultrasonic. One superheater was removed from service because of anomalous performance and was replaced with an evaporator which was removed, examined, and converted into a superheater. Overall operating experience of the system has been excellent and essentially trouble free. Inspections have not revealed any conditions that are performance or life limiting. (author)

Acoustic Leak Detection under Micro and Small Water Steam Leaks into Sodium for a Protection of the SFR Steam Generator

International Nuclear Information System (INIS)

Kim, Tae-Joon; Jeong, Ji-Young; Kim, Jong-Man; Kim, Byung-Ho; Hahn, Do-Hee; Yugay, Valeriy S.

2008-01-01

The results of an experimental study of water in a sodium leak noise spectrum formation related with a leak noise attenuation and absorption, and at various rates of water into a sodium leak, smaller than 1.0 g/s, are presented. We focused on studying the micro leak dynamics with an increasing rate of water into sodium owing to a self-development from 0.005 till 0.27 g/s. Conditions and ranges for the existence of bubbling and jetting modes in a water steam outflow into circulating sodium through an injector device, for simulating a defect in a wall of a heat-transmitting tube of a sodium water steam generator were determined. On the basis of the experimental leak noise data the simple dependency of an acoustic signal level from the rate of a micro and small leak at different frequency bands is presented to understand the principal analysis for the development of an acoustic leak detection methodology used in a K- 600 steam generator, with the operational experiences for the noise analysis and measurements in BN-600

Mechanical test of E110 cladding material oxidized in hydrogen rich steam atmosphere

International Nuclear Information System (INIS)

Windberg, P.; Perez-Fero, E.

2005-01-01

The behavior of the fuel cladding under accidental conditions has been studied at the AEKI for more than a decade. Earlier, the effect of oxygen and hydrogen content on the mechanical properties was studied separately. The present experiments can help to understand what kind of processes took place in the cleaning tank at Paks NPP (2003). The purpose of our experiments was to investigate high temperature oxidation of E110 cladding in steam + hydrogen mixture. A high temperature tube furnace was used for oxidation of the samples. The oxidation was carried out at three different temperatures (900 0 C, 1000 0 C, 1100 0 C). The hydrogen content in the steam was varied between 19-36 vol%. The oxygen content of the sample was defined as oxidation ratio. Two sizes (length: 2 and 8 mm) of cladding rings and 100 mm long E110 cladding tubes were oxidized. After the oxidation we made compression and tensile tests for rings, and ballooning experiments for 100 mm long tube. The most important conclusions were the following. Oxidation in H-rich steam atmosphere need longer time to get the same oxidation ratio compared to the steam oxidation without hydrogen. The shorter oxidation time results in a more compact oxide layer. The longer oxidation time leads to a cracked oxide layer. (author)

Evaluation of sulfur hexafluoride and helium for steam generator leak location: Final report

International Nuclear Information System (INIS)

Kassen, W.R.

1987-01-01

Since the use of sulfur hexafluoride as a tracer for identifying sources of primary to secondary leakage in PWR steam generators appeared to offer significant sensitivity advantages, the thermal stability of sulfur hexafluoride in water was evaluated at steam generator operating temperature. Significant decomposition was observed after 2 to 4 hours at temperature. Key decomposition products were fluoride and sulfide ions. Based on this observation and these limited test results, the use of SF 6 for PWR steam generator leak location can not be recommended at this time. A survey of 15 utilities was conducted in regard to their application experience with the helium tracer-mass spectroscopy technique for steam generator leak location. Although several successful steam generator integrity programs do not include use of this technique, it has proven to be a useful addition to the inspection program at some plants. No corrosion concerns appear to be associated with this technique

Prediction of localized flow velocities and turbulence in a PWR steam generator: Final report

International Nuclear Information System (INIS)

Stuhmiller, J.H.

1988-05-01

The Steam Generator Project Office (SGPO) of the Steam Generator Owners Group and Electric Power Research Institute has developed a methodology for prediction of steam generator tube buffeting and associated material wear. Turbulent buffeting of steam generator tubes causes low amplitude vibratory response which results in fretting wear at support locations. Concerns raised at the Zion Nuclear Power Plant regarding the useful life of their steam generators prompted this study, in which the SGPO methodology is applied to analysis of the Westinghouse Model 51 steam generator. The specific intent of this project was to calculate turbulent buffeting forces within the tube bank of an operating Model 51 steam generator as a first step in the overall SGPO tube vibration and wear prediction strategy. Attention is focused on flow in the vicinity of anti-vibration bars (U-bend region) and on the flow that leaves the downcomer to impact against peripheral tubes. Other projects utilized the buffeting forces calculated here to determine tube vibratory response, tube-support plate impact statistics, and material wear rates. Besides successfully calculating hydraulic buffeting loads within the tube bank, the present project has enhanced the SGPO methodology and has identified hitherto unnoticed flow phenomena that occur in the steam generator. Experiments have also been carried out to validate numerical computations of the steam generator flow field

Facility to separate water and steam

International Nuclear Information System (INIS)

Loesel, G.

1977-01-01

The water/steam mixture from the pressure vessel e.g. of a BWR is separated by means of centrifugal separators untilizing the natural separation of steam. The steam is supplied to a steam drying vessel and the water to a water collecting tank. These vessels may be combined to a common vessel or connected through additional pipes. From the water collecting tank, arranged below the steam dryer, a feedwater pipe runs back to the pressure vessel. By construction out of individual components cleaning, decontamination, and operating control are essentially simplified. (RW) 891 RW [de

Development of a water separator for steam drying in nuclear power plants

International Nuclear Information System (INIS)

Kall, H.

1979-01-01

In the wet steam of nuclear power plants with light water cooling the content of liquid phase before entering the high- and low-pressure turbine or at the inlet of the reheater connected in between is limited to a few parts per thousand. For the mechanical drying necessary to maintain this moisture limit there is suited the socalled lamellar separator, in which the vertical laminae, arranged one beside the other, confine a great number of separating ducts, bended in periodically changing sequence, through which wet steam is flowing. In the development of such a lamellar separator with internal drain channels described in this paper the droplet motion in the carrier gas stream and the discharge of the separated liquid is treated independently from each other. The mathematical and experimental study of the first-mentioned partial process leads to its complete description in a dimensionless separation diagram. During the subsequent discharge of the primarily separated droplet fluid there may be torn off droplets by the steam flow and carried away from the separator.The limit for the draining capacity reached with the occurrence of this phenomenon was determined in preliminary experiments with air and water and in experiments with wet steam carried out close to operating conditions. (orig.) 891 GL/orig. 892 RRD [de

Thermodynamic study of the injection of steam bubbles in a subcooled liquid

International Nuclear Information System (INIS)

Besset, Jacqueline.

1980-10-01

The behaviour of steam bubbles injected in a subcooled liquid has been studied. Water was the fluid chosen for this experiment for the steam and the liquid. The experiment was carried out at atmospheric pressure and the variable parameters were the ΔT subcooling (difference between the saturation temperature at the pressure under consideration and that of the liquid around the bubbles) and the steam output. We first studied the formation of the bubbles in a wide subcooling range (7 0 C 0 C). In this study a straight correlation was obtained giving the volume of the bubbles formed at the injector outlet, which is valid for a wide range of variables. The implosion of free bubbles in the liquid after they separated from the injector was then studied. In these experiments the significant implosion parameters are the Jakob Ja number, that measures the possibility of the liquid to absorb the heat given off by the condensation, and the Peclet Pe(RO) number, that defines the relative participation of conduction and convection in the heat exchanges. These numbers vary in the areas: 35 [fr

Corrosion of PWR steam generators

International Nuclear Information System (INIS)

Garnsey, R.

1979-01-01

Some designs of pressurized water reactor (PWR) steam generators have experienced a variety of corrosion problems which include stress corrosion cracking, tube thinning, pitting, fatigue, erosion-corrosion and support plate corrosion resulting in 'denting'. Large international research programmes have been mounted to investigate the phenomena. The operational experience is reviewed and mechanisms which have been proposed to explain the corrosion damage are presented. The implications for design development and for boiler and feedwater control are discussed. (author)

Drying system for steam generators, particularly for steam generators of nuclear power stations

International Nuclear Information System (INIS)

Lavalerie, Claude; Borrel, Christian.

1982-01-01

A drying system is described which allows for modular construction and which provides a significant available exchange area in a reduced volume. All the drying elements are identical and are distributed according to a ternay circular symmetry and are placed radially and associated to steam guiding facilities which alternately provide around the axis of revolution an output volume of dry steam from one element and an input volume of wet steam in the following element [fr

Taking tea in the parlour: middle-class formation and gender construction in Nova Scotia and New Brunswick, 1760-1850

OpenAIRE

Poole, Ann Judith

2007-01-01

Knowledge of tea etiquette was a significant marker of middle-class gentility and contributed to middle-class formation and gender construction in colonial Nova Scotia and New Brunswick. Early middle-class settlers brought tea ware and employed the tea ritual to set standards of gentility that determined social inclusion or exclusion. Tea drinking shifted from a predominantly masculine activity in the late eighteenth century, as commercial and political men met in their parlours over tea, to ...

