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Sample records for steam header rupture

  1. Relap5/Mod3.1 analysis of main steam header rupture in VVER- 440/213 NPP

    Energy Technology Data Exchange (ETDEWEB)

    Kral, P. [Nuclear Research Inst. Rez (Switzerland)

    1995-12-31

    The presentation is focused on two main topics. First the applied modelling of PGV-4 steam generator for RELAP5 code are described. The results of steady-state calculation under reference conditions are compared against measured data. The problem of longitudinal subdivision of SG tubes is analysed and evaluated. Secondly, a best-estimate analysis of main steam header (MSH) rupture accident in WWER-440/213 NPP is presented. The low reliability of initiation of ESFAS signal `MSH Rupture` leads in this accident to big loss of secondary coolant, full depressurization of main steam system, extremely fast cool-down of both secondary and primary system, opening of PRZ SV-bypass valve with later liquid outflow, potential reaching of secondary criticality by failure of HPIS. 7 refs.

  2. Relap5/Mod3.1 analysis of main steam header rupture in VVER- 440/213 NPP

    Energy Technology Data Exchange (ETDEWEB)

    Kral, P [Nuclear Research Inst. Rez (Switzerland)

    1996-12-31

    The presentation is focused on two main topics. First the applied modelling of PGV-4 steam generator for RELAP5 code are described. The results of steady-state calculation under reference conditions are compared against measured data. The problem of longitudinal subdivision of SG tubes is analysed and evaluated. Secondly, a best-estimate analysis of main steam header (MSH) rupture accident in WWER-440/213 NPP is presented. The low reliability of initiation of ESFAS signal `MSH Rupture` leads in this accident to big loss of secondary coolant, full depressurization of main steam system, extremely fast cool-down of both secondary and primary system, opening of PRZ SV-bypass valve with later liquid outflow, potential reaching of secondary criticality by failure of HPIS. 7 refs.

  3. Relap5/Mod3.1 analysis of main steam header rupture in VVER- 440/213 NPP

    International Nuclear Information System (INIS)

    Kral, P.

    1995-01-01

    The presentation is focused on two main topics. First the applied modelling of PGV-4 steam generator for RELAP5 code are described. The results of steady-state calculation under reference conditions are compared against measured data. The problem of longitudinal subdivision of SG tubes is analysed and evaluated. Secondly, a best-estimate analysis of main steam header (MSH) rupture accident in WWER-440/213 NPP is presented. The low reliability of initiation of ESFAS signal 'MSH Rupture' leads in this accident to big loss of secondary coolant, full depressurization of main steam system, extremely fast cool-down of both secondary and primary system, opening of PRZ SV-bypass valve with later liquid outflow, potential reaching of secondary criticality by failure of HPIS

  4. 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

  5. A LOCA analysis for AHWR caused by ECCS header rupture

    International Nuclear Information System (INIS)

    Chatterjee, B.; Gawai, Amol; Gupta, S.K.; Kushwaha, H.S.

    2000-01-01

    Loss of coolant accident (LOCA) analyses for the proposed 750 MWth Advanced Heavy Water Reactor (AHWR), initiated by the rupture of 8 inch NB ECCS header has been carried out. This paper narrates the description of AHWR and associated ECCS, postulated scenario with which the analyses is carried out, results, discussion and conclusion

  6. Twin header bore welded steam generator for pressurized water reactors

    International Nuclear Information System (INIS)

    Davies, R.J.; Hirst, B.

    1979-01-01

    A description is given of a pressurized water reactor (PWR) steam generator concept, several examples of which have been in service for up to fourteen years. Details are given of the highly successful service record of this equipment and the features which have been incorporated to minimize corrosion and deposition pockets. The design employs a vertical U tube bundle carried off two horizontal headers to which the tubes are welded by the Foster Wheeler Power Products (FWPP) bore welding process. The factors to be considered in uprating the design to meet the current operating conditions for a 1000 MW unit are discussed. (author)

  7. Optimising residual stresses at a repair in a steam header to tubeplate weld

    International Nuclear Information System (INIS)

    Soanes, T.P.T.; Bell, W.; Vibert, A.J.

    2005-01-01

    Following the discovery of incorrect weld metal in the steam side shell to tubeplate weld in a type 316H stainless steel superheater steam header, a repair strategy had to be determined. The strategy adopted was to remove the incorrect weld material, which extended around the full circumference, by machining from the inside of the header, followed by rewelding from the inside using an automatic welding process and localised post-weld heat treatment. Due to concern over potential reheat cracking of the repair after return to service, a considerable amount of residual stress modelling was carried out to support the development and optimisation of a successful repair and heat treatment strategy and thus underwrite the safety case for return to service

  8. Hot steam header of a high temperature reactor as a benchmark problem

    International Nuclear Information System (INIS)

    Demierre, J.

    1990-01-01

    The International Atomic Energy Agency (IAEA) initiated a Coordinated Research Programme (CRP) on ''Design Codes for Gas-Cooled Reactor Components''. The specialists proposed to start with a benchmark design of a hot steam header in order to get a better understanding of the methods in the participating countries. The contribution of Switzerland carried out by Sulzer. The following report summarized the detailed calculations of dimensioning procedure and analysis. (author). 5 refs, 2 figs, 2 tabs

  9. Steam generator tube rupture (SGTR) scenarios

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  10. Thermal-hydraulic analysis of Ignalina NPP compartments response to group distribution header rupture using RALOC4 code

    International Nuclear Information System (INIS)

    Urbonavicius, E.

    2000-01-01

    The Accident Localisation System (ALS) of Ignalina NPP is a containment of pressure suppression type designed to protect the environment from the dangerous impact of the radioactivity. The failure of ALS could lead to contamination of the environment and prescribed public radiation doses could be exceeded. The purpose of the presented analysis is to perform long term thermal-hydraulic analysis of compartments response to Group Distribution Header rupture and verify if design pressure values are not exceeded. (authors)

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

    Science.gov (United States)

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

    1994-01-01

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

  12. Radioactivity transport following steam generator tube rupture

    International Nuclear Information System (INIS)

    Hopenfeld, J.

    1985-03-01

    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 PWR's shows that certain experimental data are needed for reliable off-site dose predictions. This article defines five parameters which 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 conjunction with CITADEL or they can be inserted in the appropriate equations which then conveniently can be programmed as a subroutine in thermal-hydraulic system codes. A joint Westinghouse, Electric Power Research Institute and Nuclear Regulatory Commission Program aimed at obtaining the five parameters empirically is described

  13. The Reasons of Steam Pipeline Elbow Rupture

    Directory of Open Access Journals (Sweden)

    Mesjasz A.

    2016-09-01

    Full Text Available In the paper the reasons for steam pipeline’s elbow material rupture, made of steel 13CrMo4-5 (15HM that is being used in the energetics. Based on the mechanical properties in the ambient temperature (Rm, Rp0,2 and elongation A5 and in the increased temperature (Rp0,2t it was found, that the pipeline elbow’s material sampled from the ruptured area has lower Rp0,2 i Rp0,2t by around 2% than it is a requirement for 13CrMo4-5 steel in it’s base state. The damage appeared as a result of complex stress state, that substantially exceeded the admissible tensions, what was the consequence of considerable structure degradation level. As a result of the microstructure tests on HITACHI S4200 microscope, the considerable development of the creeping process associates were found. Also the advances progress of the microstructure degradation was observed, which is substantial decomposition of bainite and multiple, with varied secretion size, and in most cases forming the micro cracks chains. With the use of lateral micro sections the creeping voids were observed, that creates at some places the shrinkage porosities clusters and micro pores.

  14. Steam generator tube rupture effects on a LOCA

    International Nuclear Information System (INIS)

    LaChance, J.L.

    1979-01-01

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

  15. Header integrity assessment

    Energy Technology Data Exchange (ETDEWEB)

    Rotvel, F. [ELSAMPROJEKT, Fredericia (Denmark); Sampietri, C. [ENEL, Milano (Italy); Verelst, L. [LABORELEC, Linkebeek (Belgium); Wortel, H. van [TNO, Apeldoorn (Netherlands); Li Ying Zhi [KEMA, Arnhem (Netherlands)

    1998-12-31

    In the late eighties creep cracks in the nozzle-to-header welds of high temperature headers became internationally recognized as a problem in older steam power plants. To study the problem a 2 1/4Cr1Mo service-exposed header, which was scrapped due to creep damage, was made available for testing. A full-scale model was fabricated with partly repaired nozzle to header welds and then tested at increased temperature. Loads included internal pressure and system loads. Damage accumulation and creep crack initiation and growth were predicted and experimentally verified. Conclusions and the practical implications for power plant operation are described. (orig.) 7 refs.

  16. Header integrity assessment

    Energy Technology Data Exchange (ETDEWEB)

    Rotvel, F [ELSAMPROJEKT, Fredericia (Denmark); Sampietri, C [ENEL, Milano (Italy); Verelst, L [LABORELEC, Linkebeek (Belgium); Wortel, H van [TNO, Apeldoorn (Netherlands); Zhi, Li Ying [KEMA, Arnhem (Netherlands)

    1999-12-31

    In the late eighties creep cracks in the nozzle-to-header welds of high temperature headers became internationally recognized as a problem in older steam power plants. To study the problem a 2 1/4Cr1Mo service-exposed header, which was scrapped due to creep damage, was made available for testing. A full-scale model was fabricated with partly repaired nozzle to header welds and then tested at increased temperature. Loads included internal pressure and system loads. Damage accumulation and creep crack initiation and growth were predicted and experimentally verified. Conclusions and the practical implications for power plant operation are described. (orig.) 7 refs.

  17. CATHENA simulations of steam generator tube rupture

    International Nuclear Information System (INIS)

    Abdul-Razzak, A.; Lin, M.R.; Wright, A.C.D.

    1997-01-01

    The CATHENA thermalhydraulic computer code was used to simulate various scenarios following a CANDU 9 steam generator tube rupture (SGTR) event. The analysis included cases with class IV power and emergency core cooling system (ECCS) available and other cases with subsequent loss of class IV power (LCIVP) or impairment of ECCS injection. Two main approaches were followed in the analysis of each case. In the first approach, D 2 O feed was credited to provide conservative data for input to radionuclide release and dose calculations. Also operator actions are credited. The other approach is designed to give conservative predictions with respect to the acceptance criteria of fuel and fuel channel integrity and to prove that in case of such event, the operator will have enough time to mitigate the consequences. This is done by not crediting makeup for the inventory loss and relying on the automatic operation of safety systems. The analysis of the cases of the first approach provided the required data for radionuclide release and dose calculations and gave a good insight into the required sequence of operator timely actions to mitigate the consequences of such event. On the other hand, the cases of the second approach confirmed compliance with regulatory requirements for pressure tube and fuel integrity. The runs with ECCS available, showed the ECCS injection is effective in filling and cooling the core and that regulatory requirement's for fuel and channel integrity are met. In the event of ECCS impairment, the earliest indication of late fuel heat-up is late enough to provide the operator with an adequate time to act in mitigating the consequences of this event. (author)

  18. 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.

  19. 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

  20. Steam generator tube rupture simulation using extended finite element method

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish, E-mail: smohanty@anl.gov; Majumdar, Saurin; Natesan, Ken

    2016-08-15

    Highlights: • Extended finite element method used for modeling the steam generator tube rupture. • Crack propagation is modeled in an arbitrary solution dependent path. • The FE model is used for estimating the rupture pressure of steam generator tubes. • Crack coalescence modeling is also demonstrated. • The method can be used for crack modeling of tubes under severe accident condition. - Abstract: A steam generator (SG) is an important component of any pressurized water reactor. Steam generator tubes represent a primary pressure boundary whose integrity is vital to the safe operation of the reactor. SG tubes may rupture due to propagation of a crack created by mechanisms such as stress corrosion cracking, fatigue, etc. It is thus important to estimate the rupture pressures of cracked tubes for structural integrity evaluation of SGs. The objective of the present paper is to demonstrate the use of extended finite element method capability of commercially available ABAQUS software, to model SG tubes with preexisting flaws and to estimate their rupture pressures. For the purpose, elastic–plastic finite element models were developed for different SG tubes made from Alloy 600 material. The simulation results were compared with experimental results available from the steam generator tube integrity program (SGTIP) sponsored by the United States Nuclear Regulatory Commission (NRC) and conducted at Argonne National Laboratory (ANL). A reasonable correlation was found between extended finite element model results and experimental results.

  1. Steam generator tube rupture simulation using extended finite element method

    International Nuclear Information System (INIS)

    Mohanty, Subhasish; Majumdar, Saurin; Natesan, Ken

    2016-01-01

    Highlights: • Extended finite element method used for modeling the steam generator tube rupture. • Crack propagation is modeled in an arbitrary solution dependent path. • The FE model is used for estimating the rupture pressure of steam generator tubes. • Crack coalescence modeling is also demonstrated. • The method can be used for crack modeling of tubes under severe accident condition. - Abstract: A steam generator (SG) is an important component of any pressurized water reactor. Steam generator tubes represent a primary pressure boundary whose integrity is vital to the safe operation of the reactor. SG tubes may rupture due to propagation of a crack created by mechanisms such as stress corrosion cracking, fatigue, etc. It is thus important to estimate the rupture pressures of cracked tubes for structural integrity evaluation of SGs. The objective of the present paper is to demonstrate the use of extended finite element method capability of commercially available ABAQUS software, to model SG tubes with preexisting flaws and to estimate their rupture pressures. For the purpose, elastic–plastic finite element models were developed for different SG tubes made from Alloy 600 material. The simulation results were compared with experimental results available from the steam generator tube integrity program (SGTIP) sponsored by the United States Nuclear Regulatory Commission (NRC) and conducted at Argonne National Laboratory (ANL). A reasonable correlation was found between extended finite element model results and experimental results.

  2. Rupture of steam lines between blocks D and G

    International Nuclear Information System (INIS)

    1999-01-01

    Analysis of steam lines rupture between blocks D and G of Ignalina NPP was performed. Model for evaluation of thermo hydrodynamic parameters was developed. Structural analysis of the shaft building was done as well. State of the art codes such as RELAP5, ALGOR, NEPTUNE were used in these calculations

  3. The effect of tube rupture location on the consequences of multiple steam generator tube rupture event

    International Nuclear Information System (INIS)

    Jeong, Ji Hwan; Kweon, Young Chul

    2002-01-01

    A multiple steam generator tube rupture (MSGTR) event has never occurred in the commercial operation of nuclear reactors while single steam generator tube rupture (SGTR) events are reported to occur every 2 years. As there has been no occurrence of a MSGTR event, the understanding of transients and consequences of this event is very limited. In this study, a postulated MSGTR event in an advanced power reactor 1400 (APR 1400) is analyzed using the thermal-hydraulic system code, MARS1.4. The APR 1400 is a two-loop, 3893 MWt, PWR proposed to be built in 2010. The present study aims to understand the effects of rupture location in heat transfer tubes following a MSGTR event. The effects of five tube rupture locations are compared with each other. The comparison shows that the response of APR1400 allows the shortest time for operator action following a tube rupture in the vicinity of the hot-leg side tube sheet and allows the longest time following a tube rupture at the tube top. The MSSV lift time for rupture at the tube-top is evaluated as 24.5% larger than that for rupture at the hot-leg side tube sheet

  4. Corrosion and Rupture of Steam Generator Tubings in PWRs

    International Nuclear Information System (INIS)

    Hwang, Seong Sik; Kim, Hong Pyo

    2007-08-01

    This report is intended to provide corrosion engineers in the filed of nuclear energy with information on the corrosion and rupture behavior of steam generator tubing in PWRs. Various types of corrosion in PWR steam generator tubing have been reported all around the world, and countermeasures such as the addition of corrosion inhibitors, a water chemistry control, a tube plugging and sleeving have been applied. Steam generators equipped with alloy 600 tubing, which are not so resistant to a stress corrosion cracking (SCC), have generally been replaced with new steam generators made of alloy 690 TT (Thermally treated). Pull tube examination results which were performed of KAERI are summarized. The tubes were affected by a pitting, SCC, and a denting. Nondestructive examination method for the tubes and repair techniques are also reviewed. In addition, the regulatory guidance of some countries are reviewed. As a part of a tube integrity project in Korea, some results on a tube rupture and leak behaviors for axial cracks are also mentioned

  5. Corrosion and Rupture of Steam Generator Tubings in PWRs

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Seong Sik; Kim, Hong Pyo

    2007-08-15

    This report is intended to provide corrosion engineers in the filed of nuclear energy with information on the corrosion and rupture behavior of steam generator tubing in PWRs. Various types of corrosion in PWR steam generator tubing have been reported all around the world, and countermeasures such as the addition of corrosion inhibitors, a water chemistry control, a tube plugging and sleeving have been applied. Steam generators equipped with alloy 600 tubing, which are not so resistant to a stress corrosion cracking (SCC), have generally been replaced with new steam generators made of alloy 690 TT (Thermally treated). Pull tube examination results which were performed of KAERI are summarized. The tubes were affected by a pitting, SCC, and a denting. Nondestructive examination method for the tubes and repair techniques are also reviewed. In addition, the regulatory guidance of some countries are reviewed. As a part of a tube integrity project in Korea, some results on a tube rupture and leak behaviors for axial cracks are also mentioned.

  6. Fission product retention during faults involving steam generator tube rupture

    International Nuclear Information System (INIS)

    Rodliffe, R.S.

    1983-08-01

    In some PWR fault conditions, such as stuck open safety relief valve in the secondary circuit or main steam line break, the release of fission products to the atmosphere may be increased by the leakage of primary coolant into the secondary circuit following steam generator tube rupture. The release may be reduced by retention either within the primary circuit or within the affected steam generator unit (SGU). The mechanisms leading to retention are reviewed and quantified where possible. The parameters on which any analysis will be most critically dependent are identified. Fission product iodine and caesium may be retained in the secondary side of a SGU either by partition to retained water or by droplet deposition on surfaces and subsequent evaporation to dryness. Two extreme simplifications are considered: SGU 'dry', i.e. the secondary side is steam filled, and SGU 'wet', i.e. the tube bundle is covered with water. Consideration is given to: the distribution of fission products between gaseous and aerosol forms; mechanisms for droplet formation, deposition and resuspension; fission product retention during droplet or film evaporation primary coolant mixing and droplet scrubbing in a wet SGU; and the performance of moisture separators and steam driers. (author)

  7. Investigation of a steam generator tube rupture sequence using VICTORIA

    International Nuclear Information System (INIS)

    Bixler, N.E.; Erickson, C.M.; Schaperow, J.H.

    1995-01-01

    VICTORIA-92 is a mechanistic computer code for analyzing fission product behavior within the reactor coolant system (RCS) during a severe reactor accident. It provides detailed predictions of the release of radionuclides and nonradioactive materials from the core and transport of these materials within the RCS. The modeling accounts for the chemical and aerosol processes that affect radionuclide behavior. Coupling of detailed chemistry and aerosol packages is a unique feature of VICTORIA; it allows exploration of phenomena involving deposition, revaporization, and re-entrainment that cannot be resolved with other codes. The purpose of this work is to determine the attenuation of fission products in the RCS and on the secondary side of the steam generator in an accident initiated by a steam generator tube rupture (SGTR). As a class, bypass sequences have been identified in NUREG-1150 as being risk dominant for the Surry and Sequoyah pressurized water reactor (PWR) plants

  8. WWER-1000/320 steam generator collector rupture. Radiological consequences

    Energy Technology Data Exchange (ETDEWEB)

    Ivanova, A; Sartmadzhiev, A; Balabanov, E [Energoproekt, Sofia (Bulgaria)

    1996-12-31

    A model describing a hypothetical accident with direct release of primary coolant to the atmosphere is proposed. Cover lifting of the primary collector due to a rupture of the fixing bolts leads to a coolant release. The initial and boundary conditions of the accident scenario have been selected to provide for the most unfavorable conditions. The total release of primary coolant during the first 15 min of transient are estimated to 50.8 tons, of these 48.5 t with the initial activity in the primary coolant circuit. Without evacuation or sheltering, after 7 days of exposure, the expected dose at the boundary of the restricted zone is 0.0182 Sv for the whole body and 0.184 Sv for the thyroid gland. The effective equivalent dose on the site would be 0.0521 Sv. As a result of the analysis it is concluded that the steam generator collector rupture is not jeopardizing the core heat removal even with a minimum configuration of ECCS as the cooling is accomplished through the steam generators. The radiological consequences of the accident would be relatively small if an emergency procedure is applied at the 15-th minute of the transient. 1 ref.

  9. Indian Point 2 steam generator tube rupture analyses

    International Nuclear Information System (INIS)

    Dayan, A.

    1985-01-01

    Analyses were conducted with RETRAN-02 to study consequences of steam generator tube rupture (SGTR) events. The Indian Point, Unit 2, power plant (IP2, PWR) was modeled as a two asymmetric loops, consisting of 27 volumes and 37 junctions. The break section was modeled once, conservatively, as a 150% flow area opening at the wall of the steam generator cold leg plenum, and once as a 200% double-ended tube break. Results revealed 60% overprediction of breakflow rates by the traditional conservative model. Two SGTR transients were studied, one with low-pressure reactor trip and one with an earlier reactor trip via over temperature ΔT. The former is more typical to a plant with low reactor average temperature such as IP2. Transient analyses for a single tube break event over 500 seconds indicated continued primary subcooling and no need for steam line pressure relief. In addition, SGTR transients with reactor trip while the pressurizer still contains water were found to favorably reduce depressurization rates. Comparison of the conservative results with independent LOFTRAN predictions showed good agreement

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

  11. Using MAAP 4.0 to determine risks from steam generator tube leaks or ruptures

    International Nuclear Information System (INIS)

    Fuller, E.L.; Kenton, M.A.

    1996-01-01

    As part of the Electric Power Research Institute (EPRI) program on steam generator degradation specific management (SGDSM), the nuclear industry is investigating the effects on plant risk of severe accidents involving steam generator tube leaks or ruptures. Such accidents fall into three classes: those caused by spontaneous, steam generator tube ruptures (SGTRs) that subsequently result in core damage; those caused by design-basis accidents that lead to induced tube ruptures and subsequent core damage; and those that progress to core damage, such as a station blackout (SBO), with subsequent induced tube leakage or rupture. In each case, the potential exists for a significant fraction of the fission products released from a damaged core to reach the environment through the leaking or ruptured tubes

  12. Prognostics for Steam Generator Tube Rupture using Markov Chain model

    International Nuclear Information System (INIS)

    Kim, Gibeom; Heo, Gyunyoung; Kim, Hyeonmin

    2016-01-01

    This paper will describe the prognostics method for evaluating and forecasting the ageing effect and demonstrate the procedure of prognostics for the Steam Generator Tube Rupture (SGTR) accident. Authors will propose the data-driven method so called MCMC (Markov Chain Monte Carlo) which is preferred to the physical-model method in terms of flexibility and availability. Degradation data is represented as growth of burst probability over time. Markov chain model is performed based on transition probability of state. And the state must be discrete variable. Therefore, burst probability that is continuous variable have to be changed into discrete variable to apply Markov chain model to the degradation data. The Markov chain model which is one of prognostics methods was described and the pilot demonstration for a SGTR accident was performed as a case study. The Markov chain model is strong since it is possible to be performed without physical models as long as enough data are available. However, in the case of the discrete Markov chain used in this study, there must be loss of information while the given data is discretized and assigned to the finite number of states. In this process, original information might not be reflected on prediction sufficiently. This should be noted as the limitation of discrete models. Now we will be studying on other prognostics methods such as GPM (General Path Model) which is also data-driven method as well as the particle filer which belongs to physical-model method and conducting comparison analysis

  13. 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

  14. 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

  15. Steam generator secondary pH during a steam generator tube rupture

    International Nuclear Information System (INIS)

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

    1991-12-01

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

  16. 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)

  17. Intentional back flow effects on ruptured steam generator cooldown during a SGTR event for KSNP

    International Nuclear Information System (INIS)

    Kim, C.W.; Park, S.J.; Choi, C.J.; Seo, J.T.

    2004-01-01

    For an optimum recovery from a steam generator tube rupture (SGTR) event, the operators are directed to isolate the steam generator (SG) with ruptured tube as early as possible to minimize the radioactive material release. However, the reactor coolant system (RCS) cooldown and depressurization to the shutdown cooling system (SCS) operation conditions using the intact SG only are hard to achieve unless the ruptured SG is properly cooled since the ruptured SG, which is isolated by operator, remains at high temperature even though the RCS has been cooled down. The effects of intentional back flow from the SG secondary side to the RCS through the ruptured U-tube on the the ruptured SG cooldown were evaluated for the pressurized light water reactor, especially for the Korean standard nuclear power plant (KSNP). In order to evaluate the back flow effect, a series of analyses was conducted using the RELAP5/MOD3 computer code. For the first stage of the analysis, the cooldown process by natural circulation in the SG secondary side was simulated for the initial conditions of the ruptured SG cooldown. In the next analysis stage, two methods of the ruptured SG cooldown by using back flow after RCS cooldown were evaluated. One utilizes the steam condensation on the uncovered U-tube surface, and the other is a SG drain and fill. In the former method, SG tubes are exposed to the steam space by draining SG secondary water into the RCS in order to condense the steam directly onto the uncovered tubes. This method showed that the steam condensation decreased SG secondary pressure and temperature rapidly, demonstrating its effectiveness for cooling. However, this process has a limited applicability if the rupture is located at the lower region. The latter method, draining by back flow and filling using the feedwater system was also found to be effective in ruptured SG cooldown and depressurization even if the rupture occurred at the top of the U-tube. It is concluded that the

  18. A proven twin header design for small PWRs

    International Nuclear Information System (INIS)

    Davidov, Maurice

    1987-01-01

    A unique design of PWR steam generator, developed by Foster Wheeler in Britain more than 30 years ago, avoids the problem of tubesheet sludge accumulation. The twin header steam generator uses a vertical, inverted U-tube bundle connected to cylindrical inlet and outlet headers. The advantages of the design and operating experience are outlined. (author)

  19. Steam generator tubes rupture probability estimation - study of the axially cracked tube case

    International Nuclear Information System (INIS)

    Mavko, B.; Cizelj, L.; Roussel, G.

    1992-01-01

    The objective of the present study is to estimate the probability of a steam generator tube rupture due to the unstable propagation of axial through-wall cracks during a hypothetical accident. For this purpose the probabilistic fracture mechanics model was developed taking into account statistical distributions of influencing parameters. A numerical example considering a typical steam generator seriously affected by axial stress corrosion cracking in the roll transition area, is presented; it indicates the change of rupture probability with different assumptions focusing mostly on tubesheet reinforcing factor, crack propagation rate and crack detection probability. 8 refs., 4 figs., 4 tabs

  20. Intentional back flow effects on ruptured steam generator cooldown during a SGTR event for KSNP

    International Nuclear Information System (INIS)

    Seok, Jeong Park; Cheol, Woo Kim; Chul, Jin Choi; Jong, Tae Seo

    2001-01-01

    For an optimum recovery from a Steam Generator Tube Rupture (SGTR) event, the operators are directed to isolate the steam generator (SG) with ruptured tube(s) as early as possible in order to minimize the radioactive material release. However, the Reactor Coolant System (RCS) cooldown and depressurization to the Residual Heat Removal (RHR) System operation conditions using the intact SG only can not be readily achievable unless the affected SG is properly cooled since the isolated SG remains at high temperature even though the RCS has been cooled down. Therefore, a study on the intentional back flow from the ruptured SG secondary side to the RCS was performed to evaluate its effectiveness on the ruptured SG cooldown during a SGTR event for the pressurized light water reactor, especially for the Korean Standard Nuclear Power Plant (KSNP). In order to evaluate the intentional back flow effect, a series of analyses was conducted by using RELAP5/MOD3 computer code. In these analyses, the primary and secondary systems of KSNP are modeled including the major Nuclear Steam Supply System (NSSS) components such as the reactor vessel, steam generators, hot and cold legs, pressurizer, and reactor coolant pumps. Also, the key safety systems and control systems are modeled. Using this model, two possible methods of the ruptured SG cooldown by using back flow after RCS cooldown were evaluated: the first method is a tube uncover method, and the second method is a SG drain (back flow) and fill method. (author)

  1. 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

  2. Realistic analysis of steam generator tube rupture accident in Angra-1 reactor

    International Nuclear Information System (INIS)

    Fontes, S.W.F.

    1989-01-01

    This paper presents the analysis of different scenarios for a Steam Generator Tube Rupture accident (SGTR) in Angra-1 NPP. The results and conclusions will be used as support in the preparation of the emergency situation programs for the plant. For the analysis a SGTR simulation was performed with RETRAN-02 code. The results indicated that the core integrity and the plant itself will not affect by small ruptures in SG tubes. For large ruptures the analysis demonstrated that the accident may have harmful consequences if the operator do not actuate effectively since the initial moments of the accidents. (author) [pt

  3. 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.)

  4. Analysis of induced steam generator tube rupture using MAAP 4.0

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  5. Steam generator tube rupture: studies to improve plant procedure

    International Nuclear Information System (INIS)

    Tellier, N.; Zilliox, C.

    1984-10-01

    These accidents have the particularities to lead to atmospheric radioactive release and to be able to be determinated with appropriate operator actions. These radioactive releases are function of several parameters of which sensitivity is analyzed. The major part of the calculations were performed by EDF with an home made code called ''AXEL''. The main conclusions are: - the optimization of the safety injection monitoring to minimize radioactive releases to atmosphere, while ensuring the cooling of the core; - the radioactive releases to atmosphere are very low in any case but much more important if the filling of the steam generator secondary side cannot be avoided

  6. Ligament rupture and unstable burst behaviors of axial flaws in steam generator U-bends

    Energy Technology Data Exchange (ETDEWEB)

    Bahn, Chi Bum, E-mail: bahn@pusan.ac.kr [Pusan National University, 2 Busandaehak-ro 63 beon-gil, Geumjeong-gu, Busan 609-735 (Korea, Republic of); Oh, Young-Jin [KEPCO Engineering & Construction Co. Inc., Seongnam 463-870 (Korea, Republic of); Majumdar, Saurin [Argonne National Laboratory, Lemont, IL 60439 (United States)

    2015-11-15

    Highlights: • Ligament rupture and unstable burst pressure tests were conducted with U-bends. • In general, U-bends showed higher ligament rupture and burst pressures than straight tubes. • U-bend test data was bounded by 90% lower limit of the probabilistic models for straight tubes. • Prediction models for straight tubes could be conservatively applied to U-bends. - Abstract: Incidents of U-bend cracking in steam generator (SG) tubes have been reported, some of which have led to tube rupture. Experimental and analytical modeling efforts to determine the failure criteria of flawed SG U-bends are limited. To evaluate structural integrity of flawed U-bends, ligament rupture and unstable burst pressure tests were conducted on 57 and 152 mm bend radius U-bends with axial electrical discharge machining notches. In general, the ligament rupture and burst pressures of the U-bends were higher than those of straight tubes with similar notches. To quantitatively address the test data scatter issue, probabilistic models were introduced. All ligament rupture and burst pressures of U-bends were bounded by 90% lower limits of the probabilistic models for straight tubes. It was concluded that the prediction models for straight tubes could be applied to U-bends to conservatively evaluate the ligament rupture and burst pressures of U-bends with axial flaws.