Maintenance and repair of LMFBR steam generators: specialists` meeting, O-Arai Engineering Center, Japan, 4-8 June 1984. Summary report

Energy Technology Data Exchange (ETDEWEB)

NONE

1984-07-01

The Specialists` Meeting on "Maintenance and Repair of LMFBR Steam Generators" was held in Oarai, Japan, from 4-8 June 1984. The meeting was sponsored by the International Atomic Energy Agency on the recommendation of the IAEA International Working Group on Fast Reactors and was hosted by the Power Reactor and Nuclear Fuel Development Corporation of Japan. The purpose of the meeting was to review and discuss the experience accumulated in various countries on the general design philosophy of LMFBR steam generators from the view point of maintenance and repair, in-service inspection of steam generator tube bundles, identification and inspection of failed tubes and the cleaning and repairing of failed steam generators. The following main topical areas were discussed by participants: national review presentations on maintenance and repair of LMFBR steam generators - design philosophy for maintenance and repair; research and development work on maintenance and repair; and experience on steam generator maintenance and repair.

In-vessel coolability and steam explosion in Nordic BWRs

International Nuclear Information System (INIS)

Ma, W.; Hansson, R.; Li, L.; Kudinov, P.; Cadinu, F.; Tran, C-.T.

2010-05-01

The INCOSE project is to reduce the uncertainty in quantification of steam explosion risk and in-vessel coolability in Nordic BWR plants with the cavity flooding as a severe accident management (SAM) measure. During 2009 substantial advances and new insights into physical mechanisms were gained for studies of: (i) in-vessel corium coolability - development of the methodologies to assess the efficiency of the control rod guide tube (CRGT) cooling as a potential SAM measure; (ii) debris bed coolability - characterization of the effective particle diameter of multi-size particles and qualification of friction law for two-phase flow in the beds packed with multi-size particles; and (iii) steam explosion - investigation of the effect of binary oxides mixtures properties on steam explosion. An approach for coupling of ECM/PECM models with RELAP5 was developed to enhance predictive fidelity for melt pool heat transfer. MELCOR was employed to examine the CRGT cooling efficiency by considering an entire accident scenario, and the simulation results show that the nominal flowrate (∼10kg/s) of CRGT cooling is sufficient to maintain the integrity of the vessel in a BWR of 3900 MWth, if the water injection is activated no later than 1 hour after scram. The POMECO-FL experimental data suggest that for a particulate bed packed with multi-size particles, the effective particle diameter can be represented by the area mean diameter of the particles, while at high velocity (Re>7) the effective particle diameter is closer to the length mean diameter. The pressure drop of two-phase flow through the particulate bed can be predicted by Reed's model. The steam explosion experiments performed at high melt superheat (>200oC) using oxidic mixture of WO3-CaO didn't detect an apparent difference in steam explosion energetics and preconditioning between the eutectic and noneutectic melts. This points out that the next step of MISTEE experiment will be conducted at lower superheat. (author)

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

AGE RELATED DEGRADATION OF STEAM GENERATOR INTERNALS BASED ON INDUSTRY RESPONSES TO GENERIC LETTER 97-06

International Nuclear Information System (INIS)

SUBUDHI, M.; SULLIVAN, JR. E.J.

2002-01-01

THIS PAPER PRESENTS THE RESULTS OF AN AGING ASSESSMENT OF THE NUCLEAR POWER INDUSTRY RESPONSES TO NRC GENERIC LETTER 97-06 ON THE DEGRADATION OF STEAM GENERATOR INTERNALS EXPERIENCED AT ELECTRICITE DE FRANCE (EDF) PLANTS IN FRANCE AND AT A UNITED STATES PRESSURIZED WATER REACTOR (PWR). WESTINGHOUSE (W), COMBUSTION ENGINEERING (CE), AND BABCOCK AND WILCOX (BW) STEAM GENERATOR MODELS, CURRENTLY IN SERVICE AT U.S. NUCLEAR POWER PLANTS, POTENTIALLY COULD EXPERIENCE DEGRADATION SIMILAR TO THATFOUND AT EDF PLANTS AND THE U.S. PLANT. THE STEAM GENERATORS IN MANY OF THE U.S. PWRS HAVE BEEN REPLACED WITH STEAM GENERATORS WITH STEAM GENERATORS WITH IMPROVED DESIGNS AND MATERIALS. THESE REPLACEMENT STEAM GENERATORS HAVE BEEN MANUFACTURED IN THE U.S. AND ABROAD. DURING THIS ASSESSMENT, EACH OF THE THREE OWNERS GROUPS (W,CE, AND BW) IDENTIFIED FOR ITS STEAM GENERATOR, MODELS ALL THE POTENTIAL INTERNAL COMPONENTS THAT ARE VULNERABLE TO DEGRADATION WHILE IN SERVICE. EACH OWNERS GROUPDEVELOPED INSPEC TION AND MONITORING GUIDANCE AND RECOMMENDATIONS FOR ITS PARTICULAR STEAM GENERATOR MODELS. THE NUCLEAR ENERGY INSTITUTE INCORPORATED IN NEI 97-06 STEAM GENERATOR PROGRAM GUIDELINES, A REQUIREMENT TO MONITOR SECONDARY SIDE STEAM GENERATOR COMPONENTS IF THEIR FAILURE COULD PREVENT THE STEAM GENERATOR FROM FULFILLING ITS INTENDED SAFETY-RELATED FUNCTION. LICENSEES INDICATED THAT THEY IMPLEMENTED OR PLANNED TO IMPLEMENT, AS APPROPRIATE FOR THEIR STEAM GENERATORS, THEIR OWNERS GROUPRECOMMENDATIONS TO ADDRESS THE LONG-TERM EFFECTS OF THE POTENTIAL DEGRADATION MECHANISMS ASSOCIATED WITH THE STEAM GENERATOR INTERNALS

Study on steam pressure characteristics in various types of nozzles

Science.gov (United States)

Firman; Anshar, Muhammad

2018-03-01

Steam Jet Refrigeration (SJR) is one of the most widely applied technologies in the industry. The SJR system was utilizes residual steam from the steam generator and then flowed through the nozzle to a tank that was containing liquid. The nozzle converts the pressure energy into kinetic energy. Thus, it can evaporate the liquid briefly and release it to the condenser. The chilled water, was produced from the condenser, can be used to cool the product through a heat transfer process. This research aims to study the characteristics of vapor pressure in different types of nozzles using a simulation. The Simulation was performed using ANSYS FLUENT software for nozzle types such as convergent, convrgent-parallel, and convergent-divergent. The results of this study was presented the visualization of pressure in nozzles and was been validated with experiment data.

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

Exergy Steam Drying and Energy Integration

Energy Technology Data Exchange (ETDEWEB)

Verma, Prem; Muenter, Claes (Exergy Engineering and Consulting, SE-417 55 Goeteborg (Sweden)). e-mail: verma@exergyse.com

2008-10-15

Exergy Steam Drying technology has existed for past 28 years and many new applications have been developed during this period. But during past few years the real benefits have been exploited in connection with bio-fuel production and energy integration. The steam dryer consists of a closed loop system, where the product is conveyed by superheated and pressurised carrier steam. The carrier steam is generated by the water vapours from the product being dried, and is indirectly superheated by another higher temperature energy source such as steam, flue gas, thermal oil etc. Besides the superior heat transfer advantages of using pressurised steam as a drying medium, the energy recovery is efficient and simple as the recovered energy (80-90%) is available in the form of steam. In some applications the product quality is significantly improved. Examples presented in this paper: Bio-Combine for pellets production: Through integration of the Exergy Steam Dryer for wood with a combined heat and power (CHP) plant, together with HP steam turbine, the excess carrier steam can be utilised for district heating and/or electrical power production in a condensing turbine. Bio-ethanol production: Both for first and second generation of ethanol can the Exergy process be integrated for treatment of raw material and by-products. Exergy Steam Dryer can dry the distillers dark grains and solubles (DDGS), wood, bagasse and lignin. Bio-diesel production: Oil containing seeds and fruits can be treated in order to improve both the quality of oil and animal feed protein, thus minimizing further oil processing costs and increasing the sales revenues. Sewage sludge as bio-mass: Municipal sewage sludge can be considered as a renewable bio-fuel. By drying and incineration, the combustion heat value of the sludge is sufficient for the drying process, generation of electrical energy and production of district heat. Keywords; Exergy, bio-fuel, bio-mass, pellets, bio-ethanol, biodiesel, bio

Questions raised in developing fast reactor steam generator designs

Energy Technology Data Exchange (ETDEWEB)

Taylor, P A; Hayden, O

1975-07-01

The most important component in the achievement of satisfactory LMFBR reliability is the steam generator. When the failure statistics of other nuclear steam generators and the implications of a sodium water reaction are considered, there is some cause for concern. It is apparent that considerable improvement in technology is necessary and until more experience on operating plant is available a conservative design approach must be taken. Many solutions have been proposed, varying from forced circulation straight tube modular to large single vessel once through helical designs. The paper poses what are considered to be the main questions which arise when making a choice of fast reactor steam generator type and tube configuration. The aim is to promote discussion amongst the assembled experts on their relative design approaches and the importance placed upon the various factors in reaching our common goal of ensuring the success of the LMFBR in its essential role of conserving world energy resources. (author)

Questions raised in developing fast reactor steam generator designs

International Nuclear Information System (INIS)

Taylor, P.A.; Hayden, O.