  7. Ligament rupture and unstable burst behaviors of axial flaws in steam generator U-bends

    International Nuclear Information System (INIS)

    Bahn, Chi Bum; Oh, Young-Jin; Majumdar, Saurin

    2015-01-01

    Highlights: • Ligament rupture and unstable burst pressure tests were conducted with U-bends. • In general, U-bends showed higher ligament rupture and burst pressures than straight tubes. • U-bend test data was bounded by 90% lower limit of the probabilistic models for straight tubes. • Prediction models for straight tubes could be conservatively applied to U-bends. - Abstract: Incidents of U-bend cracking in steam generator (SG) tubes have been reported, some of which have led to tube rupture. Experimental and analytical modeling efforts to determine the failure criteria of flawed SG U-bends are limited. To evaluate structural integrity of flawed U-bends, ligament rupture and unstable burst pressure tests were conducted on 57 and 152 mm bend radius U-bends with axial electrical discharge machining notches. In general, the ligament rupture and burst pressures of the U-bends were higher than those of straight tubes with similar notches. To quantitatively address the test data scatter issue, probabilistic models were introduced. All ligament rupture and burst pressures of U-bends were bounded by 90% lower limits of the probabilistic models for straight tubes. It was concluded that the prediction models for straight tubes could be applied to U-bends to conservatively evaluate the ligament rupture and burst pressures of U-bends with axial flaws.

  8. CITADEL: a computer code for the analysis of iodine behavior in steam generator tube rupture accidents

    International Nuclear Information System (INIS)

    1982-04-01

    The computer code CITADEL was written to analyze iodine behavior during steam generator tube rupture accidents. The code models the transport and deposition of iodine from its point of escape at the steam generator primary break until its release to the environment. This report provides a brief description of the code including its input requirements and the nature and form of its output. A user's guide describing the manner in which the input data are required to be set up to run the code is also provided

  9. Comparative Analyses on OPR1000 Steam Generator Tube Rupture Event Emergency Operational Guideline

    International Nuclear Information System (INIS)

    Lee, Sang Won; Bae, Yeon Kyoung; Kim, Hyeong Teak

    2006-01-01

    The Steam Generator Tube Rupture (SGTR) event is one of the important scenarios in respect to the radiation release to the environment. When the SGTR occurs, containment integrity is not effective because of the direct bypass of containment via the ruptured steam generator to the MSSV and MSADV. To prevent this path, the Emergency Operational Guideline of OPR1000 indicates the use of Turbine Bypass Valves (TBVs) as an effective means to depressurize the main steam line and prevent the lifting of MSSV. However, the TBVs are not operable when the offsite power is not available (LOOP). In this situation, the RCS cool-down is achieved by opening the both intact and ruptured SG MSADV. But this action causes the large amount of radiation release to the environment. To minimize the radiation release to the environment, KSNP EOG adopts the improved strategy when the SGTR concurrently with LOOP is occurred. However, these procedures show some duplicated procedure and branch line that might confusing the operator for optimal recovery action. So, in this paper, the comparative analysis on SGTR and SGTR with LOOP is performed and optimized procedure is proposed

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  11. Pipe rupture and steam/water hammer design loads for dynamic analysis of piping systems

    International Nuclear Information System (INIS)

    Strong, B.R. Jr.; Baschiere, R.J.

    1978-01-01

    The design of restraints and protection devices for nuclear Class I and Class II piping systems must consider severe pipe rupture and steam/water hammer loadings. Limited stress margins require that an accurate prediction of these loads be obtained with a minimum of conservatism in the loads. Methods are available currently for such fluid transient load development, but each method is severely restricted as to the complexity and/or the range of fluid state excursions which can be simulated. This paper presents a general technique for generation of pipe rupture and steam/water hammer design loads for dynamic analysis of nuclear piping systems which does not have the limitations of existing methods. Blowdown thrust loadings and unbalanced piping acceleration loads for restraint design of all nuclear piping systems may be found using this method. The technique allows the effects of two-phase distributed friction, liquid flashing and condensation, and the surrounding thermal and mechanical equipment to be modeled. A new form of the fluid momentum equation is presented which incorporates computer generated fluid acceleration histories by inclusion of a geometry integral termed the 'force equivalent area' (FEA). The FEA values permit the coupling of versatile thermal-hydraulic programs to piping dynamics programs. Typical applications of the method to pipe rupture problems are presented and the resultant load histories compared with existing techniques. (Auth.)

  12. Five Tubes Rupture at Cold Side of Steam Generator Simulation Test Report Using the ATLAS

    International Nuclear Information System (INIS)

    Kang, Kyoung Ho; Park, Hyun Sik; Cho, Seok

    2010-12-01

    In this study, a postulated SGTR event of the APR1400 was experimentally investigated with the ATLAS. In order to simulate a double-ended rupture of five U-tubes in the APR1400, the SGTR-CL-02 test was performed with the ATLAS. The main objectives of this test were not only to provide a physical insight into the system response of the APR1400 during the SGTR but also to produce integral effect experimental data to validate the safety analysis code. In the present report, major thermal-hydraulic phenomena such as the system pressures, the collapsed water levels, and the break flow rate were presented and discussed. Compared to the case of a single U-tube rupture test, opening frequency of the MSSVs in the intact steam generator (SG-2) was highly reduced after 500 seconds in the present SGTR-CL-02 test. Large discharge of the primary inventory resulted in rapid depressurization of the primary system and consequently early injection of the SIP. Supply of cold ECC water by the SIPs reduced the energy transfer to the secondary side compared with the single U-tube rupture case. Less heat transfer to the secondary side had more influence on the secondary pressure of the affected steam generator than the break flow. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS and the RELAP5 as well as the SPACE code

  13. Five Tubes Rupture at Hot Side of Steam Generator Simulation Test Report Using the ATLAS

    International Nuclear Information System (INIS)

    Kang, Kyoung Ho; Park, Hyun Sik; Cho Seok

    2010-12-01

    In this study, a postulated SGTR event of the APR1400 was experimentally investigated with the ATLAS. In order to simulate a double-ended rupture of five U-tubes in the APR1400, the SGTR-HL-05 test was performed with the ATLAS. The main objectives of this test were not only to provide a physical insight into the system response of the APR1400 during the SGTR but also to produce integral effect experimental data to validate the SPACE code. In the present report, major thermal-hydraulic phenomena such as the system pressures, the collapsed water levels, and the break flow rate were presented and discussed. On the contrary to the case of a single U-tube rupture test, the MSSV of the intact steam generator was not opened any more after 1500 seconds in the present SGTR-HL-05 test. Large discharge of the primary inventory resulted in rapid depressurization of the primary system and consequently early injection of the SIP. Supply of cold ECC water by the SIPs reduced the energy transfer to the secondary side compared with the single U-tube rupture case. Less heat transfer to the secondary side had more influence on the secondary pressure of the affected steam generator than the break flow. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS and the RELAP5 as well as the SPACE code

  14. Evaluation of tube rupture simulation test (TRUST-1) for FBR steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Hayashida, Yoshihiko; Hamada, Hirotsugu [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1996-06-01

    The intermediate water leak in an FBR Steam Generator (SG) causes a high temperature and corrosive sodium-water reaction jet. In such cases, it is necessary to evaluate the wastage and overheating rupture behavior of heat transfer tubes. Especially, in the large SG that aims at high temperature of sodium and high temperature/pressure of water, the establishment of the rational evaluation method is important. In this paper, as a basic experiment to make clear the phenomenon of overheating rupture, tests and analysis of Tube Rupture Simulation Test-1 (TRUST-1) were conducted. TRUST-1 simulates the overheating rupture of the tube made of Mod.9Cr-1Mo steel by nitrogen gas pressurization and quick induction heating. The result of TRUST-1 are as follows: (1) The breaking strength predicted by the internal pressure is larger than the tensile strength of the tube material. (2) The margin of the breaking strength from the tensile strength of the tube material has a tendency of decreasing with the heating rate, especially in the lower temperature region. (3) Using an theoretical formula that is deduced from the steady creep model and appropriate experimental coefficients that are determined by the test data, the breaking strength can be reasonably evaluated. (author)

  15. Probability of a steam generator tube rupture due to the presence of axial through wall cracks

    International Nuclear Information System (INIS)

    Mavko, B.; Cizelj, L.

    1991-01-01

    Using the Leak-Before-Break (LBB) approach to define tube plugging criteria a possibility to operate with through wall crack(s) in steam generator tubes may be considered. This fact may imply an increase in tube rupture probability. Improved examination techniques (in addition to the 100% tube examination) have been developed and introduced to counterbalance the associated risk. However no estimates of the amount of total increase or decrease of risk due to the introduction of LBB have been made. A scheme to predict this change of risk is proposed in the paper, based on probabilistic fracture mechanics analysis of axial cracks combined with available data of steam generator tube nondestructive examination reliability. (author)

  16. Design guideline to prevent the pipe rupture by radiolysis gases in BWR steam piping

    International Nuclear Information System (INIS)

    Inagaki, T.; Miyagawa, M.; Ota, T.; Sato, T.; Sakata, K.

    2009-01-01

    In late 2001, pipe rupture accidents due to fast combustion of radiolysis gas occurred in Japan and elsewhere's BWR power plants. TENPES began to set up the guideline as action to such a new problem to prevent accumulation and combustion of radiolysis gas in BWR steam piping. And then, the first edition of guideline was published in October 2005. Afterwards, the experimental study about combustion/detonation of radiolysis gas have been continued. And in March 2007, TENPES published a revised edition of the guideline. This is the report of the revised edition of that guideline. According to this guideline, it became possible to design BWR's steam piping to prevent accumulation of radiolysis gas. (author)

  17. Steam-generator-tube-rupture transients for pressurized-water reactors

    International Nuclear Information System (INIS)

    Dobranich, D.; Henninger, R.J.; DeMuth, N.S.

    1982-01-01

    Steam generator tube ruptures with and without concurrent main-steam-line break are investigated for pressurized water reactors supplied by the major US vendors. The goal of these analyses is to provide thermodynamic and flow conditions for the determination of iodine transport to the environment and to provide an evaluation of the adequacy of the plant safety systems and operating procedures for controlling these transients. The automatic safety systems of the plant were found to be adequate for the mitigation of these transients. Emergency injection system flows equilibrated with the leakage flows and prevented core uncovery. Sufficient time was afforded by the plant safety systems for the operators to identify the problem and to take appropriate measures

  18. Analytical and experimental studies of mechanical consequences of a steam generator tube rupture

    International Nuclear Information System (INIS)

    Duc, B.; Sudreau, F.; Rassineux, B.

    1995-01-01

    Concerning to steam generator tubes support mechanical loadings due to the impact f the ruptured one, Electricite de France, with the support of Commissariat a l'Energie. Atomique, has undertaken a large study in order to evaluate the consequences of such loadings. This paper first presents the results of the tests performed on AQUITAINE 2 facility (CEA Cadarache research center) for nominal, faulted and boiler effect conditions. Those results are then compared with numerical dynamic elastoplastic analyses performed with CASTEM 2000 code (CEA system). (author). 1 ref., 14 figs

  19. Application of probabilistic fracture mechanics to estimate the risk of rupture of PWR steam generator tubes

    International Nuclear Information System (INIS)

    Pitner, P.; Riffard, T.; Granger, B.

    1992-01-01

    This paper describes the COMPROMIS code developed by Electricite de France (EDF) to optimize the tube bundle maintenance of steam generators. The model, based on probabilistic fracture mechanics, makes it possible to quantify the influence of in-service inspections and maintenance work on the risk of an SG tube rupture, taking all significant parameters into account as random variables (initial defect size distribution, reliability of non-destructive detection and sizing, crack initiation and propagation, critical sizes, leak before risk of break, etc.). (authors). 5 refs., 8 figs., 3 tabs

  20. Uncertainty analysis for probabilistic steam generators tube rupture in LBB applications

    Energy Technology Data Exchange (ETDEWEB)

    Durbec, V.; Pitner, P.; Pages, D. [Electricite de France, 78 - Chatou (France). Research and Development Div.; Riffard, T. [Electricite de France, 69 - Villeurbanne (France). Engineering and Construction Div.; Flesch, B. [Electricite de France, 92 - Paris la Defense (France). Generation and Transmission Div.

    1997-10-01

    Steam Generators (SG) of Pressurized Water Reactors have experienced world wide various types of tube degradations, mainly from stress corrosion cracking; because of this damage, primary-secondary leakage or tube rupture can occur. Safety against the risk of tube rupture is achieved through a combination of periodic in-service inspections (eddy current testing), surveillance of leaks during operation (leak before break concept) and tube plugging. In order to optimize the tube bundle SG maintenance, Electricite de France has developed a specific software named COMPROMIS. The model, based on probabilistic fracture mechanics makes it possible to quantify the influence of in service inspections and maintenance work on the risk of a SG Tube Rupture (SGTR), taking all significant parameters into account as random variables (initial defect size distribution, reliability of non-destructive examinations, crack initiation and propagation, critical sizes, leak before risk of break, etc...). This paper focuses on the leak rate calculation module and presents a sensitivity study of the influence of the leak before break on the conditional failure probability. (author) 8 refs.

  1. Uncertainty analysis for probabilistic steam generators tube rupture in LBB applications

    International Nuclear Information System (INIS)

    Durbec, V.; Pitner, P.; Pages, D.; Riffard, T.; Flesch, B.

    1997-10-01

    Steam Generators (SG) of Pressurized Water Reactors have experienced world wide various types of tube degradations, mainly from stress corrosion cracking; because of this damage, primary-secondary leakage or tube rupture can occur. Safety against the risk of tube rupture is achieved through a combination of periodic in-service inspections (eddy current testing), surveillance of leaks during operation (leak before break concept) and tube plugging. In order to optimize the tube bundle SG maintenance, Electricite de France has developed a specific software named COMPROMIS. The model, based on probabilistic fracture mechanics makes it possible to quantify the influence of in service inspections and maintenance work on the risk of a SG Tube Rupture (SGTR), taking all significant parameters into account as random variables (initial defect size distribution, reliability of non-destructive examinations, crack initiation and propagation, critical sizes, leak before risk of break, etc...). This paper focuses on the leak rate calculation module and presents a sensitivity study of the influence of the leak before break on the conditional failure probability. (author)

  2. Single Tube Rupture at Cold Side of Steam Generator Simulation Test Report Using the ATLAS

    International Nuclear Information System (INIS)

    Kang, Kyoung Ho; Park, Hyun Sik; Cho, Seok

    2010-12-01

    In this study, a postulated SGTR event of the APR1400 was experimentally investigated with the ATLAS. In order to simulate a double-ended rupture of a single U-tube in the APR1400, the SGTR-CL-01 test was performed with the ATLAS. The main objectives of this test were not only to provide a physical insight into the system response of the APR1400 during the SGTR but also to produce integral effect experimental data to validate the safety analysis code. In the present report, major thermal-hydraulic phenomena such as the system pressures, the collapsed water levels, and the break flow rate were presented and discussed. Following the reactor trip induced by high steam generator level (HSGL) signal, the primary system pressure decreased and the secondary system pressure increased until the MSSVs was opened. The MSSVs repeated on and off status depending on the secondary system pressure during the whole test period. Due to the break flow, the collapsed water level of the affected steam generator showed milder decrease than that of the intact steam generator. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS and the RELAP5 as well as the SPACE code. Furthermore, this data can be utilized to identify any code deficiency for a SGTR simulation, especially for DVI-adapted plants

  3. 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

  4. Analysing The Thermalhydraulic Parameters Of VVER-1000 Reactor For The Accident Of Steam Generator Tube Rupture

    International Nuclear Information System (INIS)

    Luu Nam Hai; Truong Cong Thang

    2011-01-01

    To ensure the safety operation of nuclear power plant (NPP), a lot of postulated accident scenarios were considered and analysed. This research chose and analysed the accident of steam generator tube rupture (SGTR) under the actual plant conditions by using the simulation program PCTRAN. The SGTR accident is happen when the NPP is under operation with the steady state condition (power of 3000 MWth, primary pressure of 157 bar and secondary pressure of 63 bar). The accident is initiated by creating a break with equivalent diameter of 100 mm in the area of lower row heat exchanging tubes. The result of analysis is compared with the calculation of the Shariz University, Iran using the thermal hydraulics code RELAP5/mod3.2 and the report in the PSAR data of VVER-1000. This comparison shows that it is possible for using PCTRAN to analyse accidents of VVER-1000 reactor. (author)

  5. Radioactivity release vs probability for a steam generator tube rupture accident

    International Nuclear Information System (INIS)

    Buslik, A.J.; Hall, R.E.

    1978-01-01

    A calculation of the probability of obtaining various radioactivity releases from a steam generator tube rupture (SGTR) is presented. The only radioactive isotopes considered are Iodine-131 and Xe-133. The particular accident path considered consists of a double-ended guillotine SGTR followed by loss of offsite power (LOSP). If there is no loss of offsite power, and no system fault other than the SGTR, it is judged that the consequences will be minimal, since the amount of iodine released through the condenser air ejector is expected to be quite small; this is a consequence of the fact that the concentration of iodine in the vapor released from the condenser air ejector is very small compared to that dissolved in the condensate water. In addition, in some plants the condenser air ejector flow is automatically diverted to containment or a high-activity alarm. The analysis presented here is for a typical Westinghouse PWR such as described in RESAR-3S

  6. Preliminary evaluation of steam generator tube rupture (SGTR) accident in lead cooled reactor

    International Nuclear Information System (INIS)

    Frano, R. Lo; Forasassi, G.

    2009-01-01

    In this paper some contributions are provided to the development of a European Lead-cooled System, known as the ELSY project (within EU-6 Framework Project); that will constitute a possible reference system for a large lead-cooled reactor of GEN IV. Steam generator (SG) tubing of this system type might be subject to a variety of degradation processes, such as cracking, wall thinning and potential leakage or rupture, eventually leading to the failure of one or more SG tubes that constitute a steam generator tube rupture (SGTR) accident with possible consequences for the safety of the primary systems. It is therefore of interest for the designer to know how the SG itself, as well as the vessel and internals structures, behave under impulsive loading conditions (in form of a rapid and strong increase of pressure) that can arise as consequences of the interaction between the primary and secondary coolants (lead-water interaction). The analysed initiator event, as already mentioned, is a large break (up to a double ended guillotine break) of one (or more) SG cooling tubes that may become severe enough to determine dangerous effects on the interested structures. In order to better simulate and perform the mentioned postulated SGTR accident sequence analyses, an appropriate numerical model with the available computing resources (FEM codes) was set up at the DIMNP of Pisa University. That model was used to evaluate the effects of the propagation of the blast pressure waves inside the SG structures, taking into account also the sloshing phenomenon that could be induced by the lead primary coolant motions. Therefore the SGTR effects study may be considered as a transient and non linear problem the solution of which provides the 'time histories' of hydrodynamic pressures and stresses on the reactor pressure vessel and internals walls. (author)

  7. ANALISIS KEJADIAN STEAM GENERATOR TUBE RUPTURE (SGTR BERDASARKAN SKENARIO MIHAMA UNIT 2

    Directory of Open Access Journals (Sweden)

    Andi Sofrany Ekariansyah

    2015-03-01

    Full Text Available Pada tanggal 9 Februari 1991, terjadi kecelakaan putusnya pipa pemanas pembangkit uap (Steam Generator Tube Rupture/SGTR pada PLTN Mihama Unit 2. Dari kejadian tersebut, diperoleh catatan sekuensi kecelakaan berupa aktuasi sistem proteksi dan fitur keselamatan terekayasa dalam memitigasi kebocoran dari sistem primer ke sistem sekunder. Urutan sekuensi tersebut kemudian diterapkan pada PWR standar Jepang untuk disimulasikan menggunakan program perhitungan RELAP5/SCDAP/Mod3.2. Tujuannya untuk mengevaluasi konsekuensi yang terjadi bila kecelakaan tersebut terjadi pada PWR standar Jepang. Parameter yang dibandingkan adalah laju alir kebocoran, perubahan tekanan primer dan sekunder dan perubahan level di dalam pressurizer. Hasil simulasi menunjukkan perbedaan lama waktu kejadian SGTR hingga berhentinya kebocoran yang berlangsung lebih pendek pada PWR standar Jepang. Selain itu jumlah pendingin primer yang bocor dan jumlah uap yang terlepas dari MSRV tercatat lebih besar daripada PWR Mihama unit 2. Karakter aliran kebocoran, fluktuasi tekanan primer, dan level pressurizer sedikit berbeda pada tahap-tahap awal kejadian, namun relatif sama pada tahap akhir ketika aliran kebocoran dapat dihentikan. Hasil simulasi juga menunjukkan perlunya tindakan operator secara manual yang ditunjukkan dari isolasi sistem air umpan bantu (AFW pada pembangkit uap yang bocor, aktuasi katup pelepas uap (MSRV pada pembangkit uap yang utuh dan aktuasi auxiliary spray dan power operated relief valve (PORV pada pressurizer untuk mengantisipasi kejadian sebagai bagian dari prosedur operasi darurat. Kata kunci: SGTR, PWR Mihama Unit 2, PWR standar Jepang   On February 9,1991, a Steam Generator Tube Rupture (SGTR took place at the Mihama Unit No. 2. From that event, the accident sequence representing the actuation of protection system and engineered safety feature to mitigate the leak from primary system to secondary system is recorded. That sequence is then applied on the

  8. Risk assessment of severe accident-induced steam generator tube rupture

    International Nuclear Information System (INIS)

    1998-03-01

    This report describes the basis, results, and related risk implications of an analysis performed by an ad hoc working group of the U.S. Nuclear Regulatory Commission (NRC) to assess the containment bypass potential attributable to steam generator tube rupture (SGTR) induced by severe accident conditions. The SGTR Severe Accident Working Group, comprised of staff members from the NRC's Offices of Nuclear Reactor Regulation (NRR) and Nuclear Regulatory Research (RES), undertook the analysis beginning in December 1995 to support a proposed steam generator integrity rule. The work drew upon previous risk and thermal-hydraulic analyses of core damage sequences, with a focus on the Surry plant as a representative example. This analysis yielded new results, however, derived by predicting thermal-hydraulic conditions of selected severe accident scenarios using the SCDAP/RELAP5 computer code, flawed tube failure modeling, and tube failure probability estimates. These results, in terms of containment bypass probability, form the basis for the findings presented in this report. The representative calculation using Surry plant data indicates that some existing plants could be vulnerable to containment bypass resulting from tube failure during severe accidents. To specifically identify the population of plants that may pose a significant bypass risk would require more definitive analysis considering uncertainties in some assumptions and plant- and design-specific variables. 46 refs., 62 figs., 37 tabs

  9. ARTIST: a cooperative safety project to study fission product retention in a ruptured steam generator

    International Nuclear Information System (INIS)

    Guentay, S.; Dehbi, A.; Suckow, D.; Birchley, J.

    2001-01-01

    Sequences such as a steam generator tube rupture (SGTR) with stuck-open relief valve represent a significant public risk by virtue of the open path for release of radioactivity. The release may be lessened by deposition of fission products on the steam generator (SG) tubes and other structures or by scrubbing in the secondary coolant. The absence of empirical data, the complexity of the geometry and controlling processes, however, make the retention difficult to quantify and credit for it is typically not taken in risk assessments. The ARTIST experimental program to be conducted at Paul Scherrer Institut, Switzerland, will simulate the flow and retention of aerosol-borne fission products in the SG secondary, and thus provide a unique database to support safety assessments and analytical models. The project, foreseen in seven phases, will study phenomena at the separate effect and integral levels, and also address accident management (AM) issues. The prescribed values of the controlling parameters (aerosol size, aerosol type, gas flow velocity, residence time, etc) cover the range expected in severe and design basis accident scenarios. (authors)

  10. ARTIST: An International Project Investigating Aerosol Retention in a Ruptured Steam Generator

    International Nuclear Information System (INIS)

    Guentay, S.; Dehbi, A.; Suckow, D.; Birchley, J.

    2002-01-01

    Steam generator tube ruptures (SGTR) with a concurrent stuck open safety relief valve are counted among the risk dominant accident sequences because of the potential for radioactive products to bypass the containment. Owing to the absence of relevant empirical data and the complexity of the geometry and controlling processes, the aerosol removal in the steam generator (SG) tubes and in the secondary side is not well understood. Therefore, little or no credit is usually taken for aerosol retention due to natural processes in the various components of a SG. To help reduce the uncertainties associated with fission product release following an SGTR sequence, the Paul Scherrer Institut has initiated an international experimental project to be performed in the ARTIST (AeRosol Trapping In a Steam generaTor) facility in the time period from 2002 to 2007. The ARTIST test section is a scaled model of a real SG, and is comprised of a 264-tube bundle with a maximum height of 3.8 m, as well as one full-size droplet separator and one full-size steam dryer. The ARTIST facility is capable of producing soluble and insoluble aerosols and entrain them at sonic gas flow rates (up to 0.25 kg/s, thus matching comparable values predicted by the codes. In addition, aerosols can be generated at prototypical concentrations (up to 5 g/m 3 ) and sizes (0.2-5 mm AMMD). State of the art instrumentation is used (Low-pressure impactors, photometers, on-line particle sizer, online droplet sizer, etc.). The ARTIST project will simulate the flow and retention of aerosol-borne fission products in the SG, and provide a unique database to support safety assessments and analytical models. The project is foreseen in seven phases: 1) Aerosol retention in the tube under dry secondary side conditions, 2) Aerosol retention in the near field close to break under dry conditions, 3) Aerosol retention in the bundle far field under dry conditions, 4) Aerosol retention in the separator and dryer under dry

  11. 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

  12. 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

  13. 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

  14. 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

  15. The PSI Artist Project: Aerosol Retention and Accident Management Issues Following a Steam Generator Tube Rupture

    International Nuclear Information System (INIS)

    Guntay, Salih; Dehbi, Abdel; Suckow, Detlef; Birchley, Jon

    2002-01-01

    Steam generator tube rupture (SGTR) incidents, such as those, which occurred in various operating pressurized, water reactors in the past, are serious operational concerns and remain among the most risk-dominant events. Although considerable efforts have been spent to understand tube degradation processes, develop improved modes of operation, and take preventative and corrective measures, SGTR incidents cannot be completely ruled out. Under certain conditions, high releases of radionuclides to the environment are possible during design basis accidents (DBA) and severe accidents. The severe accident codes' models for aerosol retention in the secondary side of a steam generator (SG) have not been assessed against any experimental data, which means that the uncertainties in the source term following an un-isolated SGTR concurrent with a severe accident are not currently quantified. The accident management (AM) procedures aim at avoiding or minimizing the release of fission products from the SG. The enhanced retention of activity within the SG defines the effectiveness of the accident management actions for the specific hardware characteristics and accident conditions of concern. A sound database on aerosol retention due to natural processes in the SG is not available, nor is an assessment of the effect of management actions on these processes. Hence, the effectiveness of the AM in SGTR events is not presently known. To help reduce uncertainties relating to SGTR issues, an experimental project, ARTIST (Aerosol Trapping In a Steam generator), has been initiated at the Paul Scherrer Institut to address aerosol and droplet retention in the various parts of the SG. The test section is comprised of a scaled-down tube bundle, a full-size separator and a full-size dryer unit. The project will study phenomena at the separate effect and integral levels and address AM issues in seven distinct phases: Aerosol retention in 1) the broken tube under dry secondary side conditions, 2

  16. High-temperature deformation and rupture behavior of internally-pressurized Zircaloy-4 cladding in vacuum and steam enivronments

    International Nuclear Information System (INIS)

    Chung, H.M.; Garde, A.M.; Kassner, T.F.

    1977-01-01

    The high-temperature diametral expansion and rupture behavior of Zircaloy-4 fuel-cladding tubes have been investigated in vacuum and steam environments under transient-heating conditions that are of interest in hypothetical loss-of-coolant accident situations in light-water reactors. The effects of internal pressure, heating rate, axial constraint, and localized temperature nonuniformities in the cladding on the maximum circumferential strain have been determined for burst temperatures between approximately 650 and 1350 0 C

  17. Analysis of Communication between Main Control Room Operators in Decision-making Process in Steam Generator Tube Rupture Accident

    International Nuclear Information System (INIS)

    Petkov, M.; Petkov, G.