1975-01-01

The most important component in the achievement of satisfactory LMFBR reliability is the steam generator. When the failure statistics of other nuclear steam generators and the implications of a sodium water reaction are considered, there is some cause for concern. It is apparent that considerable improvement in technology is necessary and until more experience on operating plant is available a conservative design approach must be taken. Many solutions have been proposed, varying from forced circulation straight tube modular to large single vessel once through helical designs. The paper poses what are considered to be the main questions which arise when making a choice of fast reactor steam generator type and tube configuration. The aim is to promote discussion amongst the assembled experts on their relative design approaches and the importance placed upon the various factors in reaching our common goal of ensuring the success of the LMFBR in its essential role of conserving world energy resources. (author)

Functionality of ovalbumin during Chinese steamed bread-making processing.

Science.gov (United States)

Sang, Shangyuan; Zhang, Huang; Xu, Lei; Chen, Yisheng; Xu, Xueming; Jin, Zhengyu; Yang, Na; Wu, Fengfeng; Li, Dandan

2018-07-01

Hen egg is commonly used in some cereal-based food, including cakes and bread. Ovalbumin, one of the major components of egg white protein, can affect the performance of the food product. The interaction between ovalbumin and gluten protein and its effect on property of dough and quality of Chinese steamed bread was investigated in this study. Sodium dodecyl sulfate-polyacrylamide gel electrophoresis (SDS-PAGE) patterns indicated that ovalbumin was surprisingly not incorporated in glutenins by covalent bond, whereas size-exclusion high-performance liquid chromatography showed that glutenin macropolymer content in glutenins increased slightly. Furthermore, dynamic rheology experiments indicated ovalbumin led to a decrease inG' andG″ of dough. Based on molecular dynamic simulation and SDS-PAGE results, it was inferred that ovalbumin was not hydrolyzed by endopeptidases during dough fermentation and crosslinked to gluten proteins during steaming. Finally, ovalbumin improved maximum dough height (Hm) during dough development and specific volume of Chinese steamed bread. Copyright © 2018 Elsevier Ltd. All rights reserved.

Application of new designed butterfly type intermediate valve for nuclear steam turbine

International Nuclear Information System (INIS)

Matsumura, Kazuhiro; Kawamata, Susumu; Fujita, Isao; Taketomo, Seiki.

1991-01-01

To cope with a large capacity nuclear steam turbine, a butterfly type intermediate valve has been developed. Compared to the conventional valve, or globe valve, the butterfly valve has the following design features: a) Higher thermal efficiency due to lower pressure loss, b) Easier maintenance due to simplified construction, and c) Lower station cost due to the smaller size of the valve assembly. An experiment with a scaled-down test valve was carried out using compressed air. Subsequently a full-scale valve was tested using steam under actual steam conditions. As a result, these tests gave us no problems. The first nuclear turbine (1100MW) equipped with a butterfly valve is operating satisfactorily with good performance as expected. (author)

Steam jet ejectors are examined automatically

International Nuclear Information System (INIS)

Lardiere, C.

2013-01-01

Steam jet ejectors are used in the nuclear industry particularly for the transfer of radioactive fluids. Their working is based on the Venturi effect and the conservation of energy. A steam ejector can be considered as a thermodynamical pump without mobile parts. The Descote enterprise manufactures a broad range of steam jet ejectors and the characterization and testing of the steam ejectors was made manually and empirically so far. A new test bench has been designed, the tests are led automatically and allow a more accurate characterization and optimization of the steam jet ejectors. (A.C.)

Steam generator tube integrity program

International Nuclear Information System (INIS)

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

1996-01-01

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

Derivation of guidelines for uranium residual radioactive material in soil at the New Brunswick Site, Middlesex County, New Jersey

International Nuclear Information System (INIS)

Dunning, D.; Kamboj, S.; Nimmagadda, M.; Yu, C.

1996-02-01

Residual radioactive material guidelines for uranium in soil were derived for the New Brunswick Site, located in Middlesex County, New Jersey. This site has been designated for remedial action under the Formerly Utilized Sites Remedial Action Program of the US Department of Energy (DOE). Residual radioactive material guidelines for individual radionuclides of concern and total uranium were derived on the basis of the requirement that the 50-year committed effective dose equivalent to a hypothetical individual who lives or works in the immediate vicinity of the New Brunswick Site should not exceed a dose of 30 mrem/yr following remedial action for the current-use and likely future-use scenarios or a dose of 100 mrem/yr for less likely future-use scenarios. The DOE residual radioactive material guideline computer code, RESRAD, was used in this evaluation; RESRAD implements the methodology described in the DOE manual for establishing residual radioactive material guidelines. The guidelines derived in this report are intended to apply to the remediation of these remaining residual radioactive materials at the site. The primary radionuclides of concern in these remaining materials are expected to be radium-226 and, to a lesser extent, natural uranium and thorium. The DOE has established generic cleanup guidelines for radium and thorium in soil; however, cleanup guidelines for other radionuclides must be derived on a site-specific basis

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

Steam turbines for nuclear power plants

International Nuclear Information System (INIS)

Kosyak, Yu.F.

1978-01-01

Considered are the peculiarities of the design and operation of steam turbines, condensers and supplementary equipment of steam turbines for nuclear power plants; described are the processes of steam flow in humid-steam turbines, calculation and selection principles of main parameters of heat lines. Designs of the turbines installed at the Charkov turbine plant are described in detail as well as of those developed by leading foreign turbobuilding firms

In-Space Propulsion, Logistics Reduction, and Evaluation of Steam Reformer Kinetics: Problems and Prospects

Science.gov (United States)

Jaworske, D. A.; Palaszewski, B. A.; Kulis, M. J.; Gokoglu, S. A.

2015-01-01

Human space missions generate waste materials. A 70-kg crewmember creates a waste stream of 1 kg per day, and a four-person crew on a deep space habitat for a 400+ day mission would create over 1600 kg of waste. Converted into methane, the carbon could be used as a fuel for propulsion or power. The NASA Advanced Exploration Systems (AES) Logistics Reduction and Repurposing (LRR) project is investing in space resource utilization with an emphasis on repurposing logistics materials for useful purposes and has selected steam reforming among many different competitive processes as the preferred method for repurposing organic waste into methane. Already demonstrated at the relevant processing rate of 5.4 kg of waste per day, high temperature oxygenated steam consumes waste and produces carbon dioxide, carbon monoxide, and hydrogen which can then be converted into methane catalytically. However, the steam reforming process has not been studied in microgravity. Data are critically needed to understand the mechanisms that allow use of steam reforming in a reduced gravity environment. This paper reviews the relevant literature, identifies gravity-dependent mechanisms within the steam gasification process, and describes an innovative experiment to acquire the crucial kinetic information in a small-scale reactor specifically designed to operate within the requirements of a reduced gravity aircraft flight. The experiment will determine if the steam reformer process is mass-transport limited, and if so, what level of forced convection will be needed to obtain performance comparable to that in 1-g.

Surface area-burnoff correlation for the steam--graphite reaction

International Nuclear Information System (INIS)

Stark, W.A. Jr.; Malinauskas, A.P.

1977-01-01

The oxidation of core graphite by steam of air represents a problem area of significant concern in safety analyses for the high temperature gas cooled reactor (HTGR). Core and core-support graphite integrity and strength deteriorate with oxidation of the graphite, and oxidation furthermore could affect the rate of fission product release under upset conditions. Consequently, modeling of core response during steam or air ingress conditions requires an expression for the rate of graphite interaction with those impurities. The steam--graphite reaction in particular is a complex interaction of mass transport within the graphite with chemi-sorption and reaction on accessible surfaces; experimental results from graphite to graphite are highly variable, and the description of the reaction is not yet completely consistent. A simple etch pit model relating surface area to burnoff has been proposed and shown to provide reasonable correlation with experimental data obtained from steam oxidation studies of nuclear grade H-327 graphite. Unaccounted differences between theory and experiment arise at burnoffs exceeding 3 to 5 percent. The model, while not complete nor comprehensive, is consistent with experimental observations of graphite oxidation by O 2 (air), CO 2 , or H 2 O, and could have some utility in safety analysis

Tests and analysis on steam generator tube failure propagation

International Nuclear Information System (INIS)

Tanabe, Hiromi

1990-01-01

The understanding of leak enlargement and failure propagation behavior is essential to select a design basis leak (DBL) of LMFBR steam generators. Therefore, various series of experiments, such as self-enlargement tests, target wastage tests, failure propagation tests were conducted in a wide range of leak using test facilities of SWAT at PNC/OEC. Especially, in the large leak tests, potential of overheating failure was investigated under a prototypical steam cooling condition inside target tubes. In the small leak, the difference of wastage resistivity was clarified among several tube materials such as 9-chrome steels. In regard to an analytical approach, a computer code LEAP (Leak Enlargement and Propagation) was developed on the basis of all of these experimental results. The code was used to validate the previously selected DBL of the prototype reactor, Monju, steam generator. This approach proved to be successful in spite of somewhat over-conservatism in the analysis. Moreover, LEAP clarified the effectiveness of a rapid steam dump and an enhanced leak detection system. The code improvement toward a realistic analysis is desired, however, to lessen the DBL for a future large plant and then the re-evaluation of the experimental data such as the size of secondary failure is under way. (author). 4 refs, 8 figs, 1 tab

Safety Picks up "STEAM"

Science.gov (United States)

Roy, Ken

2016-01-01

This column shares safety information for the classroom. STEAM subjects--science, technology, engineering, art, and mathematics--are essential for fostering students' 21st-century skills. STEAM promotes critical-thinking skills, including analysis, assessment, categorization, classification, interpretation, justification, and prediction, and are…

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

Steam jet mill-a prospective solution to industrial exhaust steam and solid waste.

Science.gov (United States)

Zhang, Mingxing; Chen, Haiyan

2018-04-20

Bulk industrial solid wastes occupy a lot of our resources and release large amounts of toxic and hazardous substances to the surrounding environment, demanding innovative strategies for grinding, classification, collection, and recycling for economically ultrafine powder. A new technology for grinding, classification, collection, and recycling solid waste is proposed, using the superheated steam produced from the industrial exhaust steam to disperse, grind, classify, and collect the industrial solid waste. A large-scale steam jet mill was designed to operate at an inlet steam temperature 230-300 °C and an inlet pressure of 0.2-0.6 MPa. A kind of industrial solid waste fluidized-bed combustion ashes was used to grinding tests at different steam temperatures and inlet pressures. The total process for grinding, classification, and collection is drying. Two kinds of particle sizes are obtained. One particle size is d 50  = 4.785 μm, and another particle size is d 50  = 8.999 μm. For particle size d 50  = 8.999 μm, the inlet temperature is 296 °C and an inlet pressure is 0.54 MPa for the grinding chamber. The steam flow is 21.7 t/h. The yield of superfine powder is 73 t/h. The power consumption is 3.76 kW h/t. The obtained superfine powder meets the national standard S95 slag. On the basis of these results, a reproducible and sustainable industrial ecological protocol using steam produced by industrial exhaust heat coupled to solid waste recycling is proposed, providing an efficient, large-scale, low-cost, promising, and green method for both solid waste recovery and industrial exhaust heat reutilization.

Proceedings of the third international steam generator and heat exchanger conference

International Nuclear Information System (INIS)

1998-01-01

The Third International Steam Generator and Heat Exchanger conference had the objective to present the state of knowledge of steam generator performance and life management, and also heat exchanger technology. As this conference followed on from the previous conferences held in Toronto in 1990 and 1994, the emphasis was on recent developments, particularly those of the last 4 years. The conference provided an opportunity to operators, designers and researchers in the field of steam generation associated with electricity generation by nuclear energy to present their findings and exchange ideas. The conference endeavoured to do this over the widest possible range of subject areas, including: general operating experience, life management and fitness for service strategies, maintenance and inspection, thermalhydraulics, vibration, fretting and fatigue, materials, chemistry and corrosion and the regulatory issues

Proceedings of the third international steam generator and heat exchanger conference

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-07-01

The Third International Steam Generator and Heat Exchanger conference had the objective to present the state of knowledge of steam generator performance and life management, and also heat exchanger technology. As this conference followed on from the previous conferences held in Toronto in 1990 and 1994, the emphasis was on recent developments, particularly those of the last 4 years. The conference provided an opportunity to operators, designers and researchers in the field of steam generation associated with electricity generation by nuclear energy to present their findings and exchange ideas. The conference endeavoured to do this over the widest possible range of subject areas,including: general operating experience, life management and fitness for service strategies, maintenance and inspection, thermalhydraulics, vibration, fretting and fatigue, materials, chemistry and corrosion and the regulatory issues.

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...

Specialists meeting on leak detection and location in LMFBR steam generators. Summary report

Energy Technology Data Exchange (ETDEWEB)

NONE

1978-10-01

The following topics covered at the meeting were: with leak detection and location methods and equipment, including concentration measurements, helium tests, and electromagnetic methods; acoustic leak detection and related equipment; techniques and experiences in ensuring and measuring steam generator tightness during manufacturing, installation and repair, tube inspection methods for periodic control and damage assessment following leaks, influence of these methods on design of steam generators for LMFBR type reactors.

Specialists meeting on leak detection and location in LMFBR steam generators. Summary report

International Nuclear Information System (INIS)

1978-10-01

The following topics covered at the meeting were: with leak detection and location methods and equipment, including concentration measurements, helium tests, and electromagnetic methods; acoustic leak detection and related equipment; techniques and experiences in ensuring and measuring steam generator tightness during manufacturing, installation and repair, tube inspection methods for periodic control and damage assessment following leaks, influence of these methods on design of steam generators for LMFBR type reactors

Signal analysis of steam line acoustics

International Nuclear Information System (INIS)

Martin, C. Samuel

2003-01-01

The vibration of nuclear steam piping is usually associated with pressure fluctuations emanating from flow disturbances such as steam generator nozzles, bends, or other pipe fittings. Flow separation at pipe tees and within steam chest manifolds or headers generate pressure fluctuations that propagate both upstream to steam generators as well as downstream to the steam turbine. Steady-state acoustic oscillations at various frequencies occur within the piping, possibly exciting structural vibrations. This paper focuses on the assessment of the origin of the disturbances using signal analyses of two dynamic pressure recordings from pressure transducers located along straight runs in the steam piping. The technique involves performing the cross spectrum to two dynamic pressure signals in piping between (1) the steam generator and steam chest header, and (2) between the header and steam turbine outlet. If, at a specified frequency, no causality occurs between the two signals then the cross spectra magnitude will be negligible. Of interest here is the value of the phase between the two signals for frequencies for which the magnitude of the cross spectrum is not negligible. It is shown in the paper that the direction of the dominant waves at all frequencies can be related to the phase angle from the cross spectrum. It has to be realized that pressure waves emanating from one source such as a steam generator will propagate along uniform steam pipes with little transformation or attenuation, but will be reflected at fittings and at inlets and outlets. Hence, the eventual steady-state time record at a given location in the piping is a result of not only the disturbance, but also reflections of earlier pulsations. Cross-spectral analyses has been employed to determine the direction of the dominant acoustic waves in the piping for various frequencies for which there are signals. To prove the technique, synthetic spectra are generated comprised of harmonic waves moving both

Steam-water separator

International Nuclear Information System (INIS)

Modrak, T.M.; Curtis, R.W.

1978-01-01

The steam-water separator connected downstream of a steam generator consists of a vertical centrifugal separator with swirl blades between two concentric pipes and a cyclone separator located above. The water separated in the cyclone separator is collected in the inner tube of the centrifugal separator which is closed at the bottom. This design allows the overall height of the separator to be reduced. (DG) [de

Turbulence production by a steam-driven jet in a water vessel

Energy Technology Data Exchange (ETDEWEB)

Wissen, R.J.E. van; Schreel, K.R.A.M.; Geld, C.W.M. van der [Eindhoven Univ. of Technology (Netherlands). Dept. of Mechanical Engineering; Wieringa, J. [Unilever Research and Development, Vlaardingen (Netherlands)

2004-04-01

Direct steam injection is an efficient means of heating a volume of liquid. Usually the steam is injected via a nozzle, yielding a strong jet that condenses rapidly and transforms into a self-similar single phase jet. In the experiments reported in this paper, superheated steam is injected, centrally, at the bottom of a vertical, cylindrical water vessel. The resulting jet is turbulent (Re=7.9 x 10{sup 4}-18.1 x 10{sup 4} with the length scale based on the width of the jet, r{sub 1/2} and the velocity scale based on the centerline velocity, U{sub 0}). Using PIV in a vertical plane through the central axis, instantaneous velocity fields have been measured at a rate of 15 Hz. Near the inlet, the jet is mainly steam that rapidly condenses. Further downstream, the jet is essentially single phase, although some residual air is present as microscopically small bubbles. In the area directly downstream of the steam part, the ratio of r{sub 1/2} to the vessel radius R (32.5 cm) is about 1/14. The production of turbulent kinetic energy has been quantified for the main process conditions. Its dependencies on temperature, nozzle opening and inlet steam pressure have been determined. This production of energy is related to the stresses exerted on small particles in the mixture, and break-up of particles is discussed. (author)

Improvement of degradation with non-condensable gas in micro steam injector

International Nuclear Information System (INIS)

Saihara, Atsushi; Horiki, Sachiyo; Osakabe, Masahiro; Ohmori, Shuichi

2007-01-01

Effect of non-condensable gas on a micro steam injector (MSI) to obtain a vacuum was experimentally studied. When a pure steam was used in the MSI, the high vacuum condition was obtained. However when the mass fraction of air included in the steam was larger than a cartain value, the MSI became unstable and the vacuum condition could not be obtained. It is considered that the malfunction is due to the instability triggered with the uncondensed steam remained at the throat in downstream of the condensing region. The water nozzle was expected to be a key component to mitigate the effect of non-condensable gas. Three kinds of water nozzle whose flow areas were round, star and screw shapes were used in the present experiment. The star-shaped nozzle where the increased surface area could be expected to compensate the degradation of condensation failed to improve the malfunction of MSI with the non-condensable gas. The screw nozzle expected to drive air away outside the condensing surface could mitigate the effect of non-condensable gas. (author)

Erosion corrosion in wet steam

International Nuclear Information System (INIS)

Tavast, J.