    2006-01-01

    The paper presents an investigation results for Main Control Room operators' reliability by Performance Evaluation of Teamwork method, based on FSS-1000 training archives in KNPP in case of Steam Generator Tube Rupture accident. The advantages of operators' teamwork are shown: a) group decision-making vs. individual one: b) positive influence of crew initiated communication consisting of orders and reports that are required by instruction. (authors)

  18. Analysis of steam generator tube rupture as a severe accident using MELCOR 1.8.4

    International Nuclear Information System (INIS)

    Yang Hongrun; Hidaka, Akihide; Sugimoto, Jun

    1999-03-01

    This report presents the results from the MELCOR 1.8.4 calculations for Steam Generator Tube Rupture (SGTR) with stuck open of all the safety valves in faulted SG as a severe accident. The calculations are based on Surry nuclear power plant. After performed using the once-through primary system model alone by 1.0x10 5 s, the calculations were conducted with both of the once-through and the hot leg countercurrent natural circulation models. The results, including event sequences, processes and progressions of core degradation, radionuclides release from core and reactor cavity, and source terms to the environment are described in detail. It is concluded that the availability of High Pressure Safety Injection (HPSI) can significantly delay the progression of core heat-up and approximately 7% of cesium iodide (CsI) can be released to the environment directly through the stuck open safety valve. Comparisons between the results from the two models are also given in this report. The present analyses also showed that during SGTR accident, the hot leg countercurrent natural circulation flow cannot be established well and therefore it has little effect on the mitigation of the core degradation. (author)

  19. Simulation of a loss of coolant accident with rupture in the steam generator hot collector

    International Nuclear Information System (INIS)

    1991-03-01

    The Central Research Institute for Physics of the Hungarian Academy of Sciences designed and constructed the PMK-NVH test facility, a scaled down model of the WWER-440 Paks nuclear power plant. Hungary made the PMK-NVH facility available to the IAEA. The IAEA, having identified the need for experimental data due to the difficulties of building integral test facilities and the high costs of these experiments, has accepted the offer of the Hungarian Academy of Sciences and has organized three standard problem exercises. In these exercises, experimental data from the simulation of loss of coolant accidents were compared with analytical predictions of the behaviour of the facility, calculated with computer codes. The third standard problem exercise involved a test, in which the rupture was simulated to occur at the top of the hot collector of the steam generator, therefore creating a leak from primary to secondary side. Both hydroaccumulators and high pressure injection were allowed to actuate as prescribed in the actual plant. Eighteen organizations from 15 Member States took part in the exercise presenting pre-test and some post-test analyses which were discussed in a final meeting in Vienna in August, 1990. This document presents a complete overview of the third standard problem exercise, including description of the facility, the experiment, the codes and models used by the participants and a detailed intercomparison of calculated and experimental results. It is recognized that code assessment is a long process which involves many interrelated steps; therefore, no general conclusion or optimum code or best model was reached. However, the exercise was recognized as an important contributor to code validation. 42 refs, figs and tabs

  20. Thermal-hydraulic analysis of SMART steam generator tube rupture using TASS/SMR-S code

    International Nuclear Information System (INIS)

    Kim, Hee-Kyung; Kim, Soo Hyoung; Chung, Young-Jong; Kim, Hyeon-Soo

    2013-01-01

    Highlights: ► The analysis was performed from the viewpoint of primary coolant leakage. ► The thermal hydraulic responses and the maximum leakage have been identified. ► There is no direct release into the atmosphere caused by an SGTR accident. ► SMART safety system works well against an SGTR accident. - Abstract: A steam generator tube rupture (SGTR) accident analysis for SMART was performed using the TASS/SMR-S code. SMART with a rated thermal power of 330 MWt has been developed at the Korea Atomic Energy Research Institute. The TASS/SMR-S code can analyze the thermal hydraulic phenomena of SMART in a full range of reactor operating conditions. An SGTR is one of the most important accidents from a thermal hydraulic and radiological viewpoint. A conservative analysis against a SMART SGTR was performed. The major concern of this analysis is to find the thermal hydraulic responses and maximum leakage amount from a primary to a secondary side caused by an SGTR accident. A sensitivity study searching for the conservative thermal hydraulic conditions, break locations, reactivity and other conditions was performed. The dominant parameters related with the integral leak are the high RCS pressure, low core inlet coolant temperature and low break location of the SG cassette. The largest integral leak comes to 28 tons in the most conservative case during 1 h. But there is no direct release into the atmosphere because the secondary system pressure is maintained with a sufficient margin for the design pressure. All leaks go to the condenser. The analysis results show that the primary and secondary system pressures are maintained below the design pressure and the SMART safety system is working well against an SGTR accident

  1. Remaining life assessment of carbon steel boiler headers by repeated creep testing

    Energy Technology Data Exchange (ETDEWEB)

    Drew, M. [ANSTO, Materials and Engineering Science, New Illawarra Road, Lucas Heights, PMB 1 Menai, NSW 2234 (Australia)]. E-mail: michael.drew@ansto.gov.au; Humphries, S. [ANSTO, Materials and Engineering Science, New Illawarra Road, Lucas Heights, PMB 1 Menai, NSW 2234 (Australia); Thorogood, K. [ANSTO, Materials and Engineering Science, New Illawarra Road, Lucas Heights, PMB 1 Menai, NSW 2234 (Australia); Barnett, N. [BlueScope Steel, P.O. Box 1854, Wollongong, NSW (Australia)

    2006-05-15

    The condition of carbon steel boiler headers that have been in service for over 25 years has been assessed periodically by NDT, dimensional measurements, replication and accelerated creep testing. Historical temperature records were limited, so estimates of effective header temperatures were made from replicas. These estimates were compared with header stub thermocouple readings. At about 280,000 service hours, samples were chain-drilled from the headers for accelerated creep testing. These test results indicated that the headers had satisfactory remaining life. Nine years after the original samples were taken, additional samples were removed from one header at 337,000 service hours. The creep rupture properties measured from the repeated tests were almost identical to the initial results. A mild degree of random, nodular graphite was found in the samples and its effect on creep properties is discussed.

  2. Break size effect on the transient thermal-hydraulic behavior during the steam generator tube rupture accident

    International Nuclear Information System (INIS)

    Kang, K.H.; Park, H.S.; Cho, S.; Choi, N.H.; Chu, I.C.; Yun, B.J.; Kim, K.D.; Kim, Y.S.; Baek, W.P.; Choi, K.Y.

    2011-01-01

    In order to simulate the SGTR accident of the APR1400, integral effect tests were performed by simulating a double-ended rupture of a single and five U-tubes. Following the reactor trip, the primary system pressure decreased and the secondary system pressure increased until the MSSVs was opened to reduce the secondary system pressure. Break area affected the timings of the major events observed in the tests. Less heat transfer to the secondary side caused by earlier actuation of the safety injection pumps had more influence on the secondary pressure of the affected steam generator than the break flow. (author)

  3. Coincident steam generator tube rupture and stuck-open safety relief valve carryover tests: MB-2 steam generator transient response test program

    International Nuclear Information System (INIS)

    Garbett, K.; Mendler, O.J.; Gardner, G.C.; Garnsey, R.; Young, M.Y.

    1987-03-01

    In PWR steam generator tube rupture (SGTR) faults, a direct pathway for the release of radioactive fission products can exist if there is a coincident stuck-open safety relief valve (SORV) or if the safety relief valve is cycled. In addition to the release of fission products from the bulk steam generator water by moisture carryover, there exists the possibility that some primary coolant may be released without having first mixed with the bulk water - a process called primary coolant bypassing. The MB-2 Phase II test program was designed specifically to identify the processes for droplet carryover during SGTR faults and to provide data of sufficient accuracy for use in developing physical models and computer codes to describe activity release. The test program consisted of sixteen separate tests designed to cover a range of steady-state and transient fault conditions. These included a full SGTR/SORV transient simulation, two SGTR overfill tests, ten steady-state SGTR tests at water levels ranging from very low levels in the bundle up to those when the dryer was flooded, and three moisture carryover tests without SGTR. In these tests the influence of break location and the effect of bypassing the dryer were also studied. In a final test the behavior with respect to aerosol particles in a dry steam generator, appropriate to a severe accident fault, was investigated

  4. Effect of steam corrosion on HTGR core support post strength loss. Part II. Consequences of steam generator tube rupture event

    International Nuclear Information System (INIS)

    Wichner, R.P.

    1977-01-01

    To perform the assessment, a series of eight tube-rupture events of varying severity and probability were postulated. Case 1 pertains to the situation where the moisture detection, loop isolation, and dump procedures function as planned; the remaining seven cases suppose various defects in the moisture detection system, the core auxiliary coolant system, and the integrity of the prestressed concrete reactor vessel. Core post burnoffs beneath three typical fuel zones were estimated for each postulated event from the determined impurity compositions and core post temperature history. Two separate corrosion rate expressions were assumed, as deemed most appropriate of those published for the high-oxidant level typical in tube rupture events. It was found that the nominal core post beneath the highest power factor fuel zone would lose from 0.02 to 2.5 percent of their strength, depending on an assumed corrosion rate equation and the severity of the event. The effect of hot streaking during cooldown was determined by using preliminary estimates of its magnitude. It was found that localized strength loss beneath the highest power factor zone ranges from 0.23 to 12 percent, assuming reasonably probable hot-streaking circumstances. The combined worst case, hot streaking typical for a load-following transient and most severe accident sequence, yields an estimated strength loss of from 25 to 33 percent for localized regions beneath the highest power factor zones

  5. Analysis of multiple-tube ruptures in both steam generators for the Three Mile Island-1 pressurized water reactor

    International Nuclear Information System (INIS)

    Nassersharif, B.

    1985-01-01

    The operator guidelines were followed for both transients described. Both transients resulted in SG overfill and the tube-rupture flow did not terminate in either transient. The following statements can be deducted from the results of the calculations: the tube-rupture flow could not be stopped for either case during 2600 s (43 min) of transient time; each accident scenario resulted in SG overfill; both SGs overfilled by 1600 s (27 min) and 1800 s (30 min) for Cases 1 and 2, respectively; conditions for isolation of the SGs were not reached; and core subcooling was not lost in either case but the upper head was voided in Case 2. Comparison of the cooldown rates in the two cases after 1200 s (20 min) shows that these rates are equal (i.e., restart of the RCPs did not change the primary-system cooldown rate). However, in Case 2, a steam bubble was formed in the upper head, which did not disappear during the simulated time. One of the immediate actions in the guidelines was to fill both SGs to 95% level. This step was almost unnecessary because the tube-rupture flow was large enough that the 15.4 - mK/s (100 - 0 F h) cooldown-rate limit was exceeded and AFW could not be injected. Also, the guidelines did not address the SG overfill issue

  6. Rupture

    CERN Multimedia

    Association du personnel

    2006-01-01

    Our Director-General is indifferent to the tradition of concertation foreseen in our statutes and is "culturally" unable to associate the Staff Association with problem-solving in staff matters. He drags his heels as long as possible before entering into negotiations, presents "often misleading" solutions at the last minute which he only accepts to change once a power struggle has been established. Faced with this rupture and despite its commitment to concertation between gentlemen. The results of the poll in which the staff is invited to participate this week. We therefore need your support to state our claims to the Governing Bodies. The Staff Association proposes a new medium of communication and thus hopes to show that it is ready for future negotiations. The pages devoted to the Staff Association are presented in a more informative, reactive and factual manner and in line with the evolution of the social situation at CERN. We want to establish strong and continuous ties between the members of CERN and ou...

  7. 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.)

  8. Experimental evaluation of emergency operating procedures on multiple steam generator tube rupture in INER integral system test facility

    International Nuclear Information System (INIS)

    Liu, T.J.; Lin, Y.M.; Lee, C.H.; Chang, C.Y.; Hong, W.T.

    1997-01-01

    The multiple steam generator tube rupture (SGTR) scenario in Westinghouse type pressurized water reactor (PWR) has been investigated at the Institute of Nuclear Energy Research (INER) Integral System Test (IIST) facility. This reduced-height and reduced-pressure test facility was designed to simulate the main features of Maanshan nuclear power plant. The SGTR test scenario assumes the double-ended break of one-, two- and six- tubes without other failures. The major operator actions follow the related symptom-oriented Emergency Operating Procedure (EOP) on the reference plant. This study focuses on the investigation of thermal-hydraulics phenomena and the adequacy of associated EOP to limit primary-to-secondary leakage. Through this study, it is found that the adequacy of current EOP in minimizing the radioactivity release demands early substantial operator involvement, especially in the multi-tubes break events. Also, the detailed mechanism of the main thermal-hydraulic phenomena during the SGTR transient are explored. (author)

  9. The effect of the number of condensed phases modeled on aerosol behavior during an induced steam generator tube rupture sequence

    International Nuclear Information System (INIS)

    Bixler, N.E.; Schaperow, J.H.

    1998-06-01

    VICTORIA is a mechanistic computer code designed to analyze fission product behavior within a nuclear reactor coolant system (RCS) during a severe accident. It provides detailed predictions of the release of radioactive and nonradioactive materials from the reactor core and transport and deposition of these materials within the RCS. A recently completed independent peer review of VICTORIA, while confirming the overall adequacy of the code, recommended a number of modeling improvements. One of these recommendations, to model three rather than a single condensed phase, is the focus of the work reported here. The recommendation has been implemented as an option so that either a single or three condensed phases can be treated. Both options have been employed in the study of fission product behavior during an induced steam generator tube rupture sequence. Differences in deposition patterns and mechanisms predicted using these two options are discussed

  10. Steam generators

    International Nuclear Information System (INIS)

    Hayden, R.L.J.

    1979-01-01

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

  11. Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, G.E.; Fletcher, C.D.; Davis, C.B. [and others

    1997-06-01

    This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and Ranking Table (PIRT) efforts for the Westinghouse AP600 reactor. The purpose of this PIRT is to identify important phenomena so that they may be addressed in both the experimental programs and the RELAP5/MOD3 systems analysis computer code. In Revision of this report, the responses of AP600 during small break loss-of-coolant accident, main steam line break, and steam generator tube rupture accident scenarios were evaluated by a committee of thermal-hydraulic experts. Committee membership included Idaho National Engineering and Environmental Laboratory staff and recognized thermal-hydraulic experts from outside of the laboratory. Each of the accident scenarios was subdivided into separate, sequential periods or phases. Within each phase, the plant behavior is controlled by, at most, a few thermal-hydraulic processes. The committee identified the phenomena influencing those processes, and ranked & influences as being of high, medium, low, or insignificant importance. The primary product of this effort is a series of tables, one for each phase of each accident scenario, describing the thermal-hydraulic phenomena judged by the committee to be important, and the relative ranking of that importance. The rationales for the phenomena selected and their rankings are provided. This document issue incorporates an update of the small break loss-of-coolant accident portion of the report. This revision is the result of the release of experimental evidence from AP600-related integral test facilities (ROSA/AP600, OSU, and SPES) and thermal-hydraulic expert review. The activities associated with this update were performed during the period from June 1995 through November 1996. 8 refs., 26 figs., 42 tabs.

  12. Phenomena identification and ranking tables for Westinghouse AP600 small break loss-of-coolant accident, main steam line break, and steam generator tube rupture scenarios

    International Nuclear Information System (INIS)

    Wilson, G.E.; Fletcher, C.D.; Davis, C.B.

    1997-06-01

    This report revision incorporates new experimental evidence regarding AP600 behavior during small break loss-of-coolant accidents. This report documents the results of Phenomena Identification and Ranking Table (PIRT) efforts for the Westinghouse AP600 reactor. The purpose of this PIRT is to identify important phenomena so that they may be addressed in both the experimental programs and the RELAP5/MOD3 systems analysis computer code. In Revision of this report, the responses of AP600 during small break loss-of-coolant accident, main steam line break, and steam generator tube rupture accident scenarios were evaluated by a committee of thermal-hydraulic experts. Committee membership included Idaho National Engineering and Environmental Laboratory staff and recognized thermal-hydraulic experts from outside of the laboratory. Each of the accident scenarios was subdivided into separate, sequential periods or phases. Within each phase, the plant behavior is controlled by, at most, a few thermal-hydraulic processes. The committee identified the phenomena influencing those processes, and ranked ampersand influences as being of high, medium, low, or insignificant importance. The primary product of this effort is a series of tables, one for each phase of each accident scenario, describing the thermal-hydraulic phenomena judged by the committee to be important, and the relative ranking of that importance. The rationales for the phenomena selected and their rankings are provided. This document issue incorporates an update of the small break loss-of-coolant accident portion of the report. This revision is the result of the release of experimental evidence from AP600-related integral test facilities (ROSA/AP600, OSU, and SPES) and thermal-hydraulic expert review. The activities associated with this update were performed during the period from June 1995 through November 1996. 8 refs., 26 figs., 42 tabs

  13. Pressurization rate effect on ligament rupture and burst pressures of cracked steam generator tubes

    International Nuclear Information System (INIS)

    Majumdar, S.; Kasza, K.

    2009-01-01

    The question of whether ligament rupture pressure or unstable burst pressure may vary significantly with pressurization rate at room temperature arose from the results of pressure tests by industry on tubes with machined part-throughwall notches. Slow (quasi-static) and fast 14 MPa/s (2000 psi/s) pressurization rate tests on specimens with nominally the same notch geometry appeared to show a significant effect of the rate of pressurization on the unstable burst pressure. Unfortunately, the slow and fast loading rate tests were conducted following two different test procedures, which could confound the results. The current series of tests were conducted on a variety of specimen geometries using a consistent test procedure to better establish the effect of pressurization rate on ligament rupture and burst pressures. (author)

  14. Pressurization rate effect on ligament rupture and burst pressures of cracked steam generator tubes

    Energy Technology Data Exchange (ETDEWEB)

    Majumdar, S.; Kasza, K. [Argonne National Laboratory, Nuclear Energy Division, Lemont, Illinois (United States)

    2009-07-01

    The question of whether ligament rupture pressure or unstable burst pressure may vary significantly with pressurization rate at room temperature arose from the results of pressure tests by industry on tubes with machined part-throughwall notches. Slow (quasi-static) and fast 14 MPa/s (2000 psi/s) pressurization rate tests on specimens with nominally the same notch geometry appeared to show a significant effect of the rate of pressurization on the unstable burst pressure. Unfortunately, the slow and fast loading rate tests were conducted following two different test procedures, which could confound the results. The current series of tests were conducted on a variety of specimen geometries using a consistent test procedure to better establish the effect of pressurization rate on ligament rupture and burst pressures. (author)

  15. Steam generator arrangement

    International Nuclear Information System (INIS)

    Ssinegurski, E.

    1981-01-01

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

  16. Analysis of a postulated pipe rupture and subsequent check valve slam of a PWR feedwater line

    International Nuclear Information System (INIS)

    Chang, K.C.; Adams, T.M.

    1983-01-01

    System designs criteria employed in the design of pressurized water reactors (PWR) requires that, for a postulated instantaneous guillotine rupture anywhere in the steam generator feedwater system, no more than one steam generator can be allowed to blowdown. Feedwater systems in many PWR's consist of pipe lines running from the feedwater pumps into a common feedwater header then branching into each steam generator from the header. The feedwater piping to each steam generator contains swing check valves to prevent reverse flow from the steam generator. This activation of some or all of these check valves significantly complicates the system structural analysis in that not only the blowdown forces resulting from the postulated pipe rupture, but also the water hammer loads resulting from closure of the check valve at high reverse flow velocities must be considered. The loads resulting from system blowdown and check valve closure are axial in nature. Peak loads ranging from 130000 lbs. to 180000 lbs. are not uncommon and are layout dependent. The analysis and design to withstand this transient loading deviates from the usual feedwater line design in that supports are required along the piping axis in the direction normal to the usual seismic supports. A brief and general discussion of the methods employed in the generation of the thermal-hydraulic loadings is presented. However, the discussion emphasizes the piping and piping support structural design and analysis method and approaches used in evaluating a selected portion of such a feedwater system. (orig./RW)

  17. Development of tube rupture evaluation code for FBR steam generator (II). Modification of heat transfer model in sodium side

    International Nuclear Information System (INIS)

    Hamada, H.; Kurihara, A.

    2003-05-01

    The thermal effect of sodium-water reaction jet on neighboring heat transfer tubes was examined to rationally evaluate the structural integrity of the tube for overheating rupture under a water leak in an FBR steam generator. Then, the development of new heat transfer model and the application analysis were carried out. Main results in this paper are as follows. (1) The evaluation method of heat flux and heat transfer coefficient (HTC) on the tube exposed to reaction jet was developed. By using the method, it was confirmed that the heat flux could be realistically evaluated in comparison with the previous method. (2) The HTC between reaction jet and the tube was theoretically examined in the two-phase flow model, and new heat transfer model considering the effect of fluid temperature and cover gas pressure was developed. By applying the model, a tentative experimental correlation was conservatively obtained by using SWAT-1R test data. (3) The new model was incorporated to the Tube Rupture Evaluation Code (TRUE), and the conservatism of the model was confirmed by using sodium-water reaction data such as the SWAT-3 tests. (4) In the application analysis of the PFR large leak event, there was no significant difference of calculation results between the new model and previous one; the importance of depressurization in the tube was confirmed. (5) In the application analysis of the Monju evaporator, it was confirmed that the calculation result in the previous model would be more conservative than that in the new one and that the maximum cumulative damage of 25% could be reduced in the new model. (author)

  18. A Comparison of Nuclear Power Plant Simulator with RELAP5/MOD3 code about Steam Generator Tube Rupture

    International Nuclear Information System (INIS)

    Kim, Sung Hyun; Moon, Chan Ki; Park, Sung Baek; Na, Man Gyun

    2013-01-01

    The RELAP5/MOD3 code introduced in cooperation with U. S. NRC has been utilized mainly for validation calculation of accident analysis submitted by licensee in Korea. The Korea Institute of Nuclear Safety has built a verification system of LWR accident analysis with RELAP5/MOD3 code engine. Therefore, the simulator replicates the design basis accident and its results are compared with RELAP5/MOD3 code results that will have important implications in the verification of the simulator in the future. The SGTR simulations were performed by the simulator and its results were compared with ones by RELAP5/MOD3 code in this study. Thus, the results of this study can be used as materials to build the verification system of the nuclear power plant simulator. We tried to compare with RELAP5/MOD3 verification code by replicating major parameters of steam generator tube rupture using the simulator for OPR-1000 in Yonggwang training center. By comparing the changes in temperature, pressure and inventory of the reactor coolant system and main steam system during the SGTR, it was confirmed that the main behaviors of SGTR which the simulator and RELAP5/MOD3 code showed are similar. However, the behavior of SG pressure and level that are important parameters to diagnose the accident were a little different. We estimated that RELAP5/MOD3 code was not reflected the major control systems in detail, such as FWCS, SBCS and PPCS. The different behaviors of SG level and pressure in this study should be needed an additional review. As a result of the comparison, the major simulation parameters behavior by RELAP5/MOD3 code agreed well with the one by the simulator. Therefore, it is thought that RELAP5/MOD3 code is used as a tool for validation of NPP simulator in the near future through this study

  19. A reappraisal of steam generator tube rupture in the French licensing process

    International Nuclear Information System (INIS)

    Conte, M.; Gouffon, A.; Moriette, P.

    1984-10-01

    Upon the examination of the safety options submitted by EDF (Electricite de France) for a new pressurized water reactor design (N4, 1400 MWe), the French safety authorities decided that the conventionnal list of events to take under consideration should be amended as follows: failure of 1 and 2 steam generator tubes. To meet these objectives, design improvements were decided and new operating criteria were required by the technical specifications. Various preventive measures have been adopted by EDF to reduce tube degradation risks at the design stage, at the secondary feedwater quality level, and concerning also the quality control. The radiological consequences of generator tube integrity failure can be mitigated if the primary coolant activity is low, the tube flow detection is rapid, the release time is short, and the operating procedure is suitable and easily implemented [fr

  20. Application of dynamic probabilistic safety assessment approach for accident sequence precursor analysis: Case study for steam generator tube rupture

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Han Sul; Heo, Gyun Young [Kyung Hee University, Yongin (Korea, Republic of); Kim, Tae Wan [Incheon National University, Incheon (Korea, Republic of)

    2017-03-15

    The purpose of this research is to introduce the technical standard of accident sequence precursor (ASP) analysis, and to propose a case study using the dynamic-probabilistic safety assessment (D-PSA) approach. The D-PSA approach can aid in the determination of high-risk/low-frequency accident scenarios from all potential scenarios. It can also be used to investigate the dynamic interaction between the physical state and the actions of the operator in an accident situation for risk quantification. This approach lends significant potential for safety analysis. Furthermore, the D-PSA approach provides a more realistic risk assessment by minimizing assumptions used in the conventional PSA model so-called the static-PSA model, which are relatively static in comparison. We performed risk quantification of a steam generator tube rupture (SGTR) accident using the dynamic event tree (DET) methodology, which is the most widely used methodology in D-PSA. The risk quantification results of D-PSA and S-PSA are compared and evaluated. Suggestions and recommendations for using D-PSA are described in order to provide a technical perspective.

  1. Application of dynamic probabilistic safety assessment approach for accident sequence precursor analysis: Case study for steam generator tube rupture

    International Nuclear Information System (INIS)

    Lee, Han Sul; Heo, Gyun Young; Kim, Tae Wan

    2017-01-01

    The purpose of this research is to introduce the technical standard of accident sequence precursor (ASP) analysis, and to propose a case study using the dynamic-probabilistic safety assessment (D-PSA) approach. The D-PSA approach can aid in the determination of high-risk/low-frequency accident scenarios from all potential scenarios. It can also be used to investigate the dynamic interaction between the physical state and the actions of the operator in an accident situation for risk quantification. This approach lends significant potential for safety analysis. Furthermore, the D-PSA approach provides a more realistic risk assessment by minimizing assumptions used in the conventional PSA model so-called the static-PSA model, which are relatively static in comparison. We performed risk quantification of a steam generator tube rupture (SGTR) accident using the dynamic event tree (DET) methodology, which is the most widely used methodology in D-PSA. The risk quantification results of D-PSA and S-PSA are compared and evaluated. Suggestions and recommendations for using D-PSA are described in order to provide a technical perspective

  2. Long-term rupture strength as a criterion of operational durability of steam line metal

    International Nuclear Information System (INIS)

    Gofman, Yu.M.

    2000-01-01

    The method for substantiation of the steam line service life prolongation, depending on the achieved level of the metal vulnerability to damage, is proposed. The methodology for evaluating the metal state is developed on the basis of the durability bond with the level of the vulnerability to damage through micropores and the ferrite dislocation structure state. The main changes in the metal at the 1-3 stages of its creep are presented. The micropores are absent at the 1 stage. the micropores of about 0.1 μm in diameter are identified at the beginning of the 2 stage. The ferrite grains on the transition from the 2 to the 3 creep stage are mainly fragmentary. There takes place further micropores growth on the grain boundaries up to 1 - 3 μm. Significant number of recrystallized volumes in the ferrite is observed at the 3 creep stage. The number of micropores of 1 - 3μm in size sharply increases, and, as a rule, chains of micropores are observed. The pores of 5 μm in size are formed at the pre-destruction stage, the fusion whereof leads to microcracks formation [ru

  3. Evaluation of a main steam line break with induced, multiple tube ruptures: A comparison of NUREG 1477 (Draft) and transient methodologies Palo Verde Nuclear Generating Station

    Energy Technology Data Exchange (ETDEWEB)

    Parrish, K.R.

    1995-09-01

    This paper presents the approach taken to analyze the radiological consequences of a postulated main steam line break event, with one or more tube ruptures, for the Palo Verde Nuclear Generating Station. The analysis was required to support the restart of PVNGS Unit 2 following the steam generator tube rupture event on March 14, 1993 and to justify continued operation of Units 1 and 3. During the post-event evaluation, the NRC expressed concern that Unit 2 could have been operating with degraded tubes and that similar conditions could exist in Units 1 and 3. The NRC therefore directed that a safety assessment be performed to evaluate a worst case scenario in which a non-isolable main steam line break occurs inducing one or more tube failures in the faulted steam generator. This assessment was to use the generic approach described in NUREG 1477, Voltage-Based Interim Plugging Criteria for Steam Generator Tubes - Task Group Report. An analysis based on the NUREG approach was performed but produced unacceptable results for off-site and control room thyroid doses. The NUREG methodology, however, does not account for plant thermal-hydraulic transient effects, system performance, or operator actions which could be credited to mitigate dose consequences. To deal with these issues, a more detailed analysis methodology was developed using a modified version of the Combustion Engineering Plant Analysis Code, which examines the dose consequences for a main steam line break transient with induced tube failures for a spectrum equivalent to 1 to 4 double ended guillotine U-tube breaks. By incorporating transient plant system responses and operator actions, the analysis demonstrates that the off-site and control room does consequences for a MSLBGTR can be reduced to acceptable limits. This analysis, in combination with other corrective and recovery actions, provided sufficient justification for continued operation of PVNGS Units 1 and 3, and for the subsequent restart of Unit 2.

  4. Evaluation of a main steam line break with induced, multiple tube ruptures: A comparison of NUREG 1477 (Draft) and transient methodologies Palo Verde Nuclear Generating Station

    International Nuclear Information System (INIS)

    Parrish, K.R.

    1995-01-01

    This paper presents the approach taken to analyze the radiological consequences of a postulated main steam line break event, with one or more tube ruptures, for the Palo Verde Nuclear Generating Station. The analysis was required to support the restart of PVNGS Unit 2 following the steam generator tube rupture event on March 14, 1993 and to justify continued operation of Units 1 and 3. During the post-event evaluation, the NRC expressed concern that Unit 2 could have been operating with degraded tubes and that similar conditions could exist in Units 1 and 3. The NRC therefore directed that a safety assessment be performed to evaluate a worst case scenario in which a non-isolable main steam line break occurs inducing one or more tube failures in the faulted steam generator. This assessment was to use the generic approach described in NUREG 1477, Voltage-Based Interim Plugging Criteria for Steam Generator Tubes - Task Group Report. An analysis based on the NUREG approach was performed but produced unacceptable results for off-site and control room thyroid doses. The NUREG methodology, however, does not account for plant thermal-hydraulic transient effects, system performance, or operator actions which could be credited to mitigate dose consequences. To deal with these issues, a more detailed analysis methodology was developed using a modified version of the Combustion Engineering Plant Analysis Code, which examines the dose consequences for a main steam line break transient with induced tube failures for a spectrum equivalent to 1 to 4 double ended guillotine U-tube breaks. By incorporating transient plant system responses and operator actions, the analysis demonstrates that the off-site and control room does consequences for a MSLBGTR can be reduced to acceptable limits. This analysis, in combination with other corrective and recovery actions, provided sufficient justification for continued operation of PVNGS Units 1 and 3, and for the subsequent restart of Unit 2

  5. Conceptual design of a chickpea harvesting header

    Directory of Open Access Journals (Sweden)

    H. Golpira

    2013-07-01

    Full Text Available Interest in the development of stripper headers is growing owing to the excessive losses of combine harvesters and costs of manually harvesting for chickpeas. The design of a new concept can enhance the mechanized process for chickpea harvesting. A modified stripper platform was designed, in which passive fingers with V-shape slots removes the pods from the anchored plant. The floating platform was accompanied by a reel to complete the harvesting header. Black-box modeling was used to redesign the functional operators of the header followed by an investigation of the system behavior. Physical models of the platform and reel were modified to determine the crucial variables of the header arrangement during field trials. The slot width was fixed at 40 mm, finger length at 40 mm, keyhole diameter at 10 mm and entrance width at 6 mm; the batted reel at peripheral diameter of 700 mm and speed at 50 rpm. A tractor-mounted experimental harvester was built to evaluate the work quality of the stripper header. The performance of the prototype was tested with respect to losses and results confirmed the efficiency of the modified stripper header for chickpea harvesting. Furthermore, the header with a 1.4 m working width produced the spot work rates of 0.42 ha h-1.

  6. Potential steam generator tube rupture in the presence of severe accident thermal challenge and tube flaws due to foreign object wear

    International Nuclear Information System (INIS)

    Liao, Y.; Guentay, S.

    2009-01-01

    This study develops a methodology to assess the probability for the degraded PWR steam generator to rupture first in the reactor coolant pressure boundary, under severe accident conditions with counter-current natural circulating high temperature gas in the hot leg and SG tubes. The considered SG tube flaws are caused by foreign object wear, which in recent years has emerged as a major inservice degradation mechanism for the new generation tubing materials. The first step develops the statistical distributions for the flaw frequency, size, and the flaw location with respect to the tube length and the tube's tubesheet position, based on data of hundreds of flaws reported in numerous SG inservice inspection reports. The next step performs thermal-hydraulic analysis using the MELCOR code and recent CFD findings to predict the thermal challenge to the degraded tubes and the tube-to-tube difference in thermal response at the SG entrance. The final step applies the creep rupture models in the Monte Carlo random walk to test the potential for the degraded SG to rupture before the surge line. The mean and range of the SG tube rupture probability can be applied to estimate large early release frequency in probabilistic safety assessment.

  7. HTGR power plant hot reheat steam pressure control system

    International Nuclear Information System (INIS)

    Braytenbah, A.S.; Jaegtnes, K.O.

    1975-01-01

    A control system for a high temperature gas cooled reactor (HTGR) power plant is disclosed wherein such plant includes a plurality of steam generators. Dual turbine-generators are connected to the common steam headers, a high pressure element of each turbine receiving steam from the main steam header, and an intermediate-low pressure element of each turbine receiving steam from the hot reheat header. Associated with each high pressure element is a bypass line connected between the main steam header and a cold reheat header, which is commonly connected to the high pressure element exhausts. A control system governs the flow of steam through the first and second bypass lines to provide for a desired minimum steam flow through the steam generator reheater sections at times when the total steam flow through the turbines is less than such minimum, and to regulate the hot reheat header steam pressure to improve control of the auxiliary steam turbines and thereby improve control of the reactor coolant gas flow, particularly following a turbine trip. (U.S.)

  8. 20% inlet header break analysis of Advanced Heavy Water Reactor

    International Nuclear Information System (INIS)

    Srivastava, A.; Gupta, S.K.; Venkat Raj, V.; Singh, R.; Iyer, K.

    2001-01-01

    The proposed Advanced Heavy Water Reactor (AHWR) is a 750 MWt vertical pressure tube type boiling light water cooled and heavy water moderated reactor. A passive design feature of this reactor is that the heat removal is achieved through natural circulation of primary coolant at all power levels, with no primary coolant pumps. Loss of coolant due to failure of inlet header results in depressurization of primary heat transport (PHT) system and containment pressure rise. Depressurization activates various protective and engineered safety systems like reactor trip, isolation condenser and advanced accumulator, limiting the consequences of the event. This paper discusses the thermal hydraulic transient analysis for evaluating the safety of the reactor, following 20% inlet header break using RELAP5/MOD3.2. For the analysis, the system is discretized appropriately to simulate possible flow reversal in one of the core paths during the transient. Various modeling aspects are discussed in this paper and predictions are made for different parameters like pressure, temperature, steam quality and flow in different parts of the Primary Heat Transport (PHT) system. Flow and energy discharges into the containment are also estimated for use in containment analysis. (author)

  9. Assessment of RELAP5/MOD3.1 with the LSTF SB-SG-06 experiment simulating a steam generator tube rupture transient

    International Nuclear Information System (INIS)

    Seul, K.W.; Bang, Y.S.; Lee, S.; Kim, H.J.

    1996-09-01

    The objective of the present work is to identify the predictability of RELAP5/MOD3.1 regarding thermal-hydraulic behavior during a steam generator tube rupture (SGTR). To evaluate the computed results, LSTF SB-SG-06 test data simulating the SGTR that occurred at the Mihama Unit 2 in 1991 are used. Also, some sensitivity studies of the code change in RELAP5, the break simulation model, and the break valve discharge coefficient are performed. The calculation results indicate that the RELAP5/MOD3.1 code predicted well the sequence of events and the major phenomena during the transient, such as the asymmetric loop behavior, reactor coolant system (RCS) cooldown and heat transfer by natural circulation, the primary and secondary system depressurization by the pressurizer auxiliary spray and the steam dump using the intact loop steam generator (SG) relief valve, and so on. However, there are some differences from the experimental data in the number of the relief valve cycling in the affected SG, and the flow regime of the hot leg with the pressurizer, and the break flow rates. Finally, the calculation also indicates that the coolant in the core could remain in a subcooled state as a result of the heat transfer caused by the natural circulation flow even if the reactor coolant pumps (RCPs) turned off and that the affected SG could be properly isolated to minimize the radiological release after the SGTR

  10. Assessment of RELAP5/MOD3.1 with the LSTF SB-SG-06 experiment simulating a steam generator tube rupture transient

    Energy Technology Data Exchange (ETDEWEB)

    Seul, K.W.; Bang, Y.S.; Lee, S.; Kim, H.J. [Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of)

    1996-09-01

    The objective of the present work is to identify the predictability of RELAP5/MOD3.1 regarding thermal-hydraulic behavior during a steam generator tube rupture (SGTR). To evaluate the computed results, LSTF SB-SG-06 test data simulating the SGTR that occurred at the Mihama Unit 2 in 1991 are used. Also, some sensitivity studies of the code change in RELAP5, the break simulation model, and the break valve discharge coefficient are performed. The calculation results indicate that the RELAP5/MOD3.1 code predicted well the sequence of events and the major phenomena during the transient, such as the asymmetric loop behavior, reactor coolant system (RCS) cooldown and heat transfer by natural circulation, the primary and secondary system depressurization by the pressurizer auxiliary spray and the steam dump using the intact loop steam generator (SG) relief valve, and so on. However, there are some differences from the experimental data in the number of the relief valve cycling in the affected SG, and the flow regime of the hot leg with the pressurizer, and the break flow rates. Finally, the calculation also indicates that the coolant in the core could remain in a subcooled state as a result of the heat transfer caused by the natural circulation flow even if the reactor coolant pumps (RCPs) turned off and that the affected SG could be properly isolated to minimize the radiological release after the SGTR.

  11. Analysis of steam-generator tube-rupture events combined with auxiliary-feedwater control-system failure for Three Mile Island-Unit 1 and Zion-Unit 1 pressurized water reactors

    International Nuclear Information System (INIS)

    Nassersharif, B.

    1986-01-01

    A steam-generator tube-rupture (SGTR) event combined with loss of all offsite alternating-current power and failure of the auxiliary-feedwater (AFW) control system has been investigated for the Three Mile Island-Unit 1 (TMI-1) and Zion-Unit 1 (Zion-1) pressurized water reactors. The Transient Reactor Analysis Code was used to simulate the accident sequence for each plant. The objectives of the study were to predict the plant transient response with respect to tube-rupture flow termination, extent of steam generator overfill, and thermal-hydraulic conditions in the steam lines. Two transient cases were calculated: (1) a TMI-1 SGTR and runaway-AFW transient, and (2) a Zion-1 SGTR and runaway-AFW transient. Operator actions terminated the tube-rupture flow by 1342 s (22.4 min) and 1440 s (24.0 min) for TMI-1 and Zion-1, respectively, but AFW injection was continued. The damaged steam generator (DSG) overfilled by 1273 s (21.2 min) for the TMI-1 calculation and by 1604 s (26.7 min) for the Zion-1 calculation. The DSG steam lines were completely filled by 1500 s (25 min) and 2000 s (33.3 min) for TMI-1 and Zion-1, respectively. The maximum subcooling in the steam lines was approx.63 K (approx.113 0 F) for TMI-1 and approx.44 K (approx.80 0 F) for Zion-1

  12. Multi-protocol header generation system

    Energy Technology Data Exchange (ETDEWEB)

    Roberts, David A.; Ignatowski, Michael; Jayasena, Nuwan; Loh, Gabriel

    2017-09-05

    A communication device includes a data source that generates data for transmission over a bus, and a data encoder that receives and encodes outgoing data. An encoder system receives outgoing data from a data source and stores the outgoing data in a first queue. An encoder encodes outgoing data with a header type that is based upon a header type indication from a controller and stores the encoded data that may be a packet or a data word with at least one layered header in a second queue for transmission. The device is configured to receive at a payload extractor, a packet protocol change command from the controller and to remove the encoded data and to re-encode the data to create a re-encoded data packet and placing the re-encoded data packet in the second queue for transmission.

  13. Surry Power Station secondary water chemistry improvement since steam generator replacement and the unit two feedwater pipe rupture

    International Nuclear Information System (INIS)

    Swindell, E.T.

    1988-01-01

    Surry Power Station has two Westinghouse-designed three-loop PWRs of 811 MWe design rating. The start of commercial operation was in July, 1972 in No.1 plant, and May, 1973 in No.2 plant. Both plants began the operation using controlled phosphate chemistry for the steam generators. In 1975, both plants were converted to all volatile treatment on the secondary side due to the tube wall thinning corrosion in the steam generators, which was associated with the phosphate sludge that was building up on the tube sheets and created acidic condition. Thereafter, condenser and air leakage and steam generator denting occurred, and after the operation of 8 years 2 month of No.1 plant and 5 years 9 months of No.2 plant, the steam generators were replaced. A major plant improvement program was designed and implemented from 1979 to 1980. The improvement in new steam generators, the modification for preventing corrosion, the addition of a steam generator blowdown recovery system, the reconstruction of condensers, the installation of full flow, deep bed condensate polishers, the addition of Dionex 8,000 on-line ion chromatograph, a long term maintenance agreement with Westinghouse and so on are reported. (Kako, I.)

  14. A fast prediction of plant behaviour in the steam generator tube rupture accident at Mihama unit 2 using a similar case

    International Nuclear Information System (INIS)

    Gofuku, Akio; Tanaka, Yutaka; Numoto, Atsushi; Yoshikawa, Hidekazu.

    1996-01-01

    It is important to predict fast and accurately future trend of behaviour of a nuclear power plant in an emergency situation. The case-based reasoning is a strong tool for this purpose because it solves a problem by effectively using past similar cases. This study investigates the applicability of the case-based reasoning as a fast prediction technique of plant behaviour. This paper discusses a prediction of initial plant behaviour in the steam generator tube rupture accident happened at the Mihama nuclear power plant unit 2 by using the behaviour data of an accident of the same type happened at Prairie Island nuclear power plant unit 1. The prediction results coincide well with the reported plant behaviour although there are several important differences in the detailed plant specifications and operator actions between the two SGTR accidents. (author)

  15. Rupture of a high pressure gas or steam pipe in a tunnel: a preliminary investigation of the jet thrust exerted on a tunnel barrier

    International Nuclear Information System (INIS)

    Baum, M.R.

    1988-04-01

    On power plant, if a high pressure pipe containing high temperature gas or steam were to rupture, sensitive equipment necessary for safety shutdown of the plant could possibly be incapacitated if exposed to the subsequent high temperature environment. In many plant configurations the high pressure pipework is contained in tunnels where it is possible to construct barriers which isolate one section of the plant from another, thereby restricting the spread of the high temperature fluid/air mixture. This paper describes a preliminary experimental investigation of the magnitude of the thrust likely to be exerted on such barriers by a gas jet issuing from the failed pipe. Measurements of the thrust exerted on a flat plate by normal impingement of a highly underexpanded gas jet are in agreement with a semi-quantitative analysis assuming conservation of the axial momentum of the jet. (author)

  16. Evaluation of PWR response to main-steamline break with concurrent steam-generator tube rupture and small-break LOCA

    International Nuclear Information System (INIS)

    Laaksonen, J.T.; Sheron, B.W.

    1982-12-01

    In 1980, the NRC staff raised a potential safety issue involving a coincident steamline break, steam generator tube rupture, and small-break loss-of-coolant accident (LOCA). The bases for this concern were that the system response, primarily the maintenance of core cooling, was unanalyzed and the adequacy of the present guidance to operators to respond to combination LOCAs was unknown. This report discusses the staff evaluations performed to assess the system response and the adequacy of the present emergency operator guidelines. In all of the analyzed cases the primary coolant shrinkage, caused by overcooling, and the simultaneous loss of coolant can be compensated by the high pressure emergency core cooling system. The core remains covered with liquid, and the primary coolant remains subcooled, except in the vessel upper head. If the steamline break is outside the containment and cannot be isolated, the radiological consequences could be more severe than in any accident currently analyzed in a typical plant Final Safety Evaluation Report (FSAR). To decrease the risk of elevated offsite releases, an early diagnosis of the tube rupture has to be ensured. This can be done by upgrading operator instructions. The appropriate mitigating actions are in the existing instructions

  17. The study on water ingress mass in the steam generator heat-exchange tube rupture accident of modular high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Wang Yan; Shi Lei; Li Fu; Zheng Yanhua

    2012-01-01

    The steam generator heat-exchange tube rupture (SGTR) accident is an important and particular accident which will result in water ingress to the primary loop of reactor. Water ingress will result in chemical reaction of graphite fuel and structure with water, which may cause overpressure due to generation of explosive gaseous in large quantity. The study on the water ingress accident is significant for the verification of the inherent characteristics of high temperature gas-cooled reactor. The previous research shows that the amount of water ingress mass is the dominant key factor on the severity of the accident consequence. The 200 MWe high temperature gas-cooled reactor (HTR-PM), which is the first modular pebble-bed high temperature gas-cooled reactor in China designed by the Institute of Nuclear and New Energy Technology of Tsinghua University, is selected to be analyzed in this paper. The different DBA accident scenarios of double-ended break of single heat-exchange tube are simulated respectively by the thermal-hydraulic analysis code RETRAN-02. The results show the water ingress mass through the broken heat-exchange tube is related to the break location. The amount of water ingress mass is affected obviously by the capacity of the emptier system. With the balance of safety and economical efficiency, the amount of water ingress mass from the secondary side of steam generator into the primary coolant loop will be reduced by increasing properly the diameter of the draining lines. (authors)

  18. Severe accident analysis of a steam generator tube rupture accident using MAAP-CANDU to support level 2 PSA for the Point Lepreau Generating Station Refurbishment Project

    Energy Technology Data Exchange (ETDEWEB)

    Petoukhov, S.M.; Brown, M.J. [Canadian Nuclear Laboratories, Chalk River, ON (Canada)

    2015-07-01

    A Level 2 Probabilistic Safety Assessment was performed for the Point Lepreau Generating Station. The MAAP-CANDU code was used to simulate the progression of postulated severe core damage accidents and fission product releases. This paper discusses the results for the reference case of the Steam Generator Tube Rupture initiating event. The reference case, dictated by the Level 2 Probabilistic Safety Assessment, was extreme and assumed most safety-related plant systems were not available: all steam generator feedwater; the emergency water supply; the moderator, shield and shutdown cooling systems; and all stages of emergency core cooling. The reference case also did not credit any post Fukushima lessons or any emergency mitigating equipment. The reference simulation predicted severe core damage beginning at 3.7 h, containment failure at 6.4 h, moderator boil off by 8.2 h, and calandria vessel failure at 42 h. A total release of 5.3% of the initial inventory of radioactive isotopes of Cs, Rb and I was predicted by the end of the simulation (139 h). Almost all noble gas fission products were released to the environment, primarily after the containment failure. No hydrogen/carbon monoxide burning was predicted. (author)

  19. Discussion on amount of water ingress mass in steam generator heat-exchange tube rupture accident of high- temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Wang Yan; Zheng Yanhua; Shi Lei; Li Fu; Sun Ximing

    2009-01-01

    The steam generator heat-exchange tube rupture (SGTR) accident which will result in the water ingress to the primary circuit of reactor is an important and particular accident for high-temperature gas-cooled reactor (HTGR). The analysis of the water ingress accident is significant for verifying the inherent safety characteristics of HTGR. The amount of water ingress mass is one of the decisive factors for the seriousness of the accident consequence. The 250 MW Pebble-bed Modular High-Temperature Gas-cooled Reactor (HTR-PM) designed by Institute of Nuclear and New Energy Technology of Tsinghua University was selected as an example of analysis. The analysis results show that the amount of water ingress mass is not only affected directly with the broken position and the broken area of the tubes, but also related with the diameter of draining piping and restrictor, draining control valve, action setting of emptier system. With reasonable parameters chosen, the water in steam generator could be drained effectively, so it will prevent the primary circuit of reactor from water ingress in large quantity and reduce the radioactive isotopes ingress to the secondary circuit. (authors)

  20. Pressure transients resulting from sodium-water reaction following a large leak in LMFBR steam generator

    International Nuclear Information System (INIS)

    Rajput, A.K.

    1984-01-01

    The study of sodium water reaction, following a large leak, concerns primarily with the estimation of pressure/flow transients that are developed in the steam generator and the associated secondary circuit. This paper describes the mathematical formulations used in SWRT (Sodium Water Reaction Transients) code developed to estimate such pressure transients for FBTR plant. The results, obtained using SWRT have been presented for a leak in economiser (20m from bottom water header) and for a leak in super heater portions. A time lag of 50 m sec was considered for rupture disc takes to burst once the pressure experienced by it exceeds the set value. Also described in annexure to this paper is the mathematical formulation for two phase transient flow for the better estimation of leak rate from the ruptured end of the damaged heat transfer tube. This leak model considers slip but assumes thermal equilibrium between the liquid and vapour phases

  1. 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

  2. Testing Header Component of Electricity Power Industry Boiler

    International Nuclear Information System (INIS)

    Soedardjo, S.A; Andryansyah, B; Artahari, Dewi; Natsir, Muhammad; Triyadi, Ari; Farokhi

    2000-01-01

    Testing of header component of Suralaya Unit II electricity power by replication method has been carried out. That header component is cross over pipe which interconnection between Primary and Superheater Outlet Header Secondary Superheater Outlet Header with the operation time over 14 years. The main composition of cross over pipe is 2 1/4 Cr 1 Mo or frequently specified as ferritique steel. The replication testing shown that the damage classification on those cross over pipe in A class based on failure classification from Neubauer and Wedel. Simple calculation in favor of cross over pipe remaining lifetime is about 16.5 years moreover

  3. Application of the visual system analyzer (ViSA): simulation of the steam generator tube rupture event at Ulchin unit 4

    International Nuclear Information System (INIS)

    Lee, S.W.; Kim, K.D.; Hwang, M.K.; Jeong, J.J.

    2004-01-01

    Korea Atomic Energy Research Institute (KAERI) has developed the Visual System Analyzer (ViSA) based on the best-estimate (B-E) codes, MARS and RETRAN-3D. The key features of ViSA are: (1) The use of the same input and the same level of accuracy as the original codes is guaranteed (2) Users can design their own plant mimic by a drag-and-drop from the provided indicators (3) The on-line interactive control enables users to simulate the operator's actions (4) The nodalization window is designed to display the transient temperature and void distributions. ViSA is composed of two parts; the B-E code with plant input and the Graphic User Interface (GUI) that includes the plant mimic and an interactive control function, etc. The calculation results of the B-E code are transferred to a user via the GUI and a user can apply the operator action to the B-E code using an interactive control function. Therefore, it is not necessary to prepare complex control input data to simulate the various manual operations which may occur during the plant transient. In this study, the Steam Generator Tube Rupture (SGTR) Accident, which occurred at Ulchin Unit 4 in April 2002, has been simulated using ViSA and the simulation results are compared with the measured plant data. The RETRAN-3D plant input data used in this simulation is a genetic input deck prepared for the simulation from a normal operation condition to a Small-Break LOCA. From the results of the SGTR simulation, we found that the GUI functions of ViSA and the input data for Ulchin Unit 4 have enough effectiveness and soundness. (author)

  4. Threats and surprises behind IPv6 extension headers

    NARCIS (Netherlands)

    Hendriks, Luuk; Velan, Petr; de Oliveira Schmidt, Ricardo; De Boer, Pieter Tjerk; Pras, Aiko

    2017-01-01

    The concept of Extension Headers, newly introduced with IPv6, is elusive and enables new types of threats in the Internet. Simply dropping all traffic containing any Extension Header - a current practice by operators-seemingly is an effective solution, but at the cost of possibly dropping legitimate

  5. Detection and Repair of Ligament Cracks in a 109mm Thick Superheater Outlet Header

    International Nuclear Information System (INIS)

    Day, Peter

    2006-01-01

    Conventional thermal power station boilers are constructed of drums and a series of headers which are interconnected with many hundreds of tubes. Typically feed water enters the boiler at about 250 deg C at a pressure of around 250 bar with steam outlet temperatures of 540 deg C and a pressure of 170 bar. Superheater outlet headers may be subjected to quite arduous conditions during service. Not only are they exposed to high pressure stresses but also to high thermal stresses due to varying thermal gradients through the section thickness particularly at start up and during two shift operation. The area that is exposed to the greatest thermal gradients is the narrow ligament that exists between the tube hole penetrations in the header bore. In the mid the 1980's industry wide surveys found cracking in a large percentage (25-50%) of headers after 15 years of service. Detection and sizing of ligament cracking and estimates of the rate of growth are therefore a major consideration especially in plant that is two shifted. In order to manage the risk both remote visual and ultrasonic inspection are performed during each major unit overhaul. Conclusion: Ultrasonic techniques used for this inspection need to be carefully evaluated with respect to their effectiveness. Conventional pulse echo is capable of detection but using for example a technique such as AS2207 level 1 will not show the defect size. Time of flight diffraction has shown itself to be effective in accurately sizing ligament cracking. However the complex geometry of header ligaments appears to cause a narrowing of the beam with the effect that crack tip responses can be concentrated at the centre of the ligament. Therefore great care needs to be taken during data interrogation because errors in sizing can occur. Wherever possible both 'B' and 'D' scan data should be collected. It appears that the greatest accuracy is obtained with respect to defect growth from the B scan image. With respect to the welding a

  6. 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

  7. Nuclear power plant and apparatus for superheating steam

    International Nuclear Information System (INIS)

    Schluderberg, D.C.

    1983-01-01

    The invention consists of an apparatus for superheating steam, the apparatus comprising a horizontally disposed generally cylindrical elongate shell, inlet means in the shell for receiving steam, outlet means in the shell for discharching the steam, and a bundle of inclined tubes positioned in the flow path of the steam, each of the tubes having a length which is less than the diameter of the shell and opening into and extending in an upward direction from an outlet header to an inlet header, the inlet header beeing connected to a source of vapor, and the outlet header beeing connected to a condensate drain, characterised in that the test bundle comprises two banks of the tubes, the angle at which each of the tubes of one of the banks extends relative to a vertical longitudinal centerplane, the tubes of one of the banks terminate at and open into the inlet header, and the tubes of the other banks terminate at an open into another inlet header

  8. Simulation of main steam and feedwater system of full scope simulator for Qinshan 300 MW Nuclear Power Unit

    International Nuclear Information System (INIS)

    Zhao Xiaoyu

    1996-01-01

    The simulation of main steam and feedwater system is the most important and maximal part in secondary circuit model, including all of main steam and feedwater's thermal-hydraulic properties, except heat-exchange of secondary side of steam generator. It simulates main steam header, steam power in each stage of turbine, moisture separator-reheater, deaerator, condenser, high pressure and low pressure heater, auxiliary feedwater and main steam bypass in full scope

  9. Packet Header Compression for the Internet of Things

    Directory of Open Access Journals (Sweden)

    Pekka KOSKELA

    2016-01-01

    Full Text Available Due to the extensive growth of Internet of Things (IoT, the number of wireless devices connected to the Internet is forecasted to grow to 26 billion units installed in 2020. This will challenge both the energy efficiency of wireless battery powered devices and the bandwidth of wireless networks. One solution for both challenges could be to utilize packet header compression. This paper reviews different packet compression, and especially packet header compression, methods and studies the performance of Robust Header Compression (ROHC in low speed radio networks such as XBEE, and in high speed radio networks such as LTE and WLAN. In all networks, the compressing and decompressing processing causes extra delay and power consumption, but in low speed networks, energy can still be saved due to the shorter transmission time.

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  11. Technology Corner: Analysing E-mail Headers For Forensic Investigation

    Directory of Open Access Journals (Sweden)

    M. Tariq Banday

    2011-06-01

    Full Text Available Electronic Mail (E-Mail, which is one of the most widely used applications of Internet, has become a global communication infrastructure service.  However, security loopholes in it enable cybercriminals to misuse it by forging its headers or by sending it anonymously for illegitimate purposes, leading to e-mail forgeries. E-mail messages include transit handling envelope and trace information in the form of structured fields which are not stripped after messages are delivered, leaving a detailed record of e-mail transactions.  A detailed header analysis can be used to map the networks traversed by messages, including information on the messaging software and patching policies of clients and gateways, etc. Cyber forensic e-mail analysis is employed to collect credible evidence to bring criminals to justice. This paper projects the need for e-mail forensic investigation and lists various methods and tools used for its realization. A detailed header analysis of a multiple tactic spoofed e-mail message is carried out in this paper. It also discusses various possibilities for detection of spoofed headers and identification of its originator. Further, difficulties that may be faced by investigators during forensic investigation of an e-mail message have been discussed along with their possible solutions.

  12. Dose surveys in two digital mammography units using DICOM headers

    International Nuclear Information System (INIS)

    Tsalafoutas, I.; Michalaki, C.; Papagiannopoulou, C.; Efstathopoulos, E.

    2012-01-01

    Background and objective: Digital mammography units store images in DICOM format. Thus, data regarding the acquisition parameters are available within DICOM headers, including among others, the anode/filter combination, tube potential and tube current exposure time product, compressed breast thickness, entrance surface air kerma (ESAK) and mean glandular dose (MGD). However, manual extraction of these data for the verification of the displayed values' accuracy and for dose survey purposes is time consuming. Our objective was to develop a method that enables the automation of such procedures. Materials and methods: Two hundred mammographic examinations (800 mammograms) performed in two digital units (GE, Essential) were recorded on CD-roms. Using appropriate software (DICOM Info Extractor) all dose related DICOM headers were extracted into a Microsoft Excel based spreadsheet, containing embedded algorithms for the calculation of ESAK and MGD according to Dance et al (Phys. Med. Biol. 45, 2000) methodology. Results: The ESAK and MGD values stored in the DICOM headers were compared with those calculated and in most cases were within ±10%. The basic difference among the two mammographic units is that, the older one calculates MGD assuming a breast composition 50% glandular-50% adipose tissue, while the newer one calculates the actual breast glandularity and stores this value in a DICOM header. The average MGD values were 1.21 mGy and 1.38 mGy, respectively. Conclusion: For the units studied, the ESAK and MGD values stored in DICOM headers are reliable. Utilizing tools for their automatic extraction provides an easy way to perform dose surveys. (authors)

  13. Creep-rupture, steam oxidation and recovery behaviours upon dynamic transients up to 1300 C of cold-worked 304 stainless steel tubes dedicated to nuclear core fuel cladding

    International Nuclear Information System (INIS)

    Portier, L.; Brachet, J.C.; Vandenberghe, V.; Guilbert, T.; Lezaud-Chaillioux, V.; Bernard, C.; Rabeau, V.

    2011-01-01

    An ambitious mechanical tests program was conducted on the fuel rod cladding of the CABRI facility between 2004 and 2009 to re-evaluate the cladding tubes materials behaviour. As an offspring of this major scientific investment several conclusions of interest could be drawn on the 304 stainless steel material. In particular, the specific behaviour of the materials during hypothetical and extreme 'dry-out' conditions was investigated. In such a scenario, the cladding tube materials should experience a very brief incursion at high temperatures, in a steam environment, up to 1300 C, before cladding rewetting. Some of the measurements performed in the range of interest for the safety case were on purpose developed beyond the conservatively safe domain. Some of the results obtained for these non-conventional heating rates, pressures and temperature ranges will be presented. First in order to assess the high temperature creep-rupture material behaviour under internal pressure upon dynamic transient conditions, tests have been performed on cold-worked 304 stainless cladding tubes in a steam environment, for heating rates up to 100 C*s -1 and pressure ramp rates up to 10 bar*s -1 thanks to the use of the EDGAR facility. Other tests performed at a given pressure allowed us to check the steady-state secondary creep rate of the materials in the 1100-1200 C temperature range. It was also possible to determine the rupture strength value and the failure mode as a function of the thermal and pressure loading history applied. It is worth noticing that, for very specific conditions, a surprising pure intergranular brittle failure mode of the clad has been observed. Secondly, in order to check the materials oxidation resistance of the materials, two-side steam oxidation tests have been performed at 1300 C, using the DEZIROX facility. It was shown that, thanks to the use of Ring Compression tests, the 304 cladding tube keeps significant ductility for oxidation times up to at least

  14. Analyses in support of installation of steam-dump-to-atmosphere valves at steam lines of the Dukovany NPP

    International Nuclear Information System (INIS)

    Kral, P.

    1998-01-01

    Four conservative analyses were carried out with a view to examining the cooldown capacity of the super-emergency feedwater pump (SEFWP) → steam generator (SG) → steam dump to atmosphere/main steam line (SDA/MSL) chain. This emergency cooldown capacity was investigated for a postulated accident associated with a main steam header break + main feedwater header break + closing of all main steam lines, and for an artificial accident with SCRAM + isolation of all MSLs + loss of feedwater. The RELAP5/MOD3.1 code and a detailed 3-loop input model of the Dukovany plant were employed. Conservative assumptions with respect to the initial reactor power, decay heat evolution, and other input parameters were applied. The results gave evidence that the capacity of both the 2SEFWP → 2SG → 2SDA/SG and 1SEFWP → 1SG → 1SDA/SG chains is sufficient for the decay heat to be removed from the reactor; however, a considerably long time allowing for a sufficient drop of the decay heat is necessary for a deep cooldown of the primary circuit. For the event encompassing main steam header break + main feedwater header break with isolation of all MSLs and with cooling by 2SEFWPs, a time-consuming calculation gave evidence of the feasibility of passing to the water-water regime and primary system cooldown to below 93 deg C in the hot legs

  15. CFD simulation for thermal mixing of a SMART flow mixing header assembly

    International Nuclear Information System (INIS)

    Kim, Young In; Bae, Youngmin; Chung, Young Jong; Kim, Keung Koo

    2015-01-01

    Highlights: • Thermal mixing performance of a FMHA installed in SMART is investigated numerically. • Effects of operating condition and discharge hole configuration are examined. • FMHA performance satisfies the design requirements under various abnormal conditions. - Abstract: A flow mixing header assembly (FMHA) is installed in a system-integrated modular advanced reactor (SMART) to enhance the thermal mixing capability and create a uniform core flow distribution under both normal operation and accident conditions. In this study, the thermal mixing characteristics of the FMHA are investigated for various steam generator conditions using a commercial CFD code. Simulations include investigations for the effects of FMHA discharge flow rate differences, turbulence models, and steam generator conditions. The results of the analysis show that the FMHA works effectively for thermal mixing in various conditions and makes the temperature difference at the core inlet decrease noticeably. We verified that the mixing capability of the FMHA is excellent and satisfies the design requirement in all simulation cases tested here

  16. Design and analysis of reactor headers for Narora Atomic Power Project

    International Nuclear Information System (INIS)

    Danak, M.R.