1988-03-01

The effect of different remedies against erosion corrosion in wet steam has been studied in Barsebaeck 1. Accessible steam systems were inspected in 1984, 1985 and 1986. The effect of hydrogen peroxide injection of the transport of corrosion products in the condensate and feed water systems has also been followed through chemical analyses. The most important results of the project are: - Low alloy chromium steels with a chromium content of 1-2% have shown excellent resistance to erosion corrosion in wet steam. - A thermally sprayed coating has shown good resistance to erosion corrosion in wet steam. In a few areas with restricted accessibility minor attacks have been found. A thermally sprayed aluminium oxide coating has given poor results. - Large areas in the moisture separator/reheater and in steam extraction no. 3 have been passivated by injection of 20 ppb hydrogen peroxide to the high pressure steam. In other inspected systems no significant effect was found. Measurements of the wall thickness in steam extraction no. 3 showed a reduced rate of attack. - The injection of 20 ppb hydrogen peroxide has not resulted in any significant reduction of the iron level result is contrary to that of earlier tests. An increase to 40 ppb resulted in a slight decrease of the iron level. - None of the feared disadvantages with hydrogen peroxide injection has been observed. The chromium and cobalt levels did not increase during the injection. Neither did the lifetime of the precoat condensate filters decrease. (author)

HCDA bubble experiment, (2)

International Nuclear Information System (INIS)

Sakata, Kaoru; Mashiko, Hiroyuki; Oka, Yoshiaki; An, Shigehiro; Isozaki, Tadashi.

1981-06-01

An experiment simulating the behavior of the very large steam bubbles generated at the time of an accident of core collapse was carried out with a warm water tank, and the applicability of the theory of very small bubble disappearance known at present was examined. The bubbles generated in HCDA (hypothetical core disruptive accident) are expected to be very large, containing sodium, fuel, FP gas and so on, and play important role in the mechanism of emitting radioactive substances in the safety analysis of LMFBRs. In this experiment, the degree of subcool of the warm water pool, the initial radii of steam bubbles and the blowoff pressure of steam were taken as the parameters. The radius of the steam bubbles generated in the experiment was about 6.5 cm, and the state of disappearance was different above and below the degree of unsaturation of 10 deg C. Comparing the disappearance curve obtained by the experiment with the theory of disappearance of small bubbles, the experimental values were between inertia-controlled disappearance and heat transfer-controlled disappearance, and this result was able to be explained generally with the model taking the pressure change within steam bubbles into account. The rise of bubbles was also observed. (Kako, I.)

Steam-generator replacement sets new marks

International Nuclear Information System (INIS)

Beck, R.L.

1995-01-01

This article describes how, in one of the most successful steam-generator replacement experiences at PWRs worldwide, the V C Summer retrofit exceeded plant goals for critical-path duration, radiation, exposure, and radwaste generation. Intensive planning and teamwork, combined with the firm support of station management and the use of mockups to prepare the work crews for activity in a radiological environment, were key factors in the record performance achieved by South Carolina Electric and Gas Co (SCE and G) in replacing three steam generators at V C Summer nuclear station. The 97-day, two-hour breaker-to-breaker replacement outage -- including an eight-day delay for repair of leak in a small-bore seal-injection line of a reactor coolant pump (unrelated to the replacement activities) -- surpassed the project goal by over one day. Moreover, the outage was only 13 hours shy of the world record held by Virginia Power Co's North Anna Unit 1

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

Thermal analysis of cylindrical natural-gas steam reformer for 5 kW PEMFC

Science.gov (United States)

Jo, Taehyun; Han, Junhee; Koo, Bonchan; Lee, Dohyung

2016-11-01

The thermal characteristics of a natural-gas based cylindrical steam reformer coupled with a combustor are investigated for the use with a 5 kW polymer electrolyte membrane fuel cell. A reactor unit equipped with nickel-based catalysts was designed to activate the steam reforming reaction without the inclusion of high-temperature shift and low-temperature shift processes. Reactor temperature distribution and its overall thermal efficiency depend on various inlet conditions such as the equivalence ratio, the steam to carbon ratio (SCR), and the fuel distribution ratio (FDR) into the reactor and the combustor components. These experiments attempted to analyze the reformer's thermal and chemical properties through quantitative evaluation of product composition and heat exchange between the combustor and the reactor. FDR is critical factor in determining the overall performance as unbalanced fuel injection into the reactor and the combustor deteriorates overall thermal efficiency. Local temperature distribution also influences greatly on the fuel conversion rate and thermal efficiency. For the experiments, the operation conditions were set as SCR was in range of 2.5-4.0 and FDR was in 0.4-0.7 along with equivalence ratio of 0.9-1.1; optimum results were observed for FDR of 0.63 and SCR of 3.0 in the cylindrical steam reformer.

A drier unit for steam separators

International Nuclear Information System (INIS)

Peyrelongue, J.-P.

1973-01-01

Description is given of a drier unit adapted to equip a water separator mounted in a unit for treating a wet steam fed from a high pressure enclosure, so as to dry and contingently superheat said steam prior to injecting same into a turbine low pressure stage. This drier unit is constituted by at least a stack of separating sheets maintained in parallel relationship and at a slight angle with respect to the horizontal so as to allow the water provided by wet steam to flow toward a channel communicating with a manifold, and by means for guiding the steam between the sheets and evenly distributing it. This can be applied to steam turbines in nuclear power stations [fr

CANDU steam generator life management: laboratory data and plant experience

International Nuclear Information System (INIS)

Tapping, R.L.; Nickerson, J.H.; Subash, N.; Wright, M.D.

2001-10-01

As CANDU reactors enter middle age, and the potential value of the plants in a deregulated market is realized, life management and life extension issues become increasingly important. An accurate assessment of critical components, such as the CANDU 6 steam generators (SGs), is crucial for successful life extension, and in this context, material issues are a key factor. For example, service experience with Alloy 900 tubing indicates very low levels of degradation within CANDU SGs; the same is also noted worldwide. With little field data for extrapolation, life management and life extension decisions for the tube bundles rely heavily on laboratory data. Similarly, other components of the SGs, in particular the secondary side internals, have only limited inspection data upon which to base a condition assessment. However, in this case there are also relatively little laboratory data. Decisions on life management and life extension are further complicated--not only is inspection access often restricted, but repair or replacement options for internal components are, by definition, also limited. The application of CANDU SG life management and life extension requires a judicious blend of in-service data, laboratory research and development (R and D) and materials and engineering judgment. For instance, the available laboratory corrosion and fretting wear data for Alloy 800 SG tubing have been compared with plant experience (with all types of tubing), and with crevice chemistry simulations, in order to provide an appropriate inspection guide for a 50-year SG life. A similar approach has been taken with other SG components, where the emphasis has been on known degradation mechanisms worldwide. This paper provides an outline of the CANDU SG life management program, including the results to date, a summary of the supporting R and D program showing the integration with condition assessment and life management activities, and the approach taken to life extension for a typical

Steam explosion studies review

International Nuclear Information System (INIS)

Hwang, Moon Kyu; Kim, Hee Dong

1999-03-01

When a cold liquid is brought into contact with a molten material with a temperature significantly higher than the liquid boiling point, an explosive interaction due to sudden fragmentation of the melt and rapid evaporation of the liquid may take place. This phenomenon is referred to as a steam explosion or vapor explosion. Depending upon the amount of the melt and the liquid involved, the mechanical energy released during a vapor explosion can be large enough to cause serious destruction. In hypothetical severe accidents which involve fuel melt down, subsequent interactions between the molten fuel and coolant may cause steam explosion. This process has been studied by many investigators in an effort to assess the likelihood of containment failure which leads to large scale release of radioactive materials to the environment. In an effort to understand the phenomenology of steam explosion, extensive studies has been performed so far. The report presents both experimental and analytical studies on steam explosion. As for the experimental studies, both small scale tests which involve usually less than 20 g of high temperature melt and medium/large scale tests which more than 1 kg of melt is used are reviewed. For the modelling part of steam explosions, mechanistic modelling as well as thermodynamic modelling is reviewed. (author)

Water regime of steam power plants

International Nuclear Information System (INIS)

Oesz, Janos

2011-01-01

The water regime of water-steam thermal power plants (secondary side of pressurized water reactors (PWR); fossil-fired thermal power plants - referred to as steam power plants) has changed in the past 30 years, due to a shift from water chemistry to water regime approach. The article summarizes measures (that have been realised by chemists of NPP Paks) on which the secondary side of NPP Paks has become a high purity water-steam power plant and by which the water chemistry stress corrosion risk of heat transfer tubes in the VVER-440 steam generators was minimized. The measures can also be applied to the water regime of fossil-fired thermal power plants with super- and subcritical steam pressure. Based on the reliability analogue of PWR steam generators, water regime can be defined as the harmony of construction, material(s) and water chemistry, which needs to be provided in not only the steam generators (boiler) but in each heat exchanger of steam power plant: - Construction determines the processes of flow, heat and mass transfer and their local inequalities; - Material(s) determines the minimal rate of general corrosion and the sensitivity for local corrosion damage; - Water chemistry influences the general corrosion of material(s) and the corrosion products transport, as well as the formation of local corrosion environment. (orig.)

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)

CATALYTIC STEAM REFORMING OF CHLOROCARBONS: METHLYCHLORIDE. (R822721C633)

Science.gov (United States)

The effective destruction of trichloroethane, trichloroethylene and perchloroethylene by steam reforming with a commercial nickel catalyst has been demonstrated. Conversion levels of up to 0.99999 were attained in both laboratory and semi-pilot experiments, with the products c...

Laboratory experiment demonstrating the way in which a steam barrier prevents the dissolution of salt buried in a flooded packed bed

International Nuclear Information System (INIS)

Taylor, R.W.; Bowen, D.