    1975-01-01

    Reactor header for Narora Atomic Power Reactor is a 400 mm O.D. 10 metres long pressure vessel in the primary coolant circuit connecting 153 feeders to PHT pumps or steam generators. The vessel dimensions are restricted are by containment philosophy. The outlet connections for pumps or steam generators are to be of the size of vessel diameter and DO/t ratio for the vessel is approximately 10. The design and stresses induced meet the code requirements except that at times it is difficult to get precise stress values in absence of certain data and lack of code or available literature giving practical approach to the problem. It can be seen that the 400 mm equal tees used as part of the vessel cannot be penetrated in the light of code reinforcement requirements. However if the tees have to penetrated to retain established feeder layout, it should be established experimentally or by some detailed stress analysis that it will meet the intent of code. (author)

  17. Behaviour of a PWR with core protection system (SSN) in case of accidents due to power failure, ATWS and steam generator rupture

    International Nuclear Information System (INIS)

    Boncompagni, S.; Fulceri, P.; Oriolo, F.

    1985-01-01

    The results of the analysis of the transient fallowing internal and external power failure, without scram, in the nuclear power plant of the Italian Unified Nuclear Project are examined. The availability of ECCS is excluded while the breakage of a tube in each steam generator is supposed, togheter with the presence of an original safety system known as SSN (core protection system). Computations have been performed by using Mark 6 RELAP4 code. The study of the transient and the physical model used are briefly illustrated. Finally the results achieved are analysed

  18. Ruptured eardrum

    Science.gov (United States)

    ... eardrum ruptures. After the rupture, you may have: Drainage from the ear (drainage may be clear, pus, or bloody) Ear noise/ ... doctor to see the eardrum. Audiology testing can measure how much hearing has been lost. Treatment You ...

  19. 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

  20. Summary report on underground road header environmental control.

    CSIR Research Space (South Africa)

    Belle, BK

    2002-01-01

    Full Text Available and on monitoring should be reassessed to take into consideration the recent findings and current international trends. 5 6. No conclusive results were obtained with regard to the use of a wet cutter head in conjunction with the Bank 2000 Road Header Dust... this response time interval is lower than the T90 response time, a good indication of the methane gas trends can be obtained. To protect the methane sensors from the harsh environment around an active RH, the 22 Custodian sensors were placed in polycarbonate...

  1. Condensation induced water hammer in steam supply system

    International Nuclear Information System (INIS)

    Andrews, P.B.; Antaki, G.A.; Rawls, G.B.; Gutierrez, B.J.

    1995-01-01

    The accidental mixing of steam and water usually leads to condensation induced water hammer. This phenomenon is not uncommon in the power and process industries, and is of particular concern due to the high energies which accompany steam transients. The paper discusses the conditions which lead to a recent condensation induced water hammer in a 150 psig steam supply header. The ensuing structural damage, inspection and repairs are described. Finally, a list of design, maintenance and operational cautions are presented to help minimize the potential for condensation induced water hammer in steam lines

  2. Condensation induced water hammer in steam supply system

    International Nuclear Information System (INIS)

    Andrews, P.B.; Antaki, G.A.; Rawls, G.B.; Gutierrez, B.J.

    1995-01-01

    The accidental mixing of steam and water usually leads to condensation induced water hammer. THis phenomenon is not uncommon in the power and process industries, and is of particular concern due to the high energies which accompany steam transients. The paper discusses the conditions which lead to a recent condensation induced water hammer in a 150 psig steam supply header. The insuing structural damage, inspection and repairs are described. Finally, a list of design cautions are presented to help minimize the potential for condensation induced water hammer in steam lines

  3. Development of rupture discs for the FBTR

    International Nuclear Information System (INIS)

    Chetal, S.C.; Raju, C.; Anandkumar, V.; Seetharaman, V.; Rajan, K.K.

    1984-01-01

    Rupture discs are required as a safety device for protecting the secondary sodium circuit and its components against high pressure surges due to accidental water steam leaks in sodium heated steam generator and the consequent sodium water reaction. For identical reasons, rupture discs are also required on the vessels used for decontamination of sodium components. Reverse buckling knife blade concept with austenitic stainless steel disc has been developed for the rupture disc assemblies required for Fast Breeder Test Reactor (FBTR). Hydroforming process without any die has been used for disc fabrication. One rupture disc assembly required for steam generator is undergoing sodium endurance test and has accumulated 4,500 hours. The present status of development work as demonstrated by room temperature experimental results as well as the scope for future work are discussed. (author)

  4. Double rupture disc experience

    International Nuclear Information System (INIS)

    1979-01-01

    Result of these observations, comparisons and evaluations can be summarized in the following list of concerns regarding the use of double rupture discs coupled to the liquid space of a steam generator that is subjected to a large leak sodium water reaction event. Single rupture disc show delayed collapse characteristics in LLTR Series I and double disc assemblies are presumed to be more complex with additional delay before opening to give pressure relief. Delayed failure increases pressures in the IHTS and must be adequately covered by design requirements. With CRBR design, the first disc may fail only partially reducing the loading on the second disc with the result that relief performance may not meet requirements

  5. Safety assessment and improvement of Ignalina NPP against downcomer ruptures outside Accident Localisation System

    International Nuclear Information System (INIS)

    Rimkevicius, S.; Urbonavicius, E.

    2002-01-01

    Accident Localisation System (ALS) of Ignalina NPP is a pressure suppression type confinement, designed to prevent the release of contaminated steam-water mixture to the environment in case of Loss-of-Coolant Accident (LOCA). One of the peculiarities of Ignalina NPP with RBMK-1500 reactors is that not all of the reactor coolant circuit is enclosed within ALS. Some part of downcomers, that connect Drum Separator (DS) and suction header of main circulation pump is located outside ALS. In case of downcomer rupture in DS compartment the discharge is not confined, but flows to the environment through the safety panels installed in the ceiling of DS compartments. Numerous safety analyses were performed to assess the safety of Ignalina NPP against downcomer break outside ALS, and results were used for different applications in order to improve the safety of the plant. This paper presents the overview of the performed analyses, recommendations raised and safety improvements made to enhance the safety level of NPP. One of the applications is to present the recommendations for safety improvement if maximal allowable pressure limits are exceeded. The calculations results demonstrate that in the case of two downcomers rupture in drum separators compartment the maximum permissible pressure in the reactor hall could be exceeded. The knock-out panels from the reactor hall to the environment were recommended and installed for reactor hall overpressure protection. The evaluation of the drainage system efficiency from DS compartments was performed. In this case the especial attention was paid to analyse the water collection and drainage system behaviour in long term after postulated breaks. The analysis results showed that the modernization of the drainage system prevents the accumulation of the released water in the compartments even in the case of two downcomer pipes ruptures, and decreases the release of radioactive fission products (FP) to the environment.(author)

  6. Rupture disc

    International Nuclear Information System (INIS)

    Newton, R.G.

    1977-01-01

    The intermediate heat transport system for a sodium-cooled fast breeder reactor includes a device for rapidly draining the sodium therefrom should a sodium-water reaction occur within the system. This device includes a rupturable member in a drain line in the system and means for cutting a large opening therein and for positively removing the sheared-out portion from the opening cut in the rupturable member. According to the preferred embodiment of the invention the rupturable member includes a solid head seated in the end of the drain line having a rim extending peripherally therearound, the rim being clamped against the end of the drain line by a clamp ring having an interior shearing edge, the bottom of the rupturable member being convex and extending into the drain line. Means are provided to draw the rupturable member away from the drain line against the shearing edge to clear the drain line for outflow of sodium therethrough

  7. Globe Rupture

    Directory of Open Access Journals (Sweden)

    Reid Honda

    2017-07-01

    Full Text Available History of present illness: A 46-year-old male presented to the emergency department (ED with severe left eye pain and decreased vision after tripping and striking the left side of his head on the corner of his wooden nightstand. The patient arrived as an inter-facility transfer for a suspected globe rupture with a protective eye covering in place; thus, further physical examination of the eye was not performed by the emergency physician in order to avoid further leakage of aqueous humor. Significant findings: The patient’s computed tomography (CT head demonstrated a deformed left globe, concerning for ruptured globe. The patient had hyperdense material in the posterior segment (see green arrow, consistent with vitreous hemorrhage. CT findings that are consistent with globe rupture may include a collapsed globe, intraocular air, or foreign bodies. Discussion: A globe rupture is a full-thickness defect in the cornea, sclera, or both.1 It is an ophthalmologic emergency. Globe ruptures are almost always secondary to direct perforation via a penetrating mechanism; however, it can occur due to blunt injury if the force generated creates sufficient intraocular pressure to tear the sclera.2 Globes most commonly rupture at the insertions of the intraocular muscles or at the limbus. They are associated with a high rate of concomitant orbital floor fractures.2,3 Possible physical examination findings include a shallow anterior chamber on slit-lamp exam, hyphema, and an irregular “teardrop” pupil. Additionally, a positive Seidel sign, which is performed by instilling fluorescein in the eye and then examining for a dark stream of aqueous humor, is indicative of a globe rupture.4 CT is often used to assess for globe rupture; finds of a foreign body, intraocular air, abnormal contour or volume of the globe, or disruption of the sclera suggest globe rupture.2 The sensitivity of CT scan for diagnosis of globe rupture is only 75%; thus, high clinical

  8. RObust header compression (ROHC) performance for multimedia transmission over 3G/4G wireless networks

    DEFF Research Database (Denmark)

    Fitzek, Frank; Rein, S.; Seeling, P.

    2005-01-01

    Robust Header Compression (ROHC) has recently been proposed to reduce the large protocol header overhead when transmitting voice and other continuous meadi over IP based control stacks in wireless networks. In this paper we evaluate the real-time transmission of GSM encoded voice and H. 26L encod...

  9. Phishtest: Measuring the Impact of Email Headers on the Predictive Accuracy of Machine Learning Techniques

    Science.gov (United States)

    Tout, Hicham

    2013-01-01

    The majority of documented phishing attacks have been carried by email, yet few studies have measured the impact of email headers on the predictive accuracy of machine learning techniques in detecting email phishing attacks. Research has shown that the inclusion of a limited subset of email headers as features in training machine learning…

  10. Ruptured Spleen

    Science.gov (United States)

    ... be caused by various underlying problems, such as mononucleosis and other infections, liver disease, and blood cancers. ... cause a ruptured spleen. For instance, people with mononucleosis — a viral infection that can cause an enlarged ...

  11. Researching of Covert Timing Channels Based on HTTP Cache Headers in Web API

    Directory of Open Access Journals (Sweden)

    Denis Nikolaevich Kolegov

    2015-12-01

    Full Text Available In this paper, it is shown how covert timing channels based on HTTP cache headers can be implemented using different Web API of Google Drive, Dropbox and Facebook  Internet services.

  12. Enabling IP Header Compression in COTS Routers via Frame Relay on a Simplex Link

    Science.gov (United States)

    Nguyen, Sam P.; Pang, Jackson; Clare, Loren P.; Cheng, Michael K.

    2010-01-01

    NASA is moving toward a networkcentric communications architecture and, in particular, is building toward use of Internet Protocol (IP) in space. The use of IP is motivated by its ubiquitous application in many communications networks and in available commercial off-the-shelf (COTS) technology. The Constellation Program intends to fit two or more voice (over IP) channels on both the forward link to, and the return link from, the Orion Crew Exploration Vehicle (CEV) during all mission phases. Efficient bandwidth utilization of the links is key for voice applications. In Voice over IP (VoIP), the IP packets are limited to small sizes to keep voice latency at a minimum. The common voice codec used in VoIP is G.729. This new algorithm produces voice audio at 8 kbps and in packets of 10-milliseconds duration. Constellation has designed the VoIP communications stack to use the combination of IP/UDP/RTP protocols where IP carries a 20-byte header, UDP (User Datagram Protocol) carries an 8-byte header, and RTP (Real Time Transport Protocol) carries a 12-byte header. The protocol headers total 40 bytes and are equal in length to a 40-byte G.729 payload, doubling the VoIP latency. Since much of the IP/UDP/RTP header information does not change from IP packet to IP packet, IP/UDP/RTP header compression can avoid transmission of much redundant data as well as reduce VoIP latency. The benefits of IP header compression are more pronounced at low data rate links such as the forward and return links during CEV launch. IP/UDP/RTP header compression codecs are well supported by many COTS routers. A common interface to the COTS routers is through frame relay. However, enabling IP header compression over frame relay, according to industry standard (Frame Relay IP Header Compression Agreement FRF.20), requires a duplex link and negotiations between the compressor router and the decompressor router. In Constellation, each forward to and return link from the CEV in space is treated

  13. 40 Gbit/s NRZ Packet-Length Insensitive Header Extraction for Optical Label Switching Networks

    DEFF Research Database (Denmark)

    Seoane, Jorge; Kehayas, E; Avramopoulos, H.

    2006-01-01

    A simple method for 40 Gbit/s NRZ header extraction based on envelope detection for optical label switching networks is presented. The scheme is insensitive to packet length and spacing and can be single-chip integrated cost-effectively......A simple method for 40 Gbit/s NRZ header extraction based on envelope detection for optical label switching networks is presented. The scheme is insensitive to packet length and spacing and can be single-chip integrated cost-effectively...

  14. Orthogonal transformations for change detection, Matlab code (ENVI-like headers)

    DEFF Research Database (Denmark)

    2007-01-01

    Matlab code to do (iteratively reweighted) multivariate alteration detection (MAD) analysis, maximum autocorrelation factor (MAF) analysis, canonical correlation analysis (CCA) and principal component analysis (PCA) on image data; accommodates ENVI (like) header files.......Matlab code to do (iteratively reweighted) multivariate alteration detection (MAD) analysis, maximum autocorrelation factor (MAF) analysis, canonical correlation analysis (CCA) and principal component analysis (PCA) on image data; accommodates ENVI (like) header files....

  15. Regulation of ageing steam generators

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  16. Mechanical design of a sodium heated steam generator

    International Nuclear Information System (INIS)

    Chetal, S.C.

    1975-01-01

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

  17. Indigenous development of rupture discs for FBTR (Paper No. 028)

    International Nuclear Information System (INIS)

    Chetal, S.C.; Raju, Chander; Anandkumar, V.; Seetharaman, V.

    1987-02-01

    Rupture discs are required as a safety device for protecting the secondary sodium circuit and its components against high pressure surges due to accidental water-steam leaks in sodium heated steam generator and the consequent sodium water reaction. For identical reasons, rupture discs are also required on the vessels used for decontamination of sodium components. As an import substitution of the costly items for the FBTR Project, development of the rupture disc assemblies has been in progress at Indira Gandhi Centre for Atomic Research, Kalpakkam. Reverse buckling knife blade concept with stainless steel disc has been taken up for development. Hydroforming process without any die has been selected for disc fabrication. One rupture disc assembly required for steam generator has been tested in sodium satisfactorily. (author). 4 tables, 5 figs

  18. Creep-fatigue monitoring system for header ligaments of fossil power plants

    International Nuclear Information System (INIS)

    Chen, K.L.; Deardorf, A.F.; Copeland, J.F.; Pflasterer, R.; Beckerdite, G.

    1993-01-01

    The cracking of headers (primary and secondary superheater outlet, and reheater outlet headers) at ligament locations is an important issue for fossil power plants. A model for crack initiation and growth has been developed, based on creep-fatigue damage mechanisms. This cracking model is included in a creep-fatigue monitoring system to assess header structural integrity under high temperature operating conditions. The following principal activities are required to achieve this goal: (1) the development of transfer functions and (2) the development of a ligament cracking model. The first task is to develop stress transfer functions to convert measured (monitored) temperatures, pressures and flow rates into stresses to be used to compute damage. Elastic three-dimensional finite element analyses were performed to study transient thermal stress behavior. The sustained pressure stress redistribution due to high temperature creep was studied by nonlinear finite element analyses. The preceding results are used to derive Green's functions and pressure stress gradient transfer functions for monitoring at the juncture of the tube with the header inner surface, and for crack growth at the ligaments. The virtual crack closure method is applied to derive a stress intensity factor K solution for a corner crack at the tube/header juncture. Similarly, using the reference stress method, the steady state creep crack growth parameter C * is derived for a header corner crack. The C * solution for a small corner crack in a header can be inserted directed into the available C t solution, along with K to provide the complete transient creep solution

  19. NRC concerns about steam generator tube U-bend failures

    International Nuclear Information System (INIS)

    Dillon, R.L.

    1981-01-01

    This paper concerns itself with genralized NRC regulatory policy regarding SGT failures and staff reports and opinions which may tend to influence the developing policy specific to U-bend failures. The most significant analysis at hand in predicting NRC policy on SGT U-bend failures is Marsh's Evaluation of Steam Generator Tube Rupture Events. Marsh sets out to describe and analyze the five steam generator tube ruptures that are known to NRC. All have occurred in the period 1975 to 1980

  20. Secured Hash Based Burst Header Authentication Design for Optical Burst Switched Networks

    Science.gov (United States)

    Balamurugan, A. M.; Sivasubramanian, A.; Parvathavarthini, B.

    2017-12-01

    The optical burst switching (OBS) is a promising technology that could meet the fast growing network demand. They are featured with the ability to meet the bandwidth requirement of applications that demand intensive bandwidth. OBS proves to be a satisfactory technology to tackle the huge bandwidth constraints, but suffers from security vulnerabilities. The objective of this proposed work is to design a faster and efficient burst header authentication algorithm for core nodes. There are two important key features in this work, viz., header encryption and authentication. Since the burst header is an important in optical burst switched network, it has to be encrypted; otherwise it is be prone to attack. The proposed MD5&RC4-4S based burst header authentication algorithm runs 20.75 ns faster than the conventional algorithms. The modification suggested in the proposed RC4-4S algorithm gives a better security and solves the correlation problems between the publicly known outputs during key generation phase. The modified MD5 recommended in this work provides 7.81 % better avalanche effect than the conventional algorithm. The device utilization result also shows the suitability of the proposed algorithm for header authentication in real time applications.

  1. 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

  2. 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

  3. 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

  4. 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.)

  5. Design and Evaluation of IP Header Compression for Cellular-Controlled P2P Networks

    DEFF Research Database (Denmark)

    Madsen, T.K.; Zhang, Qi; Fitzek, F.H.P.

    2007-01-01

    In this paper we advocate to exploit terminal cooperation to stabilize IP communication using header compression. The terminal cooperation is based on direct communication between terminals using short range communication and simultaneously being connected to the cellular service access point....... The short range link is than used to provide first aid information to heal the decompressor state of the neighboring node in case of a packet loss on the cellular link. IP header compression schemes are used to increase the spectral and power efficiency loosing robustness of the communication compared...

  6. Analysis of fast reactor steam generator performance

    International Nuclear Information System (INIS)

    Hulme, G.; Curzon, A.F.

    1992-01-01

    A computer model for the prediction of flow and temperature fields within a fast reactor steam generator unit is described. The model combines a commercially available computational fluid dynamics (CFD) solver (PHOENICS) with a steam-tube calculation and provides solutions for the fully coupled flow and temperature fields on both the shell side and the tube side. The model includes the inlet and outlet headers and the bottom end stagnant zone. It also accounts for the effects of support grids and edge-gaps. Two and three dimensional and transient calculations have been performed for both straight tube and J-tube units. Examples of the application of the model are presented. (7 figures) (Author)

  7. Verification of the computer code ATHLET in the framework of the external verification group ATHLET BETHSY test 9.3 - steam generator U-tube rupture with failure of the high pressure injection. Final report; Verifikation des ATHLET-Rechenprogramms im Rahmen der externen Verifikationsgruppe ATHLET BETHSY Test 9.3 - Heizrohrbruch mit Versagen der Hochdruck-Noteinspeisung. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Krepper, E.; Schaefer, F. [Forschungszentrum Rossendorf e.V. (FZR) (Germany). Inst. fuer Sicherheitsforschung

    1998-08-01

    In the framework of the external validation of the thermalhydraulic code ATHLET MOD 1.1 CYCLE D, which is being developed by the GRS, post test analyses of two experiments were done, which were performed at the French integral test facility BETHSY. During test 9.3 the consequences of a steam generator U-tube rupture with failure of the high pressure injection and of the auxiliary feedwater supply were investigated. As accident management measures, the depressurization of the secondary sides, first of the two intact steam generators, then of the damaged steam generator and finally the primary depressurization by opening of the pressurizer valve were performed. The results show, that the code ATHLET is able to describe the complex scenario in good accordance with the experiment. The safety relevant statement could be reproduced. Deviations, which did not impose the general results, occurred concerning the break mass flow during the depressurization of the damaged steam generator and the description of the failure of the heat transfer to the damaged steam generator. Reasons are hardly to find, because these processes are highly complex. (orig.) [Deutsch] Im Rahmen der externen Validierung des von der Gesellschaft fuer Anlagen- und Reaktorsicherheit entwickelten Stoerfallcodes ATHLET, der in der Version Mod 1.1 Cycle D vorlag, wurden zwei Experimente nachgerechnet und analysiert, die an der franzoesischen Versuchsanlage BETHSY durchgefuehrt wurden. Im Test 9.3 werden die Konsequenzen untersucht, wenn bei einem Heizrohrbruch die Hochdruckeinspeisung sowie die Not-Speisewasserversorgung der Dampferzeuger versagen und nur die Druckspeicher sowie die Niederdruckeinspeisung zur Verfuegung stehen. Als Accident Management Massnahmen wurde die sekundaere Druckentlastung und schliesslich die primaere Entlastung ueber den Druckhalter untersucht. Die Analyse kommt zu dem Ergebnis, dass der Code ATHLET in der Lage ist, dieses komplexe Szenario recht gut zu beschreiben. Die

  8. Qualifying the use of RIS data for patient dose by comparison with DICOM header data

    International Nuclear Information System (INIS)

    Wilde, R.; Charnock, P.; McDonald, S.; Moores, B. M.

    2011-01-01

    A system was developed in 2008 to calculate patient doses using Radiology Information System (RIS) data and presents these data as a patient dose audit. One of the issues with this system was the quality of user-entered data. It has been shown that Digital Imaging and Communication in Medicine (DICOM) header data can be used to perform dose audits with a high level of data accuracy. This study aims to show that using RIS data for dose audits is not only a viable alternative to using DICOM header data, but that it has advantages. A new system was developed to pull header data from DICOM images easily and was installed on a workstation within a hospital department. Data were recovered for a common set of examinations using both RIS and DICOM header data. The data were compared on a result-by-result basis to check for consistency of common fields between RIS and DICOM, as well as assessing the value of data fields uncommon to both systems. The study shows that whilst RIS is not as accurate as DICOM, it does provide enough accurate data and that it has other advantages over using a DICOM approach. These results suggest that a 'best of both worlds' may be achievable using Modality Performed Procedure Step (MPPS). (authors)

  9. Multiple-output all-optical header processing technique based on two-pulse correlation principle

    NARCIS (Netherlands)

    Calabretta, N.; Liu, Y.; Waardt, de H.; Hill, M.T.; Khoe, G.D.; Dorren, H.J.S.

    2001-01-01

    A serial all-optical header processing technique based on a two-pulse correlation principle in a semiconductor laser amplifier in a loop mirror (SLALOM) configuration that can have a large number of output ports is presented. The operation is demonstrated experimentally at a 10Gbit/s Manchester

  10. Novel Scheme for Packet Forwarding without Header Modifications in Optical Networks

    DEFF Research Database (Denmark)

    Wessing, Henrik; Christiansen, Henrik Lehrmann; Fjelde, Tina

    2002-01-01

    We present a novel scheme for packet forwarding in optical packet-switched networks and we further demonstrate its good scalability through simulations. The scheme requires neither header modification nor any label distribution protocol, thus reducing component cost while simplifying network...

  11. Simulation of neuro-fuzzy model for optimization of combine header setting

    Directory of Open Access Journals (Sweden)

    S Zareei

    2016-09-01

    Full Text Available Introduction The noticeable proportion of producing wheat losses occur during production and consumption steps and the loss due to harvesting with combine harvester is regarded as one of the main factors. A grain combines harvester consists of different sets of equipment and one of the most important parts is the header which comprises more than 50% of the entire harvesting losses. Some researchers have presented regression equation to estimate grain loss of combine harvester. The results of their study indicated that grain moisture content, reel index, cutter bar speed, service life of cutter bar, tine spacing, tine clearance over cutter bar, stem length were the major parameters affecting the losses. On the other hand, there are several researchswhich have used the variety of artificial intelligence methods in the different aspects of combine harvester. In neuro-fuzzy control systems, membership functions and if-then rules were defined through neural networks. Sugeno- type fuzzy inference model was applied to generate fuzzy rules from a given input-output data set due to its less time-consuming and mathematically tractable defuzzification operation for sample data-based fuzzy modeling. In this study, neuro-fuzzy model was applied to develop forecasting models which can predict the combine header loss for each set of the header parameter adjustments related to site-specific information and therefore can minimize the header loss. Materials and Methods The field experiment was conducted during the harvesting season of 2011 at the research station of the Faulty of Agriculture, Shiraz University, Shiraz, Iran. The wheat field (CV. Shiraz was harvested with a Claas Lexion-510 combine harvester. The factors which were selected as main factors influenced the header performance were three levels of reel index (RI (forward speed of combine harvester divided by peripheral speed of reel (1, 1.2, 1.5, three levels of cutting height (CH(25, 30, 35 cm, three

  12. Steam drum level control studies of a natural circulation multi loop reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Rajesh; Contractor, A.D.; Srivastava, Abhishek; Lele, H.G. [Bhabha Atomic Research Centre, Trombay, Mumbai (India). Reactor Safety Div.; Vaze, K.K. [Bhabha Atomic Research Centre, Trombay, Mumbai (India). Reactor Design and Development Group

    2013-12-15

    The proposed heavy water moderated and light water cooled pressure tube type boiling water reactor works on natural circulation at all power levels. It has parallel inter-connected loops with 452 boiling channels in the main heat transport system configuration. These multiple (four) interconnected loops influence the steam drum level control adversely through the common reactor inlet header. Alternate design studies made earlier for efficient control of SD levels have shown favorable results. This has lead to explore further the present scheme with the compartmentalization of CRIH into four compartments catering to four loops separately. The conventional 3-element level control has been found to be working satisfactorily. The interconnections between ECCS header and inlet header compartments have also increased the safety margin for various LOCA and design basis events. The paper deals with the SD level control aspects for this novel MHT configuration which has been analyzed for various PIEs (Postulated Initiating Events) and found to be satisfactory. (orig.)

  13. Right ventricular hydatid cyst ruptured to pericardium

    Directory of Open Access Journals (Sweden)

    Feridoun Sabzi

    2015-01-01

    Full Text Available Cardiac hydatidosis is rare presentation of body hydatidosis. Incidence of cardiac involvements range from 5% to 5% of patients with hydatid disease. Most common site of hydatid cyst in heart is interventricular septum and left ventricular free wall. Right ventricular free wall involvement by cyst that ruptured to pericardial cavity is very rare presentation of hydatid cyst. Cardiac involvement may have serious consequences such as rupture to blood steam or pericardial cavity. Both the disease and its surgical treatment carry a high complication rate, including rupture leading to cardiac tamponade, anaphylaxis and also death. In the present report, a 43-year-old man with constrictive pericarditis secondary to a pericardial hydatid cyst is described.

  14. 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.)

  15. Rupture of primigravid uterus and recurrent rupture

    Directory of Open Access Journals (Sweden)

    Nahreen Akhtar

    2016-08-01

    Full Text Available Uterine rupture is a deadly obstetrical emergency endangering the life of both mother and fetus. In Bangladesh, majority of deliveries arc attended by unskilled traditional birth attendant and maternal mortality is still quite high. It is rare Ln developed country but unfortunately it is common in a developing country like Bangladesh. We report a case history of a patient age 32yrs from Daudkandi, Comilla admitted with H/0 previous two rupture uterus and repair with no living issue. We did caesarean section at her 31+ weeks of pregnancy when she developed Jabour pain. A baby of 1.4 kg was delivered. During cesarean section, focal rupture was noted in previous scar of rupture. Unfortunately the baby expired in neonatal ICU after 36 hours.

  16. Instationary discharge rates and shear factors in pipe ruptures

    International Nuclear Information System (INIS)

    Pana, P.

    1976-01-01

    The loads observed in ruptures of steam- or water-conducting pipes may occur as reactive forces on the pipes themselves or as jet forces on the structural components adjacent to the point of rupture. The present paper deals with the instationary acceleration phase directly after rupture. The general laws of conservation (mass, energy, momentum) may be used, but in their instationary form. This results in a system of partial differential equations which does not provide a comprehensive mathematical solution. However, since efficient electronic computer systems are available, difference methods are increasingly often used. Such calculations were carried out for water-steam as an ideal gas and under simplifying assumptions. (orig./AK) [de

  17. Removal of portions of tubes from steam generator of nuclear reactor

    International Nuclear Information System (INIS)

    Wilkins, R.L.; Williams, C.F.

    1983-01-01

    After the tube portion to be removed is severed from the remainder of the U-tube and its weld to the header is machined off, the internal surface of the portion is engaged internally by an ID gripper and pulled out of the header. Then the external surface is engaged by an OD gripper and pulled further out of the header. The first tube length is pulled out as far as the space under the header permits and is then cut off. Successive lengths are likewise pulled out and cut off. The apparatus for accomplishing this object includes a base secured to the header by expanded mandrel mechanisms. A carriage is suspended from the base on screws which are driven by a motor to move the carriage away from and towards the base. An OD gripper assembly is suspended from the carriage and is movable by fluoroactuated piston rods away from and towards the carriage. An ID gripper assembly extends through the OD gripper assembly. The gripper of the ID assembly is actuable to engage the internal surface of the tube portion. With its gripper so engaged the ID assembly is engaged by the gripper of the OD assembly and the engaged tube portion is pulled out of the header by the OD assembly. The ID gripper is then disengaged and the OD gripper is engaged with the tube portion in the same way that it engages the ID assembly and the tube portion is pulled out further. The apparatus also includes a tube cutter having an abrasive wheel. The wheel cuts the lengths of the tube portion at an angle so that for examination and testing the tube lengths can be matched and the orientation of any defect with respect to the plate in the steam generator which separates the inlet and outlet ends of the tubes and the U-tube supports can be identified

  18. Simulation of the behaviour of a servo actuated check valve upon rupture of the feedwater pipe

    International Nuclear Information System (INIS)

    Lucas, A.M. de; Perezagua, R.L.; Rosa, B. de la; Sanz, J.