1977-01-01

We have conducted a laboratory experiment to demonstrate a way in which a solid material can be prevented from dissolving in water. The differential solubility of salt (NaCl) in steam vs water is exploited. As long as the temperature of the area and water surrounding the salt is maintained above the boiling point of water, the salt cannot dissolve. This phenomenon, known as the thermal barrier, has far-reaching implications for preventing the dispersal of contaminants present near groundwater sources

Experiences from Loviisa Nuclear Power Station concerning the decontamination of steam generators and primary system components

International Nuclear Information System (INIS)

Jaernstroem, R.

1989-01-01

Loviisa 1 and 2 are 465 MWe PWR units of the Soviet type VVER-440. Loviisa 1 has been in commercial operation since spring 1977 and Loviisa 2 from the beginning of 1980. Decontamination of primary circuit components - even big ones as steam generators - can be performed in an efficient and quick way with good results and resonable expences. Total costs for decontamination of the two steam generators including planning, construction, documentation, operation, chemicals etc. did not rise above 100,000.00 dollars. (author) 6 figs., 2 tabs

Future development of large steam turbines

International Nuclear Information System (INIS)

Chevance, A.

1975-01-01

An attempt is made to forecast the future of the large steam turbines till 1985. Three parameters affect the development of large turbines: 1) unit output; and a 2000 to 2500MW output may be scheduled; 2) steam quality: and two steam qualities may be considered: medium pressure saturated or slightly overheated steam (light water, heavy water); light enthalpie drop, high pressure steam, high temperature; high enthalpic drop; and 3) the quality of cooling supply. The largest range to be considered might be: open system cooling for sea-sites; humid tower cooling and dry tower cooling. Bi-fluid cooling cycles should be also mentioned. From the study of these influencing factors, it appears that the constructor, for an output of about 2500MW should have at his disposal the followings: two construction technologies for inlet parts and for high and intermediate pressure parts corresponding to both steam qualities; exhaust sections suitable for the different qualities of cooling supply. The two construction technologies with the two steam qualities already exist and involve no major developments. But, the exhaust section sets the question of rotational speed [fr

Vapor generator steam drum spray heat

International Nuclear Information System (INIS)

Fasnacht, F.A. Jr.

1978-01-01

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

The new equation of steam quality and the evaluation of nonradioactive tracer method in PWR steam generators

International Nuclear Information System (INIS)

Ki Bang, Sung; Young Jin, Chang

2001-01-01

The performance of steam turbines is tested as ANSI/ASME-PTC 6. This code provides rules for the accurate testing of steam turbines for the purpose of obtaining the level of performance with a minimum uncertainty. Only the relevant portion of this code needs to process any individual case, In some case the procedure is simple. However, in complex turbines or complex operation modes, more procedures are required to test the involved provisions. Anyway, to measure the steam quality in the Wolsong PHWR with 4 SGs in Korea by the methods in the section ''Measure of steam quality methods'' of ANSI/ASME PTC 6, the result was not good though the steam generators are efficient. So, the new testing method was developed and the sophisticated equation of steam quality was introduced and uses the nonradioactive chemical tracer, Lithium hydroxide(LiOH) instead of the radioactive tracer, Na-24. (author)

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

High-temperature oxidation of Zircaloy in hydrogen-steam mixtures

International Nuclear Information System (INIS)

Chung, H.M.; Thomas, G.R.

1982-09-01

Oxidation rates of Zircaloy-4 cladding tubes have been measured in hydrogen-steam mixtures at 1200 to 1700 0 C. For a given isothermal oxidation temperature, the oxide layer thicknesses have been measured as a function of time, steam supply rate, and hydrogen overpressure. The oxidation rates in the mixtures were compared with similar data obtained in pure steam and helium-steam environments under otherwise identical conditions. The rates in pure steam and helium-steam mixtures were equivalent and comparable to the parabolic rates obtained under steam-saturated conditions and reported in the literature. However, when the helium was replaced with hydrogen of equivalent partial pressure, a significantly smaller oxidation rate was observed. For high steam-supply rates, the oxidation kinetics in a hydrogen-steam mixture were parabolic, but the rate was smaller than for pure steam or helium-steam mixtures. Under otherwise identical conditions, the ratio of the parabolic rate for hydrogen-steam to that for pure steam decreased with increasing temperature and decreasing steam-supply rate

Point Lepreau G.S. commissioning management experience

International Nuclear Information System (INIS)

Alikhan, S.; Walker, W.J.

1983-05-01

This paper shares the experience gained by New Brunswick Power in commissioning its first nuclear plant, with particular emphasis on the management techniques that were developed. Some of the areas discussed include: organization of Nuclear Operations Group responsible for commissioning and its interface with construction; outline and scope of documentation required to define, execute and report commissioning to satisfy the licensing requirements, management techniques developed to implement commissioning; and a summary of lessons learned and recommendations for the future

Improving Steam System Performance: A Sourcebook for Industry

Energy Technology Data Exchange (ETDEWEB)

2002-06-01

The sourcebook is a reference for industrial steam system users, outlining opportunities to improve steam system performance. This Sourcebook is designed to provide steam system users with a reference that describes the basic steam system components, outlines opportunities for energy and performance improvements, and discusses the benefits of a systems approach in identifying and implementing these improvement opportunities. The Sourcebook is divided into the following three main sections: Section 1: Steam System Basics--For users unfamiliar with the basics of steam systems, or for users seeking a refresher, a brief discussion of the terms, relationships, and important system design considerations is provided. Users already familiar with industrial steam system operation may want to skip this section. This section describes steam systems using four basic parts: generation, distribution, end use, and recovery. Section 2: Performance Improvement Opportunities--This section discusses important factors that should be considered when industrial facilities seek to improve steam system performance and to lower operating costs. This section also provides an overview of the finance considerations related to steam system improvements. Additionally, this section discusses several resources and tools developed by the U. S. Department of Energy's (DOE) BestPractices Steam Program to identify and assess steam system improvement opportunities. Section 3: Programs, Contacts, and Resources--This section provides a directory of associations and other organizations involved in the steam system marketplace. This section also provides a description of the BestPractices Steam Program, a directory of contacts, and a listing of available resources and tools, such as publications, software, training courses, and videos.

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.)

Steam purity in PWRs

International Nuclear Information System (INIS)

Hopkinson, J.; Passell, T.

1982-01-01

Reports that 2 EPRI studies of PWRs prove that impure steam triggers decay of turbine metals. Reveals that EPRI is attempting to improve steam monitoring and analysis, which are key steps on the way to deciding the most cost-effective degree of steam purity, and to upgrade demineralizing systems, which can then reliably maintain that degree of purity. Points out that 90% of all cracks in turbine disks have occurred at the dry-to-wet transition zone, dubbed the Wilson line. Explains that because even very clean water contains traces of chemical impurities with concentrations in the parts-per-billion range, Crystal River-3's secondary loop was designed with even more purification capability; a deaerator to remove oxygen and prevent oxidation of system metals, and full-flow resin beds to demineralize 100% of the secondary-loop water from the condenser. Concludes that focusing attention on steam and water chemistry can ward off cracking and sludge problems caused by corrosion

Report--Training Session on Roles and Responsibilities and Steps in Negotiations for the Board of Directors, New Brunswick Indian Arts and Crafts Corporation. No. 166.

Science.gov (United States)

Goddu, Roland; Nicholas, Darryl

The Board of Directors of the New Brunswick Indian Arts and Crafts Corporation attended a two-day training session in Presque Isle, Maine, to discuss and clarify roles and responsibilities of the various agencies and position holders of the Provincial and Federal Corporations. In addition, an extensive discussion of negotiations procedures took…

Sodium/water reactions in steam generators of liquid metal fast breeder reactors

International Nuclear Information System (INIS)

Hori, M.

1980-01-01

The status of the research and development on sodium/water reactions resulting from the leakage of water into sodium in LMFBR steam generators is reviewed. The importance of sodium/water reaction phenomena in the design and operation of steam generators is discussed. The effects of sodium/water reactions are evaluated and methods of protection against these phenomena are surveyed. The products of chemical reactions between sodium and water under steam generator conditions are H 2 , NaOH, Na 2 O and NaH. Together with the temperature rise due to the associated exothermic reaction, these reaction products cause effects such as self-wastage, single- and multi-target wastage, and rapid pressure increase, depending on the size of the leak hole or the magnitude of leak rate. As for the wastage phenomena of small leaks, the effects of various factors have been studied and experimental correlations, as well as some theoretical work, have been performed. To investigate the pressure phenomena of a large leak, large-scale tests have been conducted by various organizations, and the computer codes to analyse these phenomena have been developed and verified by experiments. In the design of steam generators, an initial failure up to a hypothetical double-ended guillotine rupture of a single heat transfer tube is widely used as the design basis leak. Protection systems for LMFBR plants consist of leak detection devices, leak termination devices, and reaction pressure relief devices. From analyses based on research and development activities, as well as from experience with leaks in steam generator test loops and reactor plants, it has been confirmed that protection systems can satisfactorily be designed to accommodate leak incidents in LMFBR plants. (author)

New Staphylinidae (Coleoptera records with new collection data from New Brunswick, Canada. I. Aleocharinae

Directory of Open Access Journals (Sweden)

Reginald Webster

2009-09-01

Full Text Available Forty-eight species of Aleocharinae are newly reported from New Brunswick, bringing the total number of species known from the province to 149. Two of these species, Ocyusa asperula Casey and Myllaena kaskaskia Klimaszewski, are newly recorded for Canada. Additional locality data are presented for nine species recently recorded from the province. Collection and bionomic data for all these species are presented and discussed. Colour habitus images are presented for all species included in this paper and genital images are presented for closely related Myllaena kaskaskia Klimaszewski, M. procidua Casey and M. vulpina Bernhauer. Photographs of the male genitalia of M. procidua are presented for the first time. The female spermatheca, tergite and sternite eight of Amarochara formicina Assing are illustrated for the first time.