    1995-01-01

    The steam generator replacement programme at Almaraz NPP, provides for the installation of a replacement damped non-return valve for the feedwater system. the function of this valve is to protect the steam generator in the event of a rupture in the feedwater pipe. Sudden closure of the check valve, against the flow and following rupture of the feedwater pipe, causes overpressure in the valve which is transmitted to the steam generator nozzle. It is therefore necessary to know this when designing the internal systems of the steam generator. Using the RELAP5/MODE3 code, it has been possible to simulate the dynamic behaviour of a check valve upon rupture of a feedwater pipe postulated outside the containment. The calculation model has been applied to different types of check valve. (Author)

  19. Hepatic rupture in preeclampsia

    International Nuclear Information System (INIS)

    Winer-Muram, H.T.; Muram, D.; Salazar, J.; Massie, J.D.

    1985-01-01

    The diagnosis of hepatic rupture in patients with pregnancy-induced hypertension (preeclampsia and eclampsia) is rarely made preoperatively. Diagnostic imaging can be utilized in some patients to confirm the preoperative diagnosis. Since hematoma formation precedes hepatic rupture, then, when diagnostic modalities such as sonography and computed tomography identify patients with hematomas, these patients are at risk of rupture, and should be hospitalized until the hematomas resolve

  20. Failure analysis of retired steam generator tubings

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-04-15

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

  1. LWR primary coolant pipe rupture test rig

    International Nuclear Information System (INIS)

    Yoshitoshi, Shyoji

    1978-01-01

    The rupture test rig for primary coolant pipes is constructed in the Japan Atomic Energy Research Institute to verify the reliability of the primary coolant pipes for both PWRs and BWRs. The planned test items consisted of reaction force test, restraint test, whip test, jet test and continuous release test. A pressure vessel of about 4 m 3 volume, a circulating pump, a pressurizer, a heater, an air cooler and the related instrumentation and control system are included in this test rig. The coolant test condition is 160 kg/cm 2 g, 325 deg C for PWR test, and 70 kg/cm 2 g, saturated water and steam for BWR test, 100 ton of test load for the ruptured pipe bore of 8B Schedule 160, and 20 lit/min. discharge during 20 h for continuous release of coolant. The maximum pit internal pressure was estimated for various pipe diameters and time under the PWR and BWR conditions. The spark rupturing device was adopted for the rupture mechanics in this test rig. The computer PANAFACOM U-300 is used for the data processing. This test rig is expected to operate in 1978 effectively for the improvement of reliability of LWR primary coolant pipes. (Nakai, Y.)

  2. Steaming ahead

    International Nuclear Information System (INIS)

    Anon.

    1997-01-01

    An example of the development of geothermal power in Indonesia is described. Wells are being drilled into the Salak volcano on Java, about 60km south of Jakarta. These let out high pressure hot water trapped 1 to 3km below the surface which can be flashed into steam for driving turbines. The hot water field has already produced 110MW of power since 1994 and is currently being expanded to 330MW. Some details of the drilling and civil engineering are given. Since Indonesia sits on the edge of giant tectonic boundary known as the ''Pacific ring of fire'', the potential for further development is enormous. Ultimately volcanic activity could release an estimated 27,000MW capacity. More realistically, 2,000MW of crustal power by 2020 is spoken of. (UK)

  3. Asynchronous broadcast for ordered delivery between compute nodes in a parallel computing system where packet header space is limited

    Science.gov (United States)

    Kumar, Sameer

    2010-06-15

    Disclosed is a mechanism on receiving processors in a parallel computing system for providing order to data packets received from a broadcast call and to distinguish data packets received at nodes from several incoming asynchronous broadcast messages where header space is limited. In the present invention, processors at lower leafs of a tree do not need to obtain a broadcast message by directly accessing the data in a root processor's buffer. Instead, each subsequent intermediate node's rank id information is squeezed into the software header of packet headers. In turn, the entire broadcast message is not transferred from the root processor to each processor in a communicator but instead is replicated on several intermediate nodes which then replicated the message to nodes in lower leafs. Hence, the intermediate compute nodes become "virtual root compute nodes" for the purpose of replicating the broadcast message to lower levels of a tree.

  4. Continuous monitoring of variations in the 235U enrichment of uranium in the header pipework of a centrifuge enrichment plant

    International Nuclear Information System (INIS)

    Packer, T.W.

    1991-01-01

    Non-destructive assay equipment, based on gamma-ray spectrometry and x-ray fluorescence analysis has previously been developed for confirming the presence of low enriched uranium in the header pipework of UF 6 gas centrifuge enrichment plants. However inspections can only be carried out occasionally on a limited number of pipes. With the development of centrifuge enrichment technology it has been suggested that more frequent, or ideally, continuous measurements should be made in order to improve safeguards assurance between inspections. For this purpose we have developed non-destructive assay equipment based on continuous gamma-ray spectrometry and x-ray transmission measurements. This equipment is suitable for detecting significant changes in the 235 U enrichment of uranium in the header pipework of new centrifuge enrichment plants. Results are given in this paper of continuous measurements made in the laboratory and also on header pipework of a centrifuge enrichment plant at Capenhurst

  5. Daily Press Headers as a Reinforcement to Brand Identity in Spanish Sport Newspapers both Print and Online

    Directory of Open Access Journals (Sweden)

    Belén PUEBLA MARTÍNEZ

    2015-06-01

    Full Text Available Press headers are to daily newspapers the same as brands are to products. Because a newspaper is an object itself which also participates in a double designing process: from information and from product design; and in both cases this serves the same purpose, that the reader feels attracted and comes back for more every day. On that trip, which takes place either to the newsstand or to the computer, to become visible and unique is of paramount importance and the header is the element that best identifies not only the publication but also the tone of the language that the reader expects to find in it. This study intends to dive into the Spanish sport daily press headers, both print and digital, to establish how newspapers achieve their pretended brand identity.

  6. Splenic rupture masquerading ruptured ectopic pregnancy | Kigbu ...

    African Journals Online (AJOL)

    The classical triad of presentation of delayed menses, irregular vaginal bleeding and abdominal pain may not be encountered at all! Overwhelming features of abdominal pain, amenorrhea, pallor, abdominal tenderness, shifting dullness with positive pregnancy test gave a clinical diagnosis of ruptured ectopic pregnancy.

  7. Predicting the fidelity of JPEG2000 compressed CT images using DICOM header information

    International Nuclear Information System (INIS)

    Kim, Kil Joong; Kim, Bohyoung; Lee, Hyunna; Choi, Hosik; Jeon, Jong-June; Ahn, Jeong-Hwan; Lee, Kyoung Ho

    2011-01-01

    Purpose: To propose multiple logistic regression (MLR) and artificial neural network (ANN) models constructed using digital imaging and communications in medicine (DICOM) header information in predicting the fidelity of Joint Photographic Experts Group (JPEG) 2000 compressed abdomen computed tomography (CT) images. Methods: Our institutional review board approved this study and waived informed patient consent. Using a JPEG2000 algorithm, 360 abdomen CT images were compressed reversibly (n = 48, as negative control) or irreversibly (n = 312) to one of different compression ratios (CRs) ranging from 4:1 to 10:1. Five radiologists independently determined whether the original and compressed images were distinguishable or indistinguishable. The 312 irreversibly compressed images were divided randomly into training (n = 156) and testing (n = 156) sets. The MLR and ANN models were constructed regarding the DICOM header information as independent variables and the pooled radiologists' responses as dependent variable. As independent variables, we selected the CR (DICOM tag number: 0028, 2112), effective tube current-time product (0018, 9332), section thickness (0018, 0050), and field of view (0018, 0090) among the DICOM tags. Using the training set, an optimal subset of independent variables was determined by backward stepwise selection in a four-fold cross-validation scheme. The MLR and ANN models were constructed with the determined independent variables using the training set. The models were then evaluated on the testing set by using receiver-operating-characteristic (ROC) analysis regarding the radiologists' pooled responses as the reference standard and by measuring Spearman rank correlation between the model prediction and the number of radiologists who rated the two images as distinguishable. Results: The CR and section thickness were determined as the optimal independent variables. The areas under the ROC curve for the MLR and ANN predictions were 0.91 (95% CI; 0

  8. 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.

  9. Power raise through improved reactor inlet header temperature measurement at Bruce A Nuclear Generation Station

    International Nuclear Information System (INIS)

    Basu, S.; Bruggemn, D.

    1997-01-01

    Reactor Inlet Header (RIH) temperature has become a factor limiting the performance of the Ontario Hydro Bruce A units. Specifically, the RIH temperature is one of several parameters that is preventing the Bruce A units from returning to 94% power operation. RIH temperature is one of several parameters which affect the critical heat flux in the reactor channel, and hence the integrity of the fuel. Ideally, RIH temperature should be lowered, but this cannot be done without improving the heat transfer performance of the boilers and feedwater pre-heaters. Unfortunately, the physical performance of the boilers and pre-heaters has decayed and continues to decay over time and as a result the RIH temperature has been rising and approaching its defined limit. With an understanding of the current RIH temperature measurement loop and methods available to improve it, a solution to reduce the measurement uncertainty is presented

  10. Air Emission Projections During Acid Cleaning of F-Canyon Waste Header No.2

    International Nuclear Information System (INIS)

    CHOI, ALEXANDER

    2004-01-01

    The purpose of this study was to develop the air emission projections for the maintenance operation to dissolve and flush out the scale material inside the F-Canyon Waste Header No.2. The chemical agent used for the dissolution is a concentrated nitric acid solution, so the pollutant of concern is the nitric acid vapor. Under the very conservative operating scenarios considered in this study, it was determined that the highest possible rate of nitric acid emission during the acid flush would be 0.048 lb. per hr. It turns out that this worst-case air emission projection is just below the current exemption limit of 0.05 lb. per hr. for permit applications

  11. Steam process supply optimization for Arcelormittal Tubarao consumers; Otimizacao do sistema de fornecimento de vapor de processo para a usina (AMT)

    Energy Technology Data Exchange (ETDEWEB)

    Loss, Gecimar; Oliveira, Heron Domingues de; Silva, Jose Geraldo Lessa; Beccalli, Marcelo; Calente, Paulo Sergio Boni; Monteiro, Sergio Anderson [Companhia Siderurgica de Tubarao ArcelorMittal, Serra, ES (Brazil)

    2010-07-01

    The ArcelorMittal Tubarao Energy Production area is compounded by three units: Air Separation Units, Thermal Power Plants and Thermal Recovery Power Plants. The Thermo Power Plants are co-generated units responsible to generate electrical, mechanical (Blast Furnace blower) energy and also provide Steam to complement the facility internal consumption mainly provided by CDQ plant (CDQ - Coke Dry Quenching). Since RH2 (steel treatment process) start up, the steam consumption increased and the Thermal Power Plant contribution raised to attend this new demand. Solutions were needed to guarantee the steam supply by the Power Plant even in low steam header stoppages for maintenance, since the lack of steam caused by shortage in Power Plant steam supply resulting in steel production diminution in this new scenario. (author)

  12. Large-leak sodium-water reaction analysis for steam generators

    Energy Technology Data Exchange (ETDEWEB)

    Sakano, K; Shindo, Y; Hori, M

    1975-07-01

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

  13. Large-leak sodium-water reaction analysis for steam generators

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  14. 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.)

  15. Premature rupture of membranes

    Science.gov (United States)

    ... gov/ency/patientinstructions/000512.htm Premature rupture of membranes To use the sharing features on this page, ... water that surrounds your baby in the womb. Membranes or layers of tissue hold in this fluid. ...

  16. Ruptured submitral aneurysm

    Directory of Open Access Journals (Sweden)

    V. Shukla

    2016-09-01

    Full Text Available Submitral aneurysm is a rare entity, with around few hundred cases reported till date. Presentation can be varied. We describe here a case of submitral aneurysm in a young male with rupture into the left atrium cavity.

  17. Achilles Tendon Rupture

    Science.gov (United States)

    ... is a strong fibrous cord that connects the muscles in the back of your calf to your heel bone. If you overstretch your Achilles tendon, it can tear (rupture) completely or just partially. If your Achilles ...

  18. Water leak detection in steam generator of SUPER PHENIX

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  19. Fabrication of an improved tube-to-pipe header heat exchanger for the Fuel Failure Mockup (FFM) Facility

    International Nuclear Information System (INIS)

    Prislinger, J.J.; Jones, R.H.

    1977-05-01

    The procedure used in fabricating an improved tube-to-pipe header heat exchanger for the Fuel Failure Mockup (FFM) Facility is described. Superior performance is accomplished at reduced cost with adherence to the ASME Boiler and Pressure Vessel Code. The techniques used and the method of fabrication are described in detail

  20. Novel scheme for efficient and cost-effective forwarding of packets in optical networks without header modification

    DEFF Research Database (Denmark)

    Wessing, Henrik; Fjelde, Tina; Christiansen, Henrik Lehrmann

    2001-01-01

    We present a novel scheme for addressing the outputs in optical packet switches and demonstrate its good scalability. The scheme requires neither header modification nor distribution of routing information to the packet switches, thus reducing optical component count while simplifying network...

  1. Ruptured cornual pregnancy

    International Nuclear Information System (INIS)

    Hussain, M.; Yasmeen, H.; Noorani, K.

    2003-01-01

    A case of ruptured cornual pregnancy is presented here. The patient presented with history of 30 weeks gestational amenorrhoea and pain in the lower abdomen and epigastrium for the last seven days. Ultrasound revealed a 29 weeks abdominal pregnancy with blood in the pelvic cavity. On laparotomy; there was a ruptured right cornual pregnancy, treated cornual resection and uterine repair. An alive male baby of one kg weight was delivered from the resected cornua of the uterus. (author)

  2. The effect of the flow direction inside the header on two-phase flow distribution in parallel vertical channels

    International Nuclear Information System (INIS)

    Marchitto, A.; Fossa, M.; Guglielmini, G.

    2012-01-01

    Uniform fluid distribution is essential for efficient operation of chemical-processing equipment such as contactors, reactors, mixers, burners and in most refrigeration equipment, where two phases are acting together. To obtain optimum distribution, proper consideration must be given to flow behaviour in the distributor, flow conditions upstream and downstream of the distributor, and the distribution requirements (fluid or phase) of the equipment. Even though the principles of single phase distribution have been well developed for more than three decades, they are frequently not taken in the right account by equipment designers when a mixture is present, and a significant fraction of process equipment consequently suffers from maldistribution. The experimental investigation presented in this paper is aimed at understanding the main mechanisms which drive the flow distribution inside a two-phase horizontal header in order to design improved distributors and to optimise the flow distribution inside compact heat exchanger. Experimentation was devoted to establish the influence of the inlet conditions and of the channel/distributor geometry on the phase/mass distribution into parallel vertical channels. The study is carried out with air–water mixtures and it is based on the measurement of component flow rates in individual channels and on pressure drops across the distributor. The effects of the operating conditions, the header geometry and the inlet port nozzle were investigated in the ranges of liquid and gas superficial velocities of 0.2–1.2 and 1.5–16.5 m/s, respectively. In order to control the main flow direction inside the header, different fitting devices were tested; the insertion of a co-axial, multi-hole distributor inside the header has confirmed the possibility of greatly improving the liquid and gas flow distribution by the proper selection of position, diameter and number of the flow openings between the supplying distributor and the system of

  3. 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

  4. CFD modeling of a boiler's tubes rupture

    International Nuclear Information System (INIS)

    Rahimi, Masoud; Khoshhal, Abbas; Shariati, Seyed Mehdi

    2006-01-01

    This paper reports the results of a study on the reason for tubes damage in the superheater Platen section of the 320 MW Bisotoun power plant, Iran. The boiler has three types of superheater tubes and the damage occurs in a series of elbows belongs to the long tubes. A three-dimensional modeling was performed using an in-house computational fluid dynamics (CFD) code in order to explore the reason. The code has ability of simultaneous solving of the continuity, the Reynolds-Averaged Navier-Stokes (RANS) equations and employing the turbulence, combustion and radiation models. The whole boiler including; walls, burners, air channels, three types of tubes, etc., was modeled in the real scale. The boiler was meshed into almost 2,000,000 tetrahedral control volumes and the standard k-ε turbulence model and the Rosseland radiation model were used in the model. The theoretical results showed that the inlet 18.9 MPa saturated steam becomes superheated inside the tubes and exit at a pressure of 17.8 MPa. The predicted results showed that the temperature of the steam and tube's wall in the long tubes is higher than the short and medium size tubes. In addition, the predicted steam mass flow rate in the long tube was lower than other ones. Therefore, it was concluded that the main reason for the rupture in the long tubes elbow is changing of the tube's metal microstructure due to working in a temperature higher than the design temperature. In addition, the structural fatigue tension makes the last elbow of the long tube more ready for rupture in comparison with the other places. The concluded result was validated by observations from the photomicrograph of the tube's metal samples taken from the damaged and undamaged sections

  5. Analysis of the Steam Generator Tubes Rupture Initiating Event

    International Nuclear Information System (INIS)

    Trillo, A.; Minguez, E.; Munoz, R.; Melendez, E.; Sanchez-Perea, M.; Izquierd, J.M.

    1998-01-01

    In PSA studies, Event Tree-Fault Tree techniques are used to analyse to consequences associated with the evolution of an initiating event. The Event Tree is built in the sequence identification stage, following the expected behaviour of the plant in a qualitative way. Computer simulation of the sequences is performed mainly to determine the allowed time for operator actions, and do not play a central role in ET validation. The simulation of the sequence evolution can instead be performed by using standard tools, helping the analyst obtain a more realistic ET. Long existing methods and tools can be used to automatism the construction of the event tree associated to a given initiator. These methods automatically construct the ET by simulating the plant behaviour following the initiator, allowing some of the systems to fail during the sequence evolution. Then, the sequences with and without the failure are followed. The outcome of all this is a Dynamic Event Tree. The work described here is the application of one such method to the particular case of the SGTR initiating event. The DYLAM scheduler, designed at the Ispra (Italy) JRC of the European Communities, is used to automatically drive the simulation of all the sequences constituting the Event Tree. Similarly to the static Event Tree, each time a system is demanded, two branches are open: one corresponding to the success and the other to the failure of the system. Both branches are followed by the plant simulator until a new system is demanded, and the process repeats. The plant simulation modelling allows the treatment of degraded sequences that enter into the severe accident domain as well as of success sequences in which long-term cooling is started. (Author)

  6. 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

  7. 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.

  8. 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

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

    International Nuclear Information System (INIS)

    Bassett, B.

    1983-01-01

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

  10. 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

  11. 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…

  12. 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

  13. 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.

  14. 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.

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

    International Nuclear Information System (INIS)

    Mazufri, C.M

    2000-01-01

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

  16. Evaluation of intergranular cracks on the ring header cross at Grand Gulf Unit No. 1

    International Nuclear Information System (INIS)

    Czajkowski, C.J.

    1987-01-01

    A metallurgical investigation was performed on a sample of cracked ring header cross material from the Grand Gulf Unit No. 1 Nuclear Power Station. The cracks were located in a 6-7 in (15-17.5 cm) width band running circumferentially below the cross to cap weld with a similar band above the cross to discharger pipe weld. The indications were up to 19 mm in length and 6.0 mm in depth. This particular sample was cut from a cross which had not seen actual service but which had been used to qualify the induction heating stress improvement (IHSI) technique for the Grand Gulf units. The base material was SA 182 material manufactured to SA 403-type WP 304 stainless steel. The investigation consisted of visual/dye penetrant examination, chemical analysis, hardness testing, optical microscopy, scanning electron microscopy and energy dispersive spectroscopy. The evaluated cracks were intergranular and initiated on the forging's exterior surface. The grain size of the material was larger than ASTM 00 and no definitive corrosive species were found by Energy Dispersive Spectroscopy (EDS). The cracking is considered to be the result of the forging having been overheated/burned during manufacture. (author)

  17. Dynamic simulation of steam generator failures

    Energy Technology Data Exchange (ETDEWEB)

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

    1988-07-01

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

  18. Dynamic simulation of steam generator failures

    International Nuclear Information System (INIS)

    Meister, G.

    1988-01-01

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

  19. Rupture, waves and earthquakes.

    Science.gov (United States)

    Uenishi, Koji

    2017-01-01

    Normally, an earthquake is considered as a phenomenon of wave energy radiation by rupture (fracture) of solid Earth. However, the physics of dynamic process around seismic sources, which may play a crucial role in the occurrence of earthquakes and generation of strong waves, has not been fully understood yet. Instead, much of former investigation in seismology evaluated earthquake characteristics in terms of kinematics that does not directly treat such dynamic aspects and usually excludes the influence of high-frequency wave components over 1 Hz. There are countless valuable research outcomes obtained through this kinematics-based approach, but "extraordinary" phenomena that are difficult to be explained by this conventional description have been found, for instance, on the occasion of the 1995 Hyogo-ken Nanbu, Japan, earthquake, and more detailed study on rupture and wave dynamics, namely, possible mechanical characteristics of (1) rupture development around seismic sources, (2) earthquake-induced structural failures and (3) wave interaction that connects rupture (1) and failures (2), would be indispensable.

  20. How safe is defect specific maintenance of steam generator tubes?

    International Nuclear Information System (INIS)

    Dvorsek, T.; Cizelj, L.

    1995-01-01

    Outside diameter stress corrosion cracking at the tube to tube support plate intersections is assessed in the paper. The impact of defect specific maintenance on steam generator operation safety and reliability was investigated. This was performed by comparing efficiencies of defect specific and traditional maintenance strategy. The efficiency was studied through expected primary-to-secondary leak rate and tube rupture probability in a case of postulated accidental operating conditions, and number of tubes which shall be plugged using both maintenance strategies. In general, the efficiency of specific maintenance is function of particular steam generator and operating cycle. (author)

  1. 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.)

  2. 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

  3. Studies of the steam generator degraded tubes behavior on BRUTUS test loop

    Energy Technology Data Exchange (ETDEWEB)

    Chedeau, C.; Rassineux, B. [EDF/DER/MTC, Moret Sur Loing (France); Flesch, B. [EDF/EPN/DMAINT, Paris (France)] [and others

    1997-04-01

    Studies for the evaluation of steam generator tube bundle cracks in PWR power plants are described. Global tests of crack leak rates and numerical calculations of crack opening area are discussed in some detail. A brief overview of thermohydraulic studies and the development of a mechanical probabilistic design code is also given. The COMPROMIS computer code was used in the studies to quantify the influence of in-service inspections and maintenance work on the risk of a steam generator tube rupture.

  4. 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…

  5. 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…

  6. 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…

  7. 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

  8. 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.)

  9. 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.

  10. 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

  11. Damage distribution and remnant life assessment of a super-heater outlet header used for long time

    Energy Technology Data Exchange (ETDEWEB)

    Hiroyuki, Okamura [Science Univ. of Tokyo (Japan); Ryuichi, Ohotani [Kyoto Univ. (Japan); Kazuya, Fujii [Japan Power Engineering and Inspection Corp., Tokyo (Japan); Masashi, Nakashiro; Fumio, Takemasa; Hideo, Umaki; Tomiyasu, Masumura [Ishikawajima-Harima Heavy Industries Co. Ltd., Tokyo (Japan)

    1998-11-01

    This paper presents the results of investigation on evaluating damage distribution to base metals and welded joints in the thickness direction and evaluate damage on ligaments. Thick wall tested sample was the superheater outlet header component long term serviced in high pressure and temperature condition in thermal power plant. The simulate unused steel of component material was made from sample by suitable heat treatment, and the extent of damage was assessed based on a comparison of nondestructive and destructive test results between simulate unused and aged samples. Damage evaluation was also made by FEM structural stress analysis. (orig./MM)

  12. [Achilles tendon rupture].

    Science.gov (United States)

    Thermann, H; Hüfner, T; Tscherne, H

    2000-03-01

    The treatment of acute of Achilles tendon rupture experienced a dynamic development in the last ten years. Decisive for this development was the application of MRI and above all the ultrasonography in the diagnostics of the pathological changes and injuries of tendons. The question of rupture morphology as well as different courses of healing could be now evaluated objectively. These advances led consequently to new modalities in treatment concepts and rehabilitation protocols. The decisive input for improvements of the outcome results and particularly the shortening of the rehabilitation period came with introduction of the early functional treatment in contrast to immobilizing plaster treatment. In a prospective randomized study (1987-1989) at the Trauma Dept. of the Hannover Medical School could show no statistical differences comparing functional non-operative with functional operative therapy with a special therapy boot (Variostabil/Adidas). The crucial criteria for therapy selection results from the sonographically measured position of the tendon stumps in plantar flexion (20 degrees). With complete adaptation of the tendons' ends surgical treatment does not achieve better results than non-operative functional treatment in term of tendon healing and functional outcome. Regarding the current therapeutic standards each method has is advantages and disadvantages. Both, the operative and non-operative functional treatment enable a stable tendon healing with a low risk of re-rupture (1-2%). Meanwhile there is consensus for early functional after-treatment of the operated Achilles' tendons. There seems to be a trend towards non-operative functional treatment in cases of adequate sonographical findings, or to minimal invasive surgical techniques.

  13. A computational fluid dynamics and effectiveness-NTU based co-simulation approach for flow mal-distribution analysis in microchannel heat exchanger headers

    International Nuclear Information System (INIS)

    Huang, Long; Lee, Moon Soo; Saleh, Khaled; Aute, Vikrant; Radermacher, Reinhard

    2014-01-01

    Refrigerant flow mal-distribution is a practical challenge in most microchannel heat exchangers (MCHXs) applications. Geometry design, uneven heat transfer and pressure drop in the different microchannel tubes are three main reasons leading to the flow mal-distribution. To efficiently and accurately account for these three effects, a new MCHX co-simulation approach is proposed in this paper. The proposed approach combines a detailed header simulation based on computational fluid dynamics (CFD) and a robust effectiveness-based finite volume tube-side heat transfer and refrigerant flow modeling tool. The co-simulation concept is demonstrated on a ten-tube MCHX case study. Gravity effect and uneven airflow effect were numerically analyzed using both water and condensing R134a as the working fluids. The approach was validated against experimental data for an automotive R134a condenser. The inlet header was cut open after the experimental data had been collected. The detailed header geometry was reproduced using the proposed CFD header model. Good prediction accuracy was achieved compared to the experimental data. The presented co-simulation approach is capable of predicting detailed refrigerant flow behavior while accurately predicts the overall heat exchanger performance. - Highlights: •MCHX header flow distribution is analyzed by a co-simulation approach. •The proposed method is capable of simulating both single-phase and two-phase flow. •An actual header geometry is reproduced in the CFD header model. •The modeling work is experimentally validated with good accuracy. •Gravity effect and air side mal-distribution are accounted for

  14. Slow rupture of frictional interfaces

    OpenAIRE

    Sinai, Yohai Bar; Brener, Efim A.; Bouchbinder, Eran

    2011-01-01

    The failure of frictional interfaces and the spatiotemporal structures that accompany it are central to a wide range of geophysical, physical and engineering systems. Recent geophysical and laboratory observations indicated that interfacial failure can be mediated by slow slip rupture phenomena which are distinct from ordinary, earthquake-like, fast rupture. These discoveries have influenced the way we think about frictional motion, yet the nature and properties of slow rupture are not comple...

  15. Slow rupture of frictional interfaces

    Science.gov (United States)

    Bar Sinai, Yohai; Brener, Efim A.; Bouchbinder, Eran

    2012-02-01

    The failure of frictional interfaces and the spatiotemporal structures that accompany it are central to a wide range of geophysical, physical and engineering systems. Recent geophysical and laboratory observations indicated that interfacial failure can be mediated by slow slip rupture phenomena which are distinct from ordinary, earthquake-like, fast rupture. These discoveries have influenced the way we think about frictional motion, yet the nature and properties of slow rupture are not completely understood. We show that slow rupture is an intrinsic and robust property of simple non-monotonic rate-and-state friction laws. It is associated with a new velocity scale cmin, determined by the friction law, below which steady state rupture cannot propagate. We further show that rupture can occur in a continuum of states, spanning a wide range of velocities from cmin to elastic wave-speeds, and predict different properties for slow rupture and ordinary fast rupture. Our results are qualitatively consistent with recent high-resolution laboratory experiments and may provide a theoretical framework for understanding slow rupture phenomena along frictional interfaces.

  16. Stress analysis of HTR-10 steam generator heat exchanging tubes

    International Nuclear Information System (INIS)

    Dong Jianling; Zhang Xiaohang; Yin Dejian; Fu Jiyang

    2001-01-01

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

  17. LHCb: Dynamically Adaptive Header Generator and Front-End Source Emulator for a 100 Gbps FPGA Based DAQ

    CERN Multimedia

    Srikanth, S

    2014-01-01

    The proposed upgrade for the LHCb experiment envisages a system of 500 Data sources each generating data at 100 Gbps, the acquisition and processing of which is a big challenge even for the current state of the art FPGAs. This requires an FPGA DAQ module that not only handles the data generated by the experiment but also is versatile enough to dynamically adapt to potential inadequacies of other components like the network and PCs. Such a module needs to maintain real time operation while at the same time maintaining system stability and overall data integrity. This also creates a need for a Front-end source Emulator capable of generating the various data patterns, that acts as a testbed to validate the functionality and performance of the Header Generator. The rest of the abstract briefly describes these modules and their implementation. The Header Generator is used to packetize the streaming data from the detectors before it is sent to the PCs for further processing. This is achieved by continuously scannin...

  18. Experimental use of road header (AM-50) as face cutting machine for extraction of coal in longwall panel

    Energy Technology Data Exchange (ETDEWEB)

    Passi, K.K.; Kumar, C.R.; Prasad, P. [DGMS, Dhanbad (India)

    2001-07-01

    The scope of this paper has been limited to the use of available machines and techniques for attaining higher and more efficient production in underground coal mines. Under certain conditions of strata and higher degree of gassiness, the longwall method with hydraulic sand stowing is the only appropriate method of work for extraction of thick seam. In Moonidih Jitpur Colliery of M/S IISCO, No. 14 seam, Degree III gassy seam, 9.07 m thick, is extracted in multilift system with hydraulic sand stowing. In general, the bottom lift is extracted by Single Ended Ranging Arm Shearer and the middle and top lift are extracted by conventional method. However, in one of the panels spare road header machine was used as face cutting machine in bottom lift, on an experimental basis. This paper presents a successful case study of extraction of bottom lift coal by the longwall method with hydraulic sand stowing using road header (AM 50) as the face cutting machines. 9 figs.