Kids Inspire Kids for STEAM

OpenAIRE

Fenyvesi, Kristof; Houghton, Tony; Diego-Mantecón, José Manuel; Crilly, Elizabeth; Oldknow, Adrian; Lavicza, Zsolt; Blanco, Teresa F.

2017-01-01

Abstract The goal of the Kids Inspiring Kids in STEAM (KIKS) project was to raise students' awareness towards the multi- and transdisciplinary connections between the STEAM subjects (Science, Technology, Engineering, Arts & Mathematics), and make the learning about topics and phenomena from these fields more enjoyable. In order to achieve these goals, KIKS project has popularized the STEAM-concept by projects based on the students inspiring other students-approach and by utilizing new tec...

[Effects of steaming and baking on content of alkaloids in Aconite Lateralis Radix (Fuzi)].

Science.gov (United States)

Yang, Chang-lin; Huang, Zhi-fang; Zhang, Yi-han; Liu, Yu-hong; Liu, Yun-huan; Chen, Yan; Yi, Jin-hai

2014-12-01

To study the effect of steaming and baking process on contents of alkaloids in Aconite Lateralis Radix (Fuzi), 13 alkaloids were analyzed by UPLC-MS/MS equipped with ESI ion source in MRM mode. In steaming process, the contents of diester-diterpenoid alkaloids decreased rapidly, the contents of monoester-diterpenoid alkaloids firstly increased, reached the peak at 40 min, and then deceased gradually. The contents of aconine alkaloids (mesaconine, aconine and hypaconine) increased all the time during processing, while the contents of fuziline, songorine, karacoline, salsolionl were stable or slightly decreased. In baking process, dynamic variations of alkaloids were different from that in the steaming process. Diester-diterpenoid alkaloids were degraded slightly slower than in steaming process. Monoester-diterpenoid alkaloids, aconine alkaloids and the total alkaloids had been destroyed at different degrees, their contents were significantly lower than the ones in steaming Fuzi at the same processing time. This experiment revealed the dynamic variations of alkaloids in the course of steaming and baking. Two processing methods which can both effectively remove the toxic ingredients and retain the active ingredients are simple and controllable, and are valuable for popularization and application.

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

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...

Steampunk: Full Steam Ahead

Science.gov (United States)

Campbell, Heather M.

2010-01-01

Steam-powered machines, anachronistic technology, clockwork automatons, gas-filled airships, tentacled monsters, fob watches, and top hats--these are all elements of steampunk. Steampunk is both speculative fiction that imagines technology evolved from steam-powered cogs and gears--instead of from electricity and computers--and a movement that…

Evolution of design of steam generator for sodium cooled reactors

International Nuclear Information System (INIS)

Chetal, S.C.; Vaidyanathan

1997-01-01

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

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.

7 CFR 29.3058 - Steam-dried.

Science.gov (United States)

2010-01-01

... 7 Agriculture 2 2010-01-01 2010-01-01 false Steam-dried. 29.3058 Section 29.3058 Agriculture... Steam-dried. The condition of unfermented tobacco as customarily prepared for storage by means of a redrying machine or other steam-conditioning equipment. [24 FR 8771, Oct. 29, 1959. Redesignated at 47 FR...

Steam atmosphere drying concepts using steam exhaust recompression

Energy Technology Data Exchange (ETDEWEB)

DiBella, F.A. (TECOGEN, Inc., Waltham, MA (United States))

1992-08-01

In the US industrial drying accounts for approximately 1.5 quads of energy use per year. Annual industrial dryer expenditures are estimated to be in the $500 million range. Industrial drying is a significant energy and monetary expense. For the thermal drying processes in which water is removed via evaporation from the feedstock, attempts have been made to reduce the consumption of energy using exhaust waste heat recovery techniques, improved dryer designs, or even the deployment of advanced mechanical dewatering techniques. Despite these efforts, it is obvious that a large amount of thermal energy is often still lost if the latent heat of evaporation from the evaporated water cannot be recovered and/or in some way be utilized as direct heat input into the dryer. Tecogen Inc. is conducting research and development on an industrial drying concept. That utilizes a directly or indirectly superheated steam cycle atmosphere with exhaust steam recompression to recover the latent heat in the exhaust that would otherwise be lost. This approach has the potential to save 55 percent of the energy required by a conventional air dryer. Other advantages to the industrial dryer user include: A 35-percent reduction in the yearly cost per kg[sub evap] to dry wet feedstock, Reduced airborne emissions, Reduced dry dust fire/explosion risks, Hot product not exposed to oxygen thus, the product quality is enhanced, Constant rate drying in steam atmosphere, Reduced dryer size and cost, Reduced dryer heat losses due to lower dryer inlet temperatures. Tecogen has projected that the steam atmosphere drying system is most suitable as a replacement technology for state-of-the-art spray, flash, and fluidized bed drying systems. Such systems are utilized in the food and kindred products; rubber products; chemical and allied products; stone, clay, and glass; textiles; and pulp and paper industrial sectors.

Steam atmosphere drying concepts using steam exhaust recompression

Energy Technology Data Exchange (ETDEWEB)

DiBella, F.A. [TECOGEN, Inc., Waltham, MA (United States)

1992-08-01

In the US industrial drying accounts for approximately 1.5 quads of energy use per year. Annual industrial dryer expenditures are estimated to be in the $500 million range. Industrial drying is a significant energy and monetary expense. For the thermal drying processes in which water is removed via evaporation from the feedstock, attempts have been made to reduce the consumption of energy using exhaust waste heat recovery techniques, improved dryer designs, or even the deployment of advanced mechanical dewatering techniques. Despite these efforts, it is obvious that a large amount of thermal energy is often still lost if the latent heat of evaporation from the evaporated water cannot be recovered and/or in some way be utilized as direct heat input into the dryer. Tecogen Inc. is conducting research and development on an industrial drying concept. That utilizes a directly or indirectly superheated steam cycle atmosphere with exhaust steam recompression to recover the latent heat in the exhaust that would otherwise be lost. This approach has the potential to save 55 percent of the energy required by a conventional air dryer. Other advantages to the industrial dryer user include: A 35-percent reduction in the yearly cost per kg{sub evap} to dry wet feedstock, Reduced airborne emissions, Reduced dry dust fire/explosion risks, Hot product not exposed to oxygen thus, the product quality is enhanced, Constant rate drying in steam atmosphere, Reduced dryer size and cost, Reduced dryer heat losses due to lower dryer inlet temperatures. Tecogen has projected that the steam atmosphere drying system is most suitable as a replacement technology for state-of-the-art spray, flash, and fluidized bed drying systems. Such systems are utilized in the food and kindred products; rubber products; chemical and allied products; stone, clay, and glass; textiles; and pulp and paper industrial sectors.

Cycle improvement for nuclear steam power plant

International Nuclear Information System (INIS)

Silvestri, G.J. Jr.

1976-01-01

A pressure-increasig ejector element is disposed in an extraction line intermediate to a high pressure turbine element and a feedwater heater. The ejector utilizes high pressure fluid from a reheater drain as the motive fluid to increase the pressure at which the extraction steam is introduced into the feedwater heater. The increase in pressure of the extraction steam entering the feedwater heater due to the steam passage through the ejector increases the heat exchange capability of the extraction steam thus increasing the overall steam power plant efficiency

Service experience and stress corrosion of Inconel 600 bellows expansion joints in turbine steam environments

International Nuclear Information System (INIS)

Kramer, L.D.; Michael, S.T.; Pement, F.W.

The purpose of this paper is to discuss the service history of Inconel 600 expansion bellows, to illustrate a typical case of failure, propose S.C.C. mechanisms, and to rationalize the most probable mechanism. Inconel 600 is fully resistant to high-purity power plant steam (720 deg F maximum) for on-going service lifetimes which greatly exceed the incubation periods which are reported or postulated in the literature for delayed stress corrosion cracking in high-purity water tests (630-660 deg F). The only observed stress corrosion environments which are sufficiently rapidly deleterious to be consistent with failure lifetimes are molten NaOH in superheated steam or a very concentrated aqueous caustic solution containing silica contamination. (author)

The Concept of Steam Pressure Control by Changing the Feedwater Flow during Heatup Operation for an Integral Reactor with a Once-Through Steam Generator

International Nuclear Information System (INIS)

Seo, Jae Kwang; Choi, Ki Yong; Kang, Han Ok; Kim, Young In; Yoon, Ju Hyeon; Zee, Sung Qunn

2005-01-01

contrived. Even though a mathematical model for the startup cooling line itself is not complicated, the working principle of the startup cooling line with a fixed flow resistance, in conjunction with the feedwater flow change, should be confirmed. In order to check that the required passage points are passed through under given plant operational conditions, its thermal-hydraulic performance during a heatup operation is analyzed by using the PZRTR code, which is a computer code developed for the analysis of a heatup operation. The feasibility of the concept of the steam pressure control proposed in this study will be discussed with the results from both of a simulation and an experiment

Steam generator tube rupture risk impact on design and operation of French PWR plants

International Nuclear Information System (INIS)

Depond, G.; Sureau, H.