  19. Detailed evaluation of RCS boundary rupture during high-pressure severe accident sequences

    International Nuclear Information System (INIS)

    Park, Rae-Joon; Hong, Seong-Wan

    2011-01-01

    A depressurization possibility of the reactor coolant system (RCS) before a reactor vessel rupture during a high-pressure severe accident sequence has been evaluated for the consideration of direct containment heating (DCH) and containment bypass. A total loss of feed water (TLOFW) and a station blackout (SBO) of the advanced power reactor 1400 (APR 1400) has been evaluated from an initiating event to a creep rupture of the RCS boundary by using the SCDAP/RELAP5 computer code. In addition, intentional depressurization of the RCS using power-operated safety relief valves (POSRVs) has been evaluated. The SCDAPRELAP5 results have shown that the pressurizer surge line broke before the reactor vessel rupture failure, but a containment bypass did not occur because steam generator U tubes did not break. The intentional depressurization of the RCS using POSRV was effective for the DCH prevention at a reactor vessel rupture. (author)

  20. French steam generator

    International Nuclear Information System (INIS)

    Remond, A.

    1986-01-01

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

  1. 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

  2. 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.)

  3. 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.)

  4. 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

  5. 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

  6. Untreated silicone breast implant rupture

    DEFF Research Database (Denmark)

    Hölmich, Lisbet R; Vejborg, Ilse M; Conrad, Carsten

    2004-01-01

    Implant rupture is a well-known complication of breast implant surgery that can pass unnoticed by both patient and physician. To date, no prospective study has addressed the possible health implications of silicone breast implant rupture. The aim of the present study was to evaluate whether untre...

  7. Blunt traumatic diaphragmatic rupture

    Directory of Open Access Journals (Sweden)

    Antonio Carlos Nogueira

    2011-09-01

    Full Text Available Traumatic injury of the diaphragm ranges from 0.6 to 1.2% and rise up to 5%among patients who were victims of blunt trauma and underwent laparotomy.Clinical suspicion associated with radiological assessment contributes to earlydiagnosis. Isolated diaphragmatic injury has a good prognosis. Generallyworse outcomes are associated with other trauma injuries. Bilateral andright diaphragmatic lesions have worse prognosis. Multi detector computed tomography (MDCT scan of the chest and abdomen provides better diagnosticaccuracy using the possibility of image multiplanar reconstruction. Surgicalrepair via laparotomy and/ or thoracotomy in the acute phase of the injury hasa better outcome and avoids chronic complications of diaphragmatic hernia.The authors present the case of a young male patient, victim of blunt abdominaltrauma due to motor vehicle accident with rupture of the diaphragm, spleenand kidney injuries. The diagnosis was made by computed tomography of thethorax and abdomen and was confirmed during laparotomy.

  8. Effect of Environment on Stress-Rupture Behavior of a Carbon Fiber-Reinforced Silicon Carbide (C/SiC) Ceramic Matrix Composite

    Science.gov (United States)

    Verrilli, Michael J.; Opila, Elizabeth J.; Calomino, Anthony; Kiser, J. Douglas

    2002-01-01

    Stress-rupture tests were conducted in air, vacuum, and steam-containing environments to identify the failure modes and degradation mechanisms of a carbon fiber-reinforced silicon carbide (C/SiC) composite at two temperatures, 600 and 1200 C. Stress-rupture lives in air and steam containing environments (50 - 80% steam with argon) are similar for a composite stress of 69 MPa at 1200 C. Lives of specimens tested in a 20% steam/argon environment were about twice as long. For tests conducted at 600 C, composite life in 20% steam/argon was 20 times longer than life in air. Thermogravimetric analysis of the carbon fibers was conducted under similar conditions to the stress-rupture tests. The oxidation rate of the fibers in the various environments correlated with the composite stress-rupture lives. Examination of the failed specimens indicated that oxidation of the carbon fibers was the primary damage mode for specimens tested in air and steam environments at both temperatures.

  9. Investigation of the high temperature steam oxidation of Zircaloy 4 cladding tubes

    International Nuclear Information System (INIS)

    Leistikow, S.; Berg, H. v.; Kraft, R.; Pott, E.; Schanz, G.

    1979-01-01

    Also for the ORNL Zircaloy 4 cladding material, an intermediate decrease of the proportion of the ZrO 2 /α-phase layer was found, followed by an drastic increase when the breakaway of the ZrO 2 -scale occurred. Other reasons for small divergencies were evaluated, for instance temperature and time measurements, metallographic evaluation of layer thicknesses, consequences of one-sided (ORNL) and double-sided (KfK) oxidation. The so-called anomalous effect of steam oxidation during temperature transients was reproduced qualitatively and-in case that a reduced gain of oxygen was observed-explained by the predominant existence of the monoclinic oxide phase. The creep-rupture tests below 800 0 C showed a moderate prolongation of time-to-rupture when the tests were performed in steam (or after preoxidation in steam) instead of argon. Also slightly reduced maximum circumferential strain could be measured. (orig./RW) [de

  10. 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)

  11. Flaw analysis in steam generator tube

    International Nuclear Information System (INIS)

    Hutin, J.P.; Billon, F.

    1985-08-01

    Operating more than 30 PWR units, Electricite de France has to face several steam generator tube problems. One of the most serious difficulties is the stress corrosion cracking due to primary fluid, just above the tube sheet, in the roll transition region. With regard to availability it is, of course, a major concern; with regard to safety, the point is that tube rupture should be preceded by a significant primary-to-secondary leak during normal operation so that the reactor should be shut down before failure occurs. The demonstration of this assessment asks for experimental and analytical evidences. In 1981, Elecricite de France started a comprehensive program on that subject. A general description of this program and the main results are to be presented during the SMIRT-8 Conference. The purpose of the present paper is to develop in greater detail the analytical part of the work

  12. Steam generator sludge removal apparatus

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  13. Mechanical properties of Ni-base superalloys in high temperature steam environments

    International Nuclear Information System (INIS)

    Jang, Changheui; Kim, Donghoon; Sah, Injin; Lee, Ho Jung

    2015-01-01

    The effects of environmental damages on the mechanical properties of Ni-base superalloys, Alloy 617 and Haynes 230, were evaluated for VHTR-HTSE applications. Tensile tests were carried out at room temperature after ageing at 900 deg. C in vacuum, steam, and steam + 20 vol.% H2 environments up to 3 000 h. Also, creep rupture test were performed in air, steam, and steam + 20 vol.% H2 environments. The degradations such as oxidation, decarburization, and redistribution of carbides were studied in view of the interaction of materials with the environment. During the long-term ageing at 900 deg. C in vacuum, secondary phases such as M23C6 and M6C were precipitated and coarsened, which caused increase in tensile strength and decrease in ductility. For the specimens aged in steam environments, surface and internal oxides acted as preferential sites for crack initiation and consequently, decreased the tensile and creep strength. Also, the formation of decarburization region resulted in glide plane failure during tensile test and reduction in creep rupture life due to grain boundary migration and recrystallisation. During creep tests, tensile stress caused the crack and void formation in oxide layer. Consequently, fast diffusion of oxidant occurred and environmental damage were accelerated. Among the test conditions, such environmental damage was much severe in steam environments. (authors)

  14. Simultaneous bilateral patellar tendon rupture ?

    OpenAIRE

    Moura, Diogo Lino; Marques, Jos? Pedro; Lucas, Francisco Manuel; Fonseca, Fernando Pereira

    2016-01-01

    Bilateral patellar tendon rupture is a rare entity, often associated with systemic diseases and patellar tendinopathy. The authors report a rare case of a 34-year-old man with simultaneous bilateral rupture of the patellar tendon caused by minor trauma. The patient is a retired basketball player with no past complaints of chronic knee pain and a history of steroid use. Surgical management consisted in primary end-to-end tendon repair protected temporarily with cerclage wiring, followed by a s...

  15. Water leak detection in steam generator of Super Phenix

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  16. The diagnosis of breast implant rupture

    DEFF Research Database (Denmark)

    Hölmich, Lisbet R; Vejborg, Ilse; Conrad, Carsten

    2005-01-01

    participated in either one or two study MRI examinations, aiming at determining the prevalence and incidence of silent implant rupture, respectively, and who subsequently underwent explantation. Implant rupture status was determined by four independent readers and a consensus diagnosis of either rupture...... were in fact ruptured at surgery. Thirty-four of the 43 intact implants were described as intact at surgery. When categorising possible ruptures as ruptures, there were one false positive and nine false negative rupture diagnoses at MRI yielding an accuracy of 92%, a sensitivity of 89...

  17. 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)

  18. Rupture detection device for pipeline in reactor

    International Nuclear Information System (INIS)

    Murakoshi, Toshinori; Kanamori, Shigeru; Shirasawa, Hirofumi.

    1991-01-01

    A difference between each of the pressures in a plurality of pipelines disposed in a shroud a reactor container and a pressure outside of the shroud is detected, thereby enabling safety and reliable detection even for simultaneous rapture and leakage of the pipelines. That is, a difference between the pressure of a steam phase outside of the shroud and a pressure in each of a plurality of low pressure injection pipelines in an emergency core cooling system opened to the inside of the shroud in the reactor container is detected by a difference pressure detector for each of them. Then, an average value for each of the pressure difference is determined, which is compared with the difference pressure obtained from each of the detectors in a comparator. Then, if openings should be caused by rupture, leakage or the like in any of the pipelines, the pressure in that pipeline is lowered to a vicinity of an atmospheric pressure and at the vapor phase pressure at the lowest. If the pressure is compared with the average value by the comparator, a negative difference is caused. Accordingly, an alarming unit generates an alarm based on the pressure difference signal, thereby enabling to specify the failed pipeline and provide an announce of the failure. (I.S.)

  19. Steam-water jet analysis. Final report

    International Nuclear Information System (INIS)

    Kashiwa, B.A.; Harlow, F.H.; Demuth, R.B.; Ruppel, H.M.

    1984-05-01

    This report presents the results of a theoretical study on the effects of the steam-water jet emitted from a hypothetical rupture in the high-pressure piping pf a nuclear power plant. A set of calculations is presented, incorporating increasingly complex formulations for mass and momentum exchange between the liquid and vapor flow fields. Comparisons between theory and detailed experimental data are given. The study begins with a thorough evaluation of the specification of equilibrium mass and momentum exchange (homogeneous equilibrium) throughout the flow region, a model that generally overpredicts the rate of jet momentum divergence. The study finds that a near-equilibrium momentum exchange rate and a strongly nonequilibrium momentum exchange rate are needed in the region of large vapor-volume fraction to explain the impingement data for fully developed two-phase jets. This leads to the viewpoint that the large-scale jet is characterized by a flow of large liquid entities that travel relatively unaffected by the strongly diverging vapor flow field. The study also finds circumstances in which a persistent core of metastable superheated water can cause much larger impingement pressures than would otherwise be possible. Existing engineering methods are evaluated for jet-loading predictions in plant design. The existing methods appear to be conservative in most possible rupture circumstances with one exception: when the impingement target is about one pipe-diameter away, large enough to capture the full jet, and the rupture flow area is equal to the full pipe flow area, the existing method can produce loadings that are slightly lower than observed for subcooled, flashing discharge. Recommendations have been made to improve the prediction of existing methods under these conditions

  20. Ruptured Ectopic Pregnancy

    Directory of Open Access Journals (Sweden)

    Valentina Park

    2016-09-01

    Full Text Available History of present illness: A 21-year-old female presented with sudden onset suprapubic abdominal pain associated with dysuria. The patient also experienced near syncope during bowel movements three times three days ago without falling or losing consciousness. She denied fever, nausea, and vomiting. She stated that she was five weeks pregnant by last menstrual period. She had an ultrasound a few weeks before that showed no intrauterine pregnancy, but she had not followed up for additional testing. Significant findings: The patient’s serum beta-hCG was 5,637 mIU/mL. The transvaginal ultrasound showed an empty uterus with free fluid posteriorly in the pelvis and Pouch of Douglas (00:00. A 4.5 cm heterogeneous mass was visible in the left adnexa concerning for an ectopic pregnancy (00:10. Discussion: Ectopic pregnancies are a leading cause of maternal morbidity and mortality, as well as decreased fertility.1,2 Differentiating between an ectopic pregnancy and a normal early pregnancy may be difficult, since ultrasound and quantitative beta-hCG may show inconclusive results.3,4 Patients who have used fertility treatment may further complicate the picture because they are at risk for heterotypic pregnancies.5 Ectopic pregnancies most commonly implant in the fallopian tube, but may alternatively implant in the ovary, cervix, abdomen, or uterine cornua.4 Ultrasonography may show an empty uterus, adnexal mass, pelvic free fluid, or an extra-uterine gestational sac, yolk sac, and/or embryo.6 Treatment options for ectopic pregnancy include surgery or methotrexate.2,4 Some patients may be candidates for close outpatient surveillance if the diagnosis is unclear or in very limited cases for early, non-ruptured ectopic pregnancies.2,4

  1. Steam reforming of ethanol

    DEFF Research Database (Denmark)

    Trane-Restrup, Rasmus; Dahl, Søren; Jensen, Anker Degn

    2013-01-01

    Steam reforming (SR) of oxygenated species like bio-oil or ethanol can be used to produce hydrogen or synthesis gas from renewable resources. However, deactivation due to carbon deposition is a major challenge for these processes. In this study, different strategies to minimize carbon deposition...

  2. Consolidated nuclear steam generator

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  3. Steam generators and furnaces

    Energy Technology Data Exchange (ETDEWEB)

    Swoboda, E

    1978-04-01

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

  4. Watt steam governor stability

    Science.gov (United States)

    Denny, Mark

    2002-05-01

    The physics of the fly-ball governor, introduced to regulate the speed of steam engines, is here analysed anew. The original analysis is generalized to arbitrary governor geometry. The well-known stability criterion for the linearized system breaks down for large excursions from equilibrium; we show approximately how this criterion changes.

  5. 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

  6. CFD modeling of a boiler's tubes rupture

    Energy Technology Data Exchange (ETDEWEB)

    Rahimi, Masoud; Khoshhal, Abbas; Shariati, Seyed Mehdi [Chemical Engineering Department, Faculty of Engineering, Razi University, Kermanshah (Iran)

    2006-12-15

    This paper reports the results of a study on the reason for tubes damage in the superheater Platen section of the 320MW Bisotoun power plant, Iran. The boiler has three types of superheater tubes and the damage occurs in a series of elbows belongs to the long tubes. A three-dimensional modeling was performed using an in-house computational fluid dynamics (CFD) code in order to explore the reason. The code has ability of simultaneous solving of the continuity, the Reynolds-Averaged Navier-Stokes (RANS) equations and employing the turbulence, combustion and radiation models. The whole boiler including; walls, burners, air channels, three types of tubes, etc., was modeled in the real scale. The boiler was meshed into almost 2,000,000 tetrahedral control volumes and the standard k-{epsilon} turbulence model and the Rosseland radiation model were used in the model. The theoretical results showed that the inlet 18.9MPa saturated steam becomes superheated inside the tubes and exit at a pressure of 17.8MPa. The predicted results showed that the temperature of the steam and tube's wall in the long tubes is higher than the short and medium size tubes. In addition, the predicted steam mass flow rate in the long tube was lower than other ones. Therefore, it was concluded that the main reason for the rupture in the long tubes elbow is changing of the tube's metal microstructure due to working in a temperature higher than the design temperature. In addition, the structural fatigue tension makes the last elbow of the long tube more ready for rupture in comparison with the other places. The concluded result was validated by observations from the photomicrograph of the tube's metal samples taken from the damaged and undamaged sections. (author)

  7. Effect of Environment on the Stress- Rupture Behavior of a C/SiC Composite Studied

    Science.gov (United States)

    Verrilli, Michael J.; Kiser, J. Douglas; Opila, Elizabeth J.; Calomino, Anthony M.

    2002-01-01

    Advanced reusable launch vehicles will likely incorporate fiber-reinforced ceramic matrix composites (CMC's) in critical propulsion and airframe components. The use of CMC's is highly desirable to save weight, improve reuse capability, and increase performance. One of the candidate CMC materials is carbon-fiber-reinforced silicon carbide (C/SiC). In potential propulsion applications, such as turbopump rotors and nozzle exit ramps, C/SiC components will be subjected to a service cycle that includes mechanical loading under complex, high-pressure environments containing hydrogen, oxygen, and steam. Degradation of both the C fibers and the SiC matrix are possible in these environments. The objective of this effort was to evaluate the mechanical behavior of C/SiC in various environments relevant to reusable launch vehicle applications. Stress-rupture testing was conducted at the NASA Glenn Research Center on C/SiC specimens in air and steam-containing environments. Also, the oxidation kinetics of the carbon fibers that reinforce the composite were monitored by thermogravimetric analysis in the same environments and temperatures used for the stress-rupture tests of the C/SiC composite specimens. The stress-rupture lives obtained for C/SiC tested in air and in steam/argon mixtures are shown in the following bar chart. As is typical for most materials, lives obtained at the lower temperature (600 C) are longer than for the higher temperature (1200 C). The effect of environment was most pronounced at the lower temperature, where the average test duration in steam at 600 C was at least 30 times longer than the lives obtained in air. The 1200 C data revealed little difference between the lives of specimens tested in air and steam at atmospheric pressure.

  8. Physics of Earthquake Rupture Propagation

    Science.gov (United States)

    Xu, Shiqing; Fukuyama, Eiichi; Sagy, Amir; Doan, Mai-Linh

    2018-05-01

    A comprehensive understanding of earthquake rupture propagation requires the study of not only the sudden release of elastic strain energy during co-seismic slip, but also of other processes that operate at a variety of spatiotemporal scales. For example, the accumulation of the elastic strain energy usually takes decades to hundreds of years, and rupture propagation and termination modify the bulk properties of the surrounding medium that can influence the behavior of future earthquakes. To share recent findings in the multiscale investigation of earthquake rupture propagation, we held a session entitled "Physics of Earthquake Rupture Propagation" during the 2016 American Geophysical Union (AGU) Fall Meeting in San Francisco. The session included 46 poster and 32 oral presentations, reporting observations of natural earthquakes, numerical and experimental simulations of earthquake ruptures, and studies of earthquake fault friction. These presentations and discussions during and after the session suggested a need to document more formally the research findings, particularly new observations and views different from conventional ones, complexities in fault zone properties and loading conditions, the diversity of fault slip modes and their interactions, the evaluation of observational and model uncertainties, and comparison between empirical and physics-based models. Therefore, we organize this Special Issue (SI) of Tectonophysics under the same title as our AGU session, hoping to inspire future investigations. Eighteen articles (marked with "this issue") are included in this SI and grouped into the following six categories.

  9. The relative impact of sizing errors on steam generator tube failure probability

    International Nuclear Information System (INIS)

    Cizelj, L.; Dvorsek, T.

    1998-01-01

    The Outside Diameter Stress Corrosion Cracking (ODSCC) at tube support plates is currently the major degradation mechanism affecting the steam generator tubes made of Inconel 600. This caused development and licensing of degradation specific maintenance approaches, which addressed two main failure modes of the degraded piping: tube rupture; and excessive leakage through degraded tubes. A methodology aiming at assessing the efficiency of a given set of possible maintenance approaches has already been proposed by the authors. It pointed out better performance of the degradation specific over generic approaches in (1) lower probability of single and multiple steam generator tube rupture (SGTR), (2) lower estimated accidental leak rates and (3) less tubes plugged. A sensitivity analysis was also performed pointing out the relative contributions of uncertain input parameters to the tube rupture probabilities. The dominant contribution was assigned to the uncertainties inherent to the regression models used to correlate the defect size and tube burst pressure. The uncertainties, which can be estimated from the in-service inspections, are further analysed in this paper. The defect growth was found to have significant and to some extent unrealistic impact on the probability of single tube rupture. Since the defect growth estimates were based on the past inspection records they strongly depend on the sizing errors. Therefore, an attempt was made to filter out the sizing errors and to arrive at more realistic estimates of the defect growth. The impact of different assumptions regarding sizing errors on the tube rupture probability was studied using a realistic numerical example. The data used is obtained from a series of inspection results from Krsko NPP with 2 Westinghouse D-4 steam generators. The results obtained are considered useful in safety assessment and maintenance of affected steam generators. (author)

  10. Analysis of Ruptured Heater Tube of Degasser Condenser in Wolsong Unit 4

    International Nuclear Information System (INIS)

    Kim, Hong Pyo; Kim, J. S.; Lim, Y. S.; Kim, S. S.; Hwang, S. S.; Kim, D. J.; Kim, S. W.; Jeong, M. K.; Hong, J. H.

    2007-08-01

    In a degasser condenser in Wolsong unit 4, the cracks were found in the heater tube no. 6 and no. 7. To avoid additional damages in the specimen during a decontamination process for the previous analysis, the cracks were analyzed without any decontamination process in this work. We performed the investigation of the ruptured surface morphology, the EDS analysis of the ruptured surface, the microstructural analysis of Alloy 800H sheath tube and literature survey to find the failure mechanism. From the results, it was expected that the sheath tube has been exposed in a steam condition as the coolant level was decreased in the degasser condenser, leading to the rupture of the sheath tube

  11. Steam generators in PWR's

    International Nuclear Information System (INIS)

    Michel, R.

    1974-01-01

    The steam generator of the PWR operates according to the principle of natural circulation. It consists of a U-shaped tube bundle whose free ends are welded to a bottom plate. The tube bundle is surrounded by a cylinder jacket which has slots closely above the bottom or tube plate. The feed water mixed with boiling water enters the tube bundle through these slots. Because of its buoyancy, the steam-water mixture flows upwards. Below the tube plate there are chambers for distributing and collecting pressurized water separated by means of a partition wall. By omitting some tubes, a free alloy is created so that the tubes in the center get sufficient water, too. By asymmetrical arrangement of the partition wall it is further possible to limit the tube alloy only to the inlet side for pressurized water. The flow over the tube plate is thus improved on the inlet side. (DG) [de

  12. Vertical steam generator

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  13. Simultaneous bilateral patellar tendon rupture.

    Science.gov (United States)

    Moura, Diogo Lino; Marques, José Pedro; Lucas, Francisco Manuel; Fonseca, Fernando Pereira

    2017-01-01

    Bilateral patellar tendon rupture is a rare entity, often associated with systemic diseases and patellar tendinopathy. The authors report a rare case of a 34-year-old man with simultaneous bilateral rupture of the patellar tendon caused by minor trauma. The patient is a retired basketball player with no past complaints of chronic knee pain and a history of steroid use. Surgical management consisted in primary end-to-end tendon repair protected temporarily with cerclage wiring, followed by a short immobilization period and intensive rehabilitation program. Five months after surgery, the patient was able to fully participate in sport activities.

  14. Simultaneous bilateral patellar tendon rupture

    Directory of Open Access Journals (Sweden)

    Diogo Lino Moura

    Full Text Available ABSTRACT Bilateral patellar tendon rupture is a rare entity, often associated with systemic diseases and patellar tendinopathy. The authors report a rare case of a 34-year-old man with simultaneous bilateral rupture of the patellar tendon caused by minor trauma. The patient is a retired basketball player with no past complaints of chronic knee pain and a history of steroid use. Surgical management consisted in primary end-to-end tendon repair protected temporarily with cerclage wiring, followed by a short immobilization period and intensive rehabilitation program. Five months after surgery, the patient was able to fully participate in sport activities.

  15. Spontaneous rupture of vaginal enterocele

    DEFF Research Database (Denmark)

    Svendsen, J H; Galatius, H; Hansen, P K

    1985-01-01

    Spontaneous rupture of an enterocele is a rare complication. Only 24 cases including the present case have been reported in the literature. The patients were elderly and had had at least one vaginal operation. The patients were remarkably unaffected symptomatically on admission.......Spontaneous rupture of an enterocele is a rare complication. Only 24 cases including the present case have been reported in the literature. The patients were elderly and had had at least one vaginal operation. The patients were remarkably unaffected symptomatically on admission....

  16. Rupture luminescence from natural fibers

    Science.gov (United States)

    Li, W.; Haneman, D.

    1999-12-01

    Fibers of cotton and wool, and samples of paper, have been ruptured in tension in vacuum and in air, and give detectable luminescence in the visible range. All have a common emission peak at around 2.0 eV, which is ascribed to the deexcitation of states excited by the rupture of organic chain molecule bonds. Rubber bands give stronger emission in air, but no emission in vacuum, suggesting the material breaks only at weak interchain bonds. Mohair, cat, and horse hair also give emission in air. The phenomena reveal effects that would occur widely in nature.

  17. Steam generator life management

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  18. 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

  19. Analysis of 30 breast implant rupture cases.

    Science.gov (United States)

    Tark, Kwan Chul; Jeong, Hii Sun; Roh, Tae Suk; Choi, Jong Woo

    2005-01-01

    Breast implants used for augmentation mammoplasty or breast reconstruction could rupture from various causes such as trauma or spontaneous failure. The objectives of this study were to investigate the relationships between the causes of implant rupture and the degree of capsular contracture, and then to evaluate the relative efficacies of specific signs on magnetic resonance imaging (MRI) known to be beneficial for diagnosing the rupture. A retrospective review identified patients with prosthetic implant rupture or impending rupture treated by the senior author. The 30 cases of implant rupture available for review were classified into two groups: intracapsular and extracapsular ruptures. The 30 cases of breast implant ruptures were analyzed with respect to the clinical symptoms and signs, the causes of rupture, the degree of capsular contracture, and therapeutic plans. Among the 30 cases, 14 patients who had undergone MRI during the diagnostic period were analyzed with respect to the relationships between MRI readings and operative findings. Spontaneous rupture of membranes was most common (80%), followed by failure because of trauma (7%) and valve or implant base (4%). The symptoms during implant rupture were contour deformity, palpated mass-like lesions, pain, and focal inflammation. According to the analysis of specific MRI signs, the sensitivity and specificity of the linguine sign were 87% and 100%, respectively, for intracapsular rupture. For extracapsular rupture, the sensitivity and specificity of the linguine sign were, respectively, 67% and 75%. The sensitivity and specificity of the rat-tail sign and tear drop sign were 14% and 50%, respectively. Breast implant rupture was correlated with the degree of capsular contracture in our study. Among the various specific MRI signs used in diagnosing the rupture, the linguine sign was reliable and had a high sensitivity and specificity, especially in cases of intracapsular rupture. On the other hand, the rat

  20. Rupture of Achilles Tendon : Usefulness of Ultrasonography

    International Nuclear Information System (INIS)

    Kim, Nam Hyeon; Ki, Won Woo; Yoon, Kwon Ha; Kim, Song Mun; Shin, Myeong Jin; Kwon, Soon Tae

    1996-01-01

    To differentiate a complete rupture of Achilles tendon from an incomplete one which is important because its treatment is quite different. And it is necessary to know the exact site of the rupture preoperatively. Fifteen cases of fourteen patients which were diagnosed as Achilles tendon rupture by ultrasonography and surgery were reviewed. We compared sonographic rupture site with surgical findings. Ultrasonographic criteria for differentiation of complete and incomplete rupture was defined as follows : the discreteness, which means the proximal intervening hypoechogenicity to the interface echogenicity of distal margin of ruptured tendon : the slant sign, which represents the interface of ruptured distal margin which was seen over the 3/4 of the thickness of the tendon without intervening low echogeneicity : the invagination sign, which means the echogenic invagination from Kager triangle into posterior aspect of Achilles tendon over the half thickness of the tendon. The sites of complete tendon rupture were exactly corresponded to surgical finding in four cases of ten complete ruptures. And the discrepancy between sonographic and surgical findings in the site of complete rupture was 1.2 ± 0.4 cm in six cases. Three of ten complete ruptures showed the discreteness sign, all of ten showed the slant sign and two of ten showed the invagination sign. It is helpful to differentiate a complete from incomplete rupture of the Achilles tendon and to localize the site of the complete rupture with the ultrasonographic evaluation

  1. 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.

  2. Fatigue and creep cracking of nickel alloys for 700 C steam turbines

    International Nuclear Information System (INIS)

    Berger, C.; Granacher, J.; Thoma, A.; Roesler, J.; Del Genovese, D.

    2001-01-01

    Four materials of the types Inconel 706 (two heat treatment states), Inconel 617, and Waspaloy were tested as shaft materials for 700 to 720 C steam turbines. At an extrapolation time ratio of 10, Waspaloy was expected to have the highest creep strength (about 270 MPa at 700 C), with values of about 140 MPa at 700 C for Inconel 617. A preliminary evaluation of the 700 C creep rupture tests showed the highest creep rupture resistance for Inconel 617, followed by Waspaloy and Inconel 706 [de

  3. 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)

  4. Thermal hydraulic aspects of steam drum level control philosophy for the natural circulation based heavy water reactor

    International Nuclear Information System (INIS)

    Gupta, S.K.; Gaikwad, A.J.; Kumar, Rajesh

    2004-01-01

    From safety considerations advanced nuclear reactors rely more and more on passive systems such as natural circulation for primary heat removal. A natural circulation based water reactor is relatively larger in size so as to reduce flow losses and channel type for proper flow distribution. From the size of steam drum considerations it has to be multi loop but has a common inlet header. Normally the turbine follows the reactor. This paper addresses the thermal hydraulic aspects of the steam drum pressure and level control philosophy for a four drum, natural circulation based, channel type boiling water advanced reactor. Three philosophies may be followed for drum control viz. individual drum control, one control drum approach and an average of all the four drums. For drum pressure control, the steam flow to the turbine is be regulated. A single point pressure control is better than individual drum pressure control. This is discussed in the paper. But the control point has to be at a place down steam the point where all steam line from individual drum meet. This may lead to different pressure in all the four drums depending on the power produced in the respective loops. The difference in pressure cannot be removed even if the four drums are directly connected through pipes. Also the pressure control scheme with/without interconnection is discussed. For level, the control of individual drum may not be normally possible because of common inlet header. As the frictional pressure drops in the large diameter downcomers are small as compared to elevation pressure drops, the level in all the steam drum tend to equalize. Consequently a single representative drum level may be chosen as a control variable for controlling level in all the four drums. But in case, where all the four loops are producing different powers and single point pressure control is effective, the scheme may not work satisfactorily. the level in a drum may depend on the power produced in the loop

  5. 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

  6. Pacemaker syndrome with sub-acute left ventricular systolic dysfunction in a patient with a dual-chamber pacemaker: consequence of lead switch at the header.