1984-01-01

The experience of steam generator tube leaks incidents in PWR plants has resulted in an increase of EDF analysis leading to improvements in design and post-accidental operation for new projects and operating plants. The accident consequences are minimized for each of the NSSS three barriers: first barrier: safeguard systems design and operating procedures relying upon core safety allow to maintain a low level of primary radioactivity, second barrier: steam generator design and periodic inspection allow to reduce tube ruptures risks and third barrier: atmospheric releases are reduced as a result of optimal recovery procedures, detection improvements and atmospheric steam valves design improvements. (orig.)

Corrosion of several metals in supercritical steam at 5380C

International Nuclear Information System (INIS)

McCoy, H.E.; McNabb, B.

1977-05-01

The corrosion of several iron- and nickel-base alloys in supercritical steam at 24.1 MPa (3500 psi) and 538 0 C was measured to 7.92 x 10 7 s (22,000 h). The experiments were carried out in TVA's Bull Run Steam Plant. Corrosion was measured almost entirely by weight change and visual appearance; a few samples were evaluated by more descriptive analytical techniques. The corrosion rates of low-alloy ferritic steels containing from 1.1 to 8.7 percent Cr and 0.5 to 1.0 percent Mo differed by less than a factor of 2 in steam. Several modified compositions of Hastelloy N were evaluated and found to corrode at about equivalent rates. Of the alloys studied, the lowest weight gain in 3.6 x 10 7 sec (10,000 hr) was 0.01 mg/cm 2 for Inconel 718 and the highest 10 mg/cm 2 for the low-alloy ferritic steels. 25 figures, 3 tables

Amazing & extraordinary facts the steam age

CERN Document Server

Holland, Julian

2012-01-01

Respected transport author Julian Holland delves into the intriguing world of steam in his latest book, which is full of absorbing facts and figures on subjects ranging from Cornish beam engines, steam railway locomotives, road vehicles and ships through to traction engines, steam rollers and electricity generating stations and the people who designed and built them. Helped along the way by the inventive minds of James Watt, Richard Trevithick and George Stephenson, steam became the powerhouse that drove the Industrial Revolution in Britain in the late 18th and 19th centuries.

First principles calculations and experimental insight into methane steam reforming over transition metal catalysts

DEFF Research Database (Denmark)

Jones, Glenn; Jakobsen, Jon Geest; Shim, Signe Sarah

2008-01-01

This paper presents a detailed analysis of the steam reforming process front first-principles calculations, supported by insight from experimental investigations. In the present work we employ recently recognised scaling relationships for adsorption energies of simple molecules adsorbed at pure...... metal Surfaces to develop an overview of the steam reforming process catalyzed by a range of transition metal surfaces. By combining scaling relationships with thermodynamic and kinetic analysis, we show that it is possible to determine the reactivity trends of the pure metals for methane steam...... in situ TEM measurements under a hydrogen atmosphere. The overall agreement between theory and experiment (at 773 K, 1 bar pressure and 10% conversion) is found to be excellent with Ru and Rh being the most active pure transition metals for methane steam reforming, while Ni, Ir, Pt, and Pd...

7 CFR 29.2552 - Steam-dried.

Science.gov (United States)

2010-01-01

... 7 Agriculture 2 2010-01-01 2010-01-01 false Steam-dried. 29.2552 Section 29.2552 Agriculture...-Cured Tobacco (u.s. Types 22, 23, and Foreign Type 96) § 29.2552 Steam-dried. The condition of unfermented tobacco as customarily prepared for storage by means of a redrying machine or other steam...

7 CFR 29.2300 - Steam-dried.

Science.gov (United States)

2010-01-01

... 7 Agriculture 2 2010-01-01 2010-01-01 false Steam-dried. 29.2300 Section 29.2300 Agriculture... INSPECTION Standards Official Standard Grades for Virginia Fire-Cured Tobacco (u.s. Type 21) § 29.2300 Steam... machine or other steam-conditioning equipment. [37 FR 13521, July 11, 1972. Redesignated at 51 FR 40406...

7 CFR 29.3548 - Steam-dried.

Science.gov (United States)

2010-01-01

... 7 Agriculture 2 2010-01-01 2010-01-01 false Steam-dried. 29.3548 Section 29.3548 Agriculture... Type 95) § 29.3548 Steam-dried. The condition of unfermented tobacco as customarily prepared for storage by means of a redrying machine or other steam-conditioning equipment. [30 FR 9207, July 23, 1965...

7 CFR 29.1060 - Steam-dried.

Science.gov (United States)

2010-01-01

... 7 Agriculture 2 2010-01-01 2010-01-01 false Steam-dried. 29.1060 Section 29.1060 Agriculture... Type 92) § 29.1060 Steam-dried. The condition of unfermented tobacco as customarily prepared for storage by means of a redrying machine or other steam-conditioning equipment. [42 FR 21092, Apr. 25, 1977...

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 microturbines in distributed cogeneration

CERN Document Server

Kicinski, Jan

2014-01-01

This book presents the most recent trends and concepts in power engineering, especially with regard to prosumer and civic energy generation. In so doing, it draws widely on his experience gained during the development of steam microturbines for use in small combined heat and power stations based on the organic Rankine cycle (CHP-ORC). Major issues concerning the dynamic properties of mechanical systems, in particular rotating systems, are discussed, and the results obtained when using unconventional bearing systems, presented. Modeling and analysis of radial-flow and axial-flow microturbines a

Modeling and field studies of fouling in once-through steam generators

International Nuclear Information System (INIS)

Thompson, R.; Gaudreau, T.

1995-01-01

Efforts of the past 10 years to minimize fouling of the Crystal River-3 once-through steam generators are reviewed. The major focus has been on improving at-temperature pH control in the secondary cycle. Various concentrations of different pH control agents were tested in the field for hundreds of days to determine their effect on steam generator fouling. High concentrations of morpholine (50--100 ppm) in the feedwater were found to apparently produce de-fouling of the steam generators without an associated decrease in feedwater iron concentration as compared to that at lower levels of morpholine. Computer modeling of the pH(t) within the OTSG for the various chemistries tested indicates that the pH can change significantly with elevation within the steam generator by varying the pH control agent or its concentration. It is postulated that these variations in pH may change the surface charge of the tubes, tube support plates, and/or corrosion product particles in solution, to favor either deposition or repulsion of the particles, and thereby producing conditions that either favor fouling or de-fouling of the OTSG. Crystal River-3 experience indicates that corrosion product deposition and release processes inside the steam generator can be chemically manipulated to favor release, and thereby maximize plant performance, and delay or avoid costly hydraulic or chemical cleanings

Detonation cell size measurements and predictions in hydrogen-air-steam mixtures at elevated temperatures

International Nuclear Information System (INIS)

Ciccarelli, G.; Ginsberg, T.; Boccio, J.; Economos, C.

1994-01-01

The present research reports on the effect of initial mixture temperature on the experimentally measured detonation cell size for hydrogen-air-steam mixtures. Experimental and theoretical research related to combustion phenomena in hydrogen-air-steam mixtures has been ongoing for many years. However, detonation cell size data currently exists or hydrogen-air-steam mixtures up to a temperature of only 400K. Sever accident scenarios have been identified for light water reactors (LWRs) where hydrogen-air mixture temperatures in excess of 400K could be generated within containment. The experiments in this report focus on extending the cell size data base for initial mixture temperatures in excess of 400K. The experiments were carried out in a 10-cm inner-diameter, 6.1-m long heated detonation tube with a maximum operating temperature of 700K and spatial temperature uniformity of ±14K. Detonation cell size measurements provide clear evidence that the effect of hydrogen-air initial gas mixture temperature, in the range 300K--650K, is to decrease cell size and, hence, to increase the sensitivity of the mixture to undergo detonations. The effect of steam content, at any given temperature, is to increase the cell size and, thereby, to decrease the sensitivity of stoichiometric hydrogen-air mixtures. The hydrogen-air detonability limits for the 10-cm inside-diameter test vessel, based upon the onset of single-head spin, decreased from 15 percent by hydrogen at 300K down to about 9 percent hydrogen at 650K. The one-dimensional ZND model does a very good job at predicting the overall trends in the cell size data over the range of hydrogen-air-steam mixture compositions and temperature studied in the experiments

Patchouli Oil Production for Fixative Aromatheraphy: A Case Study Design of Condenser for Steam Distillation

Directory of Open Access Journals (Sweden)

Egi Agustian

2015-06-01

Full Text Available Patchouli oil is an important material for aromatherapy industries and made from patchouli leaves by steam distillation. The important parts of steam distillation is condenser. The problem is process of farmer's patchouli based on experience not used chemical process of steam distillation cause not effective and efficient. The study was designed of condensers on steam distillation for patchouli leaves. Specification of condenser type is shell and tube, stainless steel material and water as coolant. Tubing sized is ¼ inch with 500 mm of length and 24 pieces. Shell measuring 2.5 inch and 600 mm of length. The performance of condenser gave the efficiency at 94.51%. The process condition of ratio stems to leaves is 5:6 and 4.5 hours was considered ideal for steam distillation. The used of energy is lowest cause the efficiency condenser and yield will be higher. Patchouli oil yield of 2.39% and patchouli alcohol content is 35.65%.