    Science.gov (United States)

    Khurwolah, Mohammad Reeaze; Vezi, Brian Zwelethini

    In the daily practice of pacemaker insertion, the occurrence of atrial and ventricular lead switch at the pacemaker box header is a rare and unintentional phenomenon, with less than five cases reported in the literature. The lead switch may have dire consequences, depending on the indication for the pacemaker. One of these consequences is pacemaker syndrome, in which the normal sequence of atrial and ventricular activation is impaired, leading to sub-optimal ventricular filling and cardiac output. It is important for the attending physician to recognise any worsening of symptoms in a patient who has recently had a permanent pacemaker inserted. In the case of a dual-chamber pacemaker, switching of the atrial and ventricular leads at the pacemaker box header should be strongly suspected. We present an unusual case of pacemaker syndrome and right ventricular-only pacinginduced left ventricular systolic dysfunction in a patient with a dual-chamber pacemaker.

  7. A study on the evaluation of vibration effect and the development of vibration reduction method for Wolsung unit 1 main steam piping

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hyun; Kim, Yeon Whan [Korea Electric Power Corp. (KEPCO), Taejon (Korea, Republic of). Research Center; Kim, Tae Ryong; Park, Jin Ho [Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of)

    1996-08-01

    The main steam piping of nuclear power plant which runs between steam generator and high pressure turbine has been experienced to have a severe effect on the safe operation of the plant due to the vibration induced by the steam flowing inside the piping. The imposed cyclic loads by the vibration could result in the degradation of the related structures such as connection parts between main instruments, valves, pipe supports and building. The objective of the study is to reduce the vibration level of Wolsung nuclear power plant unit 1 main steam pipeline by analyzing vibration characteristics of the piping, identifying sources of the vibration and developing a vibration reduction method .The location of the maximum vibration is piping between the main steam header and steam chest .The stress level was found to be within the allowable limit .The main vibration frequency was found to be 4{approx}6 Hz which is the same as the natural frequency from model test .A vibration reduction method using pipe supports of energy absorbing type(WEAR)is selected .The measured vibration level after WEAR installation was reduced about 36{approx}77% in displacement unit (author). 36 refs., 188 figs.

  8. PREMATURE RUPTURE OF THE MEMBRANES*

    African Journals Online (AJOL)

    In patients presenting with premature rupture of the membranes there are two factors which influence the foetal morbidity and mortality. These factors are prema- turity and intra-uterine infection. The purpose of this analysis was to elucidate which factor carried the greater risk to the foetus. Recently there has been a spate of.

  9. Ruptured Cervical Anellrysm with Neurofibromatosis

    African Journals Online (AJOL)

    rupture of an aneurysm of a lateral branch of the right thyrocervical trunk in a patient suffering from diffuse neurofibromatosis. The operative findings are reported. s. Air. Med. ... system demonstrated decreased power in all muscles of the right shoulder girdle and arm, without sensory change. The cranial nerves were normal.

  10. Spontaneous Splenic Rupture in Melanoma

    Directory of Open Access Journals (Sweden)

    Hadi Mirfazaelian

    2014-01-01

    Full Text Available Spontaneous rupture of spleen due to malignant melanoma is a rare situation, with only a few case reports in the literature. This study reports a previously healthy, 30-year-old man who came with chief complaint of acute abdominal pain to emergency room. On physical examination, abdominal tenderness and guarding were detected to be coincident with hypotension. Ultrasonography revealed mild splenomegaly with moderate free fluid in abdominopelvic cavity. Considering acute abdominal pain and hemodynamic instability, he underwent splenectomy with splenic rupture as the source of bleeding. Histologic examination showed diffuse infiltration by tumor. Immunohistochemical study (positive for S100, HMB45, and vimentin and negative for CK, CD10, CK20, CK7, CD30, LCA, EMA, and chromogranin confirmed metastatic malignant melanoma. On further questioning, there was a past history of a nasal dark skin lesion which was removed two years ago with no pathologic examination. Spontaneous (nontraumatic rupture of spleen is an uncommon situation and it happens very rarely due to neoplastic metastasis. Metastasis of malignant melanoma is one of the rare causes of the spontaneous rupture of spleen.

  11. Study of steam condensation at sub-atmospheric pressure: setting a basic research using MELCOR code

    Science.gov (United States)

    Manfredini, A.; Mazzini, M.

    2017-11-01

    One of the most serious accidents that can occur in the experimental nuclear fusion reactor ITER is the break of one of the headers of the refrigeration system of the first wall of the Tokamak. This results in water-steam mixture discharge in vacuum vessel (VV), with consequent pressurization of this container. To prevent the pressure in the VV exceeds 150 KPa absolute, a system discharges the steam inside a suppression pool, at an absolute pressure of 4.2 kPa. The computer codes used to analyze such incident (eg. RELAP 5 or MELCOR) are not validated experimentally for such conditions. Therefore, we planned a basic research, in order to have experimental data useful to validate the heat transfer correlations used in these codes. After a thorough literature search on this topic, ACTA, in collaboration with the staff of ITER, defined the experimental matrix and performed the design of the experimental apparatus. For the thermal-hydraulic design of the experiments, we executed a series of calculations by MELCOR. This code, however, was used in an unconventional mode, with the development of models suited respectively to low and high steam flow-rate tests. The article concludes with a discussion of the placement of experimental data within the map featuring the phenomenon characteristics, showing the importance of the new knowledge acquired, particularly in the case of chugging.

  12. 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...

  13. Endometriosis-related spontaneous diaphragmatic rupture.

    Science.gov (United States)

    Triponez, Frédéric; Alifano, Marco; Bobbio, Antonio; Regnard, Jean-François

    2010-10-01

    Non-traumatic, spontaneous diaphragmatic rupture is a rare event whose pathophysiology is not known. We report the case of endometriosis-related spontaneous rupture of the right diaphragm with intrathoracic herniation of the liver, gallbladder and colon. We hypothesize that the invasiveness of endometriotic tissue caused diaphragm fragility, which finally lead to its complete rupture without traumatic event. The treatment consisted of a classical management of diaphragmatic rupture, with excision of the endometriotic nodule followed by medical ovarian suppression for six months.

  14. ACL Rupture in Collegiate Wrestler

    Directory of Open Access Journals (Sweden)

    Lindsay A. Palmer

    2016-05-01

    Full Text Available Objective: To educate others on unique Anterior Cruciate Ligament tears and percentage of usage of the ACL in normal daily function. Background: Patient is an eighteen year old male participating in wrestling and football at the time of the injury. Patient now only participates in wrestling. No previous knee or chronic injuries were reported prior to this injury. Patient was playing football during the time of injury. The patient stated that he planted his foot down and was tackled at the same time when the injury occurred. The patient felt his knee twist and buckle. Patient complained of clicking inside the knee and had minimal swelling. He also complained of it being difficult to bear weight at the time. The patient did not seek further treatment until two months after the injury occurred when he received an MRI. His MRI showed a positive finding for an Anterior Cruciate Ligament rupture. His previous Athletic Trainer could not find a positive diagnosis for the patient prior to the MRI. Differential Diagnosis: Possible meniscal or ACL injury. Treatment: Doctors officially diagnosed the injury as a complete rupture of the ACL. The patient did not receive surgery immediately. Doctors have stated that he only uses about 50% of his ACL on a daily basis compared to a normal person who uses about 95% of their ACL daily. Because of this, the patient played on his rupture for seven months before receiving surgery. He played a whole season of high school football and a whole season of wrestling his senior year with the ACL ruptured. The patient only used a brace for better comfort during the seven months. The patient then received reconstructive surgery to repair the rupture. A hamstring tendon graft was used to repair the ruptured ACL. Because a tendon was taken from the hamstring, patient experienced a tight ACL and hamstring of the left leg post-surgery. The patient participated in Physical Therapy for five months to strengthen and stretch the new

  15. A Document-Based EHR System That Controls the Disclosure of Clinical Documents Using an Access Control List File Based on the HL7 CDA Header.

    Science.gov (United States)

    Takeda, Toshihiro; Ueda, Kanayo; Nakagawa, Akito; Manabe, Shirou; Okada, Katsuki; Mihara, Naoki; Matsumura, Yasushi

    2017-01-01

    Electronic health record (EHR) systems are necessary for the sharing of medical information between care delivery organizations (CDOs). We developed a document-based EHR system in which all of the PDF documents that are stored in our electronic medical record system can be disclosed to selected target CDOs. An access control list (ACL) file was designed based on the HL7 CDA header to manage the information that is disclosed.

  16. ANL/CANTIA code for steam generator tube integrity assessment

    International Nuclear Information System (INIS)

    Revankar, S.T.; Wolf, B.; Majumdar, S.; Riznic, J.R.

    2009-01-01

    Steam generator (SG) tubes have an important safety role in CANDU type reactors and Pressurized Water Reactors (PWR) because they constitute one of the primary barriers between the radioactive and non-radioactive sides of the nuclear plant. The SG tubes are susceptible to corrosion and damage. A failure of a single steam generator tube, or even a few tubes, would not be a serious safety-related event in a CANDU reactor. The leakage from a ruptured tube is within makeup capacity of the primary heat transport system, so that as long as the operator takes the correct actions, the off-site consequences will be negligible. A sufficient safety margin against tube rupture used to be the basis for a variety of maintenance strategies developed to maintain a suitable level of plant safety and reliability. Several through-wall flaws may remain in operation and potentially contribute to the total primary-to-secondary leak rate. Assessment of the conditional probabilities of tube failures, leak rates, and ultimately risk of exceeding licensing dose limits has been used for steam generator tube fitness-for-service assessment. The advantage of this type of analysis is that it avoids the excessive conservatism typically present in deterministic methodologies. However, it requires considerable effort and expense to develop all of the failure, leakage, probability of detection, and flaw growth distributions and models necessary to obtain meaningful results from a probabilistic model. The Canadian Nuclear Safety Commission (CNSC) recently developed the CANTIA methodology for probabilistic assessment of inspection strategies for steam generator tubes as a direct effect on the probability of tube failure and primary-to-secondary leak rate Recently Argonne National Laboratory has developed tube integrity and leak rate models under Integrated Steam Generator Tube Integrity Program (ISGTIP-2). These models have been incorporated in the ANL/CANTIA code. This paper presents the ANL

  17. A Retrospective Analysis of Ruptured Breast Implants

    Directory of Open Access Journals (Sweden)

    Woo Yeol Baek

    2014-11-01

    Full Text Available BackgroundRupture is an important complication of breast implants. Before cohesive gel silicone implants, rupture rates of both saline and silicone breast implants were over 10%. Through an analysis of ruptured implants, we can determine the various factors related to ruptured implants.MethodsWe performed a retrospective review of 72 implants that were removed for implant rupture between 2005 and 2014 at a single institution. The following data were collected: type of implants (saline or silicone, duration of implantation, type of implant shell, degree of capsular contracture, associated symptoms, cause of rupture, diagnostic tools, and management.ResultsForty-five Saline implants and 27 silicone implants were used. Rupture was diagnosed at a mean of 5.6 and 12 years after insertion of saline and silicone implants, respectively. There was no association between shell type and risk of rupture. Spontaneous was the most common reason for the rupture. Rupture management was implant change (39 case, microfat graft (2 case, removal only (14 case, and follow-up loss (17 case.ConclusionsSaline implants have a shorter average duration of rupture, but diagnosis is easier and safer, leading to fewer complications. Previous-generation silicone implants required frequent follow-up observation, and it is recommended that they be changed to a cohesive gel implant before hidden rupture occurs.

  18. A Study on Thermal Performance of a Novel All-Glass Evacuated Tube Solar Collector Manifold Header with an Inserted Tube

    Directory of Open Access Journals (Sweden)

    Jichun Yang

    2015-01-01

    Full Text Available A novel all-glass evacuated tube collector manifold header with an inserted tube is proposed in this paper which makes water in all-glass evacuated solar collector tube be forced circulated to improve the performance of solar collector. And a dynamic numerical model was presented for the novel all-glass evacuated tube collector manifold header water heater system. Also, a test rig was built for model validation and comparison with traditional all-glass evacuated tube collector. The experiment results show that the efficiency of solar water heater with a novel collector manifold header is higher than traditional all-glass evacuated tube collector by about 5% and the heat transfer model of water heater system is valid. Based on the model, the relationship between the average temperature of water tank and inserted tube diameter (water mass flow has been studied. The results show that the optimized diameter of inserted tube is 32 mm for the inner glass with the diameter of 47 mm and the water flow mass should be less than 1.6 Kg/s.

  19. Analysis of the dynamic response of a double rupture disc assembly to simulated sodium-water reaction pressure pulses

    International Nuclear Information System (INIS)

    Leonard, J.R.

    1980-03-01

    A series of double rupture disc experiments were conducted in 1979 to evaluate the dynamic response characteristics of this pressure relief apparatus. The tests were performed in a facility with water simulating sodium and rising pressure pulses representative of the pressure increase resulting from a water/steam leak from a steam generator into sodium in the intermediate heat transport system of a breeder reactor power plant. Maximum source pressures ranged in magnitude from 50 psi to 800 psi. Dynamic response characteristics of each of the two rupture discs were similar to those observed in larger scale sodium-water experiments conducted in the Series I and Series II Large Leak Test Program at the Energy Technology Engineering Center. The SRI double rupture disc dynamic behavior was found to be consistent and amendable to modelling in the TRANSWRAP II computer code. A series of correlations which represent rupture disc buckling parameters were developed for use in the TRANSWRAP II code. The semi-empirical modeling of the rupture discs in the TRANSWRAP II code showed very good agreement with the experimental results

  20. Aerosol retention in the flooded steam generator bundle during SGTR

    International Nuclear Information System (INIS)

    Lind, Terttaliisa; Dehbi, Abdel; Guentay, Salih

    2011-01-01

    Research highlights: → High retention of aerosol particles in a steam generator bundle flooded with water. → Increasing particle inertia, i.e., particle size and velocity, increases retention. → Much higher retention of aerosol particles in the steam generator bundle flooded with water than in a dry bundle. → Much higher retention of aerosol particles in the steam generator bundle than in a bare pool. → Bare pool models have to be adapted to be applicable for flooded bundles. - Abstract: A steam generator tube rupture in a pressurized water reactor may cause accidental release of radioactive particles into the environment. Its specific significance is in its potential to bypass the containment thereby providing a direct pathway of the radioactivity from the primary circuit to the environment. Under certain severe accident scenarios, the steam generator bundle may be flooded with water. In addition, some severe accident management procedures are designed to minimize the release of radioactivity into the environment by flooding the defective steam generator secondary side with water when the steam generator has dried out. To extend our understanding of the particle retention phenomena in the flooded steam generator bundle, tests were conducted in the ARTIST and ARTIST II programs to determine the effect of different parameters on particle retention. The effects of particle type (spherical or agglomerate), particle size, gas mass flow rate, and the break submergence on particle retention were investigated. Results can be summarized as follows: increasing particle inertia was found to increase retention in the flooded bundle. Particle shape, i.e., agglomerate or spherical structure, did not affect retention significantly. Even with a very low submergence, 0.3 m above the tube break, significant aerosol retention took place underlining the importance of the jet-bundle interactions close to the tube break. Droplets were entrained from the water surface with

  1. A worked example using the SP249 advanced assessment route: the carregado unit 6 final superheater outlet header

    Energy Technology Data Exchange (ETDEWEB)

    Brear, J.M.; Jarvis, P.; Jones, G.T. [ERA Technology (United Kingdom); Jovanovic, A.S.; Friemann, M.; Kluttig, B.; Ober, M. [Stuttgart Univ. (Germany). Staatliche Materialpruefungsanstalt; Batista, A. [EDP-PROET (Portugal); Araujo, C.L. de; Pires, A. [ISQ (Portugal)

    1995-12-31

    As a key part of its information resource, the SP249 Project contains a number of case studies, drawn from the collective experience of the partners and from the literature. The user of the system may search this data-base by component type and material or by assessment method, to find a practical example close to his own current problem. He can thus draw upon past experience as well as state-of-the-art knowledge to obtain advice. To facilitate this, a set of key-words has been defined to create links between the case studies and the overall assessment methodology. These relate to damage and failure types and causes as well as to techniques of investigation and assessment. For demonstration, validation and didactic purposes, certain of these case studies - one per end-user utility in the project - have been chosen for full elaboration as `worked-examples`. These real component evaluations are worked through by an expert group from the project team so as to provide the utility staff with `hands-on` training in both the practical techniques of component life. The assessment and the use of the knowledge based system. The exercise also provides valuable opportunity for feedback, allowing refinement of the technology package and the software. Amongst these worked examples, an assessment of EDP`s Carregado Unit 6 Final Superheater Outlet Header has been chosen for special attention - as the operators have kindly allowed direct CSS to the component during two outages. This article summarises the Carregado Case Study. It is intended to serve as a demonstration and as to how the Advanced Assessment Route (AAR) is used in practice. The actions performed and results obtained are summarised

  2. Steam generator tube performance

    International Nuclear Information System (INIS)

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

    1984-10-01

    A review of the performance of steam generator tubes in 116 water-cooled nuclear power reactors showed that tubes were plugged at 54 (46 percent) of the reactors. The number of tubes removed from service decreased from 4 692 (0.30 percent) in 1981 to 3 222 (0.20 percent) in 1982. The leading causes of tube failures were stress corrosion cracking from the primary side, stress corrosion cracking (or intergranular attack) from the secondary side and pitting corrosion. The lowest incidence of corrosion-induced defects from the secondary side occurred in reactors that have used only volatile treatment, with or without condensate demineralization

  3. Steam generator tube performance

    International Nuclear Information System (INIS)

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

    1992-03-01

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

  4. Steam generator auxiliary systems

    International Nuclear Information System (INIS)

    Heinz, A.

    1982-01-01

    The author deals with damage and defect types obtaining in auxiliary systems of power plants. These concern water/steam auxiliary systems (feed-water tank, injection-control valves, slide valves) and air/fluegas auxiliary systems (blowers, air preheaters, etc.). Operating errors and associated damage are not dealt with; by contrast, weak spots are pointed out which result from planning and design. Damage types and events are collected in statistics in order to facilitate damage evaluation for arriving at improved design solutions. (HAG) [de

  5. 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...

  6. 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)

  7. Embrittlement of pre-hydrided Zircaloy-4 by steam oxidation under simulated LOCA transients

    Energy Technology Data Exchange (ETDEWEB)

    Desquines, J., E-mail: jean.desquines@irsn.fr; Drouan, D.; Guilbert, S.; Lacote, P.

    2016-02-15

    During a Loss Of Coolant Accident (LOCA), the mechanical behavior of high temperature steam oxidized fuel rods is an important issue. In this study, as-received and pre-hydrided axial tensile samples were steam oxidized in a vertical furnace and water quenched in order to simulate a LOCA transient. The samples were then subjected to a mechanical test to determine the failure conditions. Two different rupture modes were evidenced; the first one associated to linear elastic fracture mechanics and the second one is associated to sample failure without applied load. The oxidized cladding fracture toughness was determined relying on intensive metallographic analysis. The sample failure conditions were then back predicted confirming that the main rupture parameters are well captured.

  8. 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

  9. SGTR Project: Separate Effect Studies for Vertical Steam Generators

    Energy Technology Data Exchange (ETDEWEB)

    Peyres, V.; Polo, J.; Herranz, L. E.

    2003-07-01

    The SGTR project has been carried out within the fifth EURATOM Framework Programme (Contract No FIKS-CT-1999-0007). Its main objective was to provide an experimental database and to develop and/or verify models to support definition of accident management measures in the hypothetical case of a Steam Generator tube Rupture (SGTR) sequence. The project addressed both vertical and horizontal steam generator designs. This report summarises the main results obtained in the intermediate scale experimentation that addressed Western type steam generators. The specific goal of this test programme was to investigate aerosol retention in the break stage of the secondary side of a water-empty steam generator. The test matrix consisted of 12 tests that explored the influence of variables such as break type and orientation and inlet gas flow rate. This work was performed in the PECA facility of the Laboratory for Analysis of Safety Systems (LASS). Aerosol retention at the break stage of a dry steam generator was observed to be low and non-uniform. Neither break type nor orientation affected results significantly whenever gas flowrates exceeded about 100 kg/h. However, deposition patterns guillotine breaks and fish mouth ones showed remarkable differences. For flowrates above 100 kg/hm the higher the gas flow velocity, the lower the total mass depleted on tube bundle surfaces; however, at lower flowrates this trend was not maintained. An attempt to measure gas injection velocity at the break exit by Particle Image Velocity (PIV) was done but data were highly uncertain. (Author) 2 refs.

  10. Calculation Method of Steam Generator Level for swelling and shrinking effects in YGN 1/2 Simulator

    International Nuclear Information System (INIS)

    Hwang, Do Hyun; Seo, In Yong; Park, Weon Seo; Suh, Jae Seung

    2007-01-01

    In August 2006, the development of new simulator for YGN 1/2 Simulator was completed. The NSSS (Nuclear Steam Supply System) T/H(Thermal- Hydraulic) module in this simulator was developed with ARTS code based on RETRAN, which is a best estimate thermal-hydraulic code designed to analyze several operational transients by EPRI(Electric Power Research Institute). RETRAN, however, has some limitations in real-time calculation capability and its robustness to be used in the simulator for some transient conditions. To overcome these limitations, its robustness and real-time calculation capability have been improved with simplifications and removing of discontinuities of the physical correlations of the RETRAN code. And some supplements are also developed to extend its simulation scope of the ARTS code. In comparison to KNPEC(Kori Nuclear Power Education Center) no.2 simulator, the simulator based on Younggwang Unit 1 developed in the year 2001, the ARTS code was upgraded that it extended its calculating region to the steam line and common header before turbine while it had calculated to the steam generator exit before steam line in KNPEC no.2 simulator. Consequently, the number of volume and fill/normal junction in ARTS nodalization increased to 109 and 174 from 62 and 125, respectively

  11. Header [nuclear battery

    International Nuclear Information System (INIS)

    Goslee, D.E.; Barr, H.N.

    1976-01-01

    The invention relates to a nuclear-powered microwatt thermoelectric generator, small in size, efficient in operation, and which will last for a considerable period of time. The generator is suitable for implanting within the human body for powering devices such as cardiac pacemakers

  12. Electricity from geothermal steam

    Energy Technology Data Exchange (ETDEWEB)

    Wheatcroft, E L.E.

    1959-01-01

    The development of the power station at Wairakei geothermal field is described. Wairakei is located at the center of New Zealand's volcanic belt, which lies within a major graben which is still undergoing some degree of downfaulting. A considerable number of wells, some exceeding 610 m, have been drilled. Steam and hot water are produced from both deep and shallow wells, which produce at gauge pressures of 1.5 MPa and 0.6 MPa, respectively. The turbines are fed by low, intermediate, and high pressure mains. The intermediate pressure turbine bank was installed as a replacement for a heavy water production facility which had originally been planned for the development. Stage 1 includes a 69 MW plant, and stage 2 will bring the capacity to 150 MW. A third stage, which would bring the output up to 250 MW had been proposed. The second stage involves the installation of more high pressure steam turbines, while the third stage would be powered primarily by hot water flashing. Generation is at 11 kV fed to a two-section 500 MVA board. Each section of the board feeds through a 40 MVA transformer to a pair of 220 V transmission lines which splice into the North Island grid. Other transformers feed 400 V auxiliaries and provide local supply.

  13. Steam generator materials

    International Nuclear Information System (INIS)

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

    2002-05-01

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

  14. Steam hydrocarbon cracking and reforming

    NARCIS (Netherlands)

    Golombok, M.

    2004-01-01

    Many industrial chemical processes are taught as distinct contrasting reactions when in fact the unifying comparisons are greater than the contrasts. We examine steam hydrocarbon reforming and steam hydrocarbon cracking as an example of two processes that operate under different chemical reactivity

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-15

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

  16. Renal allograft rupture: US diagnosis

    International Nuclear Information System (INIS)

    Maklad, N.F.

    1987-01-01

    The US appearances in seven pathologically and/or surgically proved cases of renal allograft rupture are presented. These include a triangular or amorphous echogenic area in the cortex and medulla in a polar location, an echogenic band or wavy, branching anechoic lines in the hyperechoic region, a subcapsular hematoma, and an extrarenal hematoma in direct continuity with the echogenic area. Duplex Doppler examination in renal allograft rupture shows marked reduction of absence of the diastolic component of the velocity waveform in the arcuate and interlobar arteries, with reduction in amplitude of the systolic wave form. Correlation of the US appearances with gross and microscopic pathologic findings indicates that the echogenic area is due to an intrarenal hematoma, while the echogenic band represents the cortical laceration with adherent blood clots. The US-duplex Doppler examination should be the primary diagnostic modality in this life-threatening condition

  17. 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

  18. 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

  19. 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

  20. Achilles tendon rupture in badminton.

    Science.gov (United States)

    Kaalund, S; Lass, P; Høgsaa, B; Nøhr, M

    1989-01-01

    The typical badminton player with an Achilles tendon rupture is 36 years old and, despite limbering up, is injured at the rear line in a sudden forward movement. He resumes work within three months and has a slight lack of dorsiflexion in the ankle as the main complication. Most patients resume badminton within one year, but some finish their sports career, mainly due to fear of a new injury. The investigation discusses predisposing factors and prophylactic measures. PMID:2605439

  1. An integrated leak detection system for the ALMR steam generator

    International Nuclear Information System (INIS)

    Dayal, Y.; Gaubatz, D.C.; Wong, K.K.; Greene, D.A.

    1995-01-01

    The steam generator (SG) of the Advanced Liquid Metal Reactor (ALMR) system serves as a heat exchanger between the shell side secondary loop hot liquid sodium and the tube side water/steam mixture. A leak in the tube will result in the injection of the higher pressure water/steam into the sodium and cause an exothermic sodium-water reaction. An initial small leak (less than 1 gm/sec) can escalate into an intermediate size leak in a relatively short time by self enlargement of the original flaw and by initiating leaks in neighboring tubes. If not stopped, complete rupture of one or more tubes can cause injection rates of thousands of gm/sec and result in the over pressurization of the secondary loop rupture disk and dumping of the sodium to relieve pressure. The down time associated with severe sodium-water reaction damage has great adverse economic consequence. An integrated leak detection system (ILDS) has been developed which utilizes both chemical and acoustic sensors for improved leak detection. The system provides SG leak status to the reactor operator and is reliable enough to trigger automatic control action to protect the SG. The ILDS chemical subsystem uses conventional in-sodium and cover gas hydrogen detectors and incorporates knowledge based effects due to process parameters for improved reliability. The ILDS acoustic subsystem uses an array of acoustic sensors and incorporates acoustic beamforming technology for highly reliable and accurate leak identification and location. The new ILDS combines the small leak detection capability of the chemical system with the reliability and rapid detection/location capability of the acoustic system to provide a significantly improved level of protection for the SG over a wide range of operation conditions. (author)

  2. 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...

  3. Steam generator tube performance

    International Nuclear Information System (INIS)

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

    1982-04-01

    The performance of steam generator tubes in water-cooled nuclear power reactors has been reviewed for 1980. Tube defects occurred at 38% of the 97 reactors surveyed. This is a marginal improvement over 1979 when defects occurred at 41% of the reactors. The number of failed tubes was also lower, 0.14% of the tubes in service in 1980 compared with 0.20% of those in service in 1979. Analysis of the causes of these failures indicates that stress corrosion cracking was the leading failure mechanism. Reactors that used all-volatile treatment of secondary water, with or without full-flow condensate demineralization since start-up showed the lowest incidence of corrosion-related defects

  4. Steam generator tube performance

    International Nuclear Information System (INIS)

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

    1983-08-01

    A review of the performance of steam generator tubes in 110 water-cooled nuclear power reactors showed that tubes were plugged at 46 (42 percent) of the reactors. The number of tubes removed from service increased from 1900 (0.14 percent) in 1980 to 4692 (0.30 percent) in 1981. The leading causes of tube failures were stress corrosion cracking from the primary side, stress corrosion cracking (or intergranular attack) from the secondary side and pitting corrosion. The lowest incidence of corrosion-induced defects from the secondary side occurred in reactors that used all-volatile treatment since start-up. At one reactor a large number of degraded tubes were repaired by sleeving which is expected to become an important method of tube repair in the future

  5. Structural and leakage integrity assessment of WWER steam generator tubes

    International Nuclear Information System (INIS)

    Splichal, K.; Otruba, J.; Keilova, E.; Krhounek, V.; Turek, J.

    1996-01-01

    The leakage and plugging limits were derived for WWER steam generators based on leak and burst tests using tubes with axial part-through and through-wall defects. The following conclusions were arrived at: (i) The permissible primary-to-secondary leak rate with respect to the permissible through-wall defect size of WWER-440 and WWER-1000 steam generator tubes is 8 l/h. (ii) The primary-to-secondary leak rate is reduced by the blocking of the tube cracks by corrosion product particles and other substances. (iii) The rate of crack penetration through the tube wall is higher than the crack widening. (iv) The validity of the criterion of instability for tubes with through-wall cracks was confirmed experimentally. For the WWER-440 and WWER-1000 steam generators, the critical size of axial through-wall cracks, for the threshold primary-to-secondary pressure difference, is 13.8 and 12.0 mm, respectively. (v) The calculated leakage for the rupture of one tube and for the assumed extreme defects is two orders and one order of magnitude, respectively, higher than the proposed primary water leakage limit of 8 l/h. (vi) The experiments gave evidence that the use of the permissible thinning limit of 80% for the heat exchange tube plugging does not bring about uncontrollable leakage or unstable crack growth. This is consistent with experience gained at WWER-440 type nuclear power plants. 4 tabs., 5 figs., 9 refs

  6. 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. 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)

  8. Hydrogen distribution in a containment with a high-velocity hydrogen-steam source

    International Nuclear Information System (INIS)

    Bloom, G.R.; Muhlestein, L.D.; Postma, A.K.; Claybrook, S.W.

    1982-09-01

    Hydrogen mixing and distribution tests are reported for a modeled high velocity hydrogen-steam release from a postulated small pipe break or release from a pressurizer relief tank rupture disk into the lower compartment of an Ice Condenser Plant. The tests, which in most cases used helium as a simulant for hydrogen, demonstrated that the lower compartment gas was well mixed for both hydrogen release conditions used. The gas concentration differences between any spatial locations were less than 3 volume percent during the hydrogen/steam release period and were reduced to less than 0.5 volume percent within 20 minutes after termination of the hydrogen source. The high velocity hydrogen/steam jet provided the dominant mixing mechanism; however, natural convection and forced air recirculation played important roles in providing a well mixed atmosphere following termination of the hydrogen source. 5 figures, 4 tables

  9. 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)

  10. Advanced technologies on steam generators

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  11. Steam reformer with catalytic combustor

    Science.gov (United States)

    Voecks, Gerald E. (Inventor)

    1990-01-01

    A steam reformer is disclosed having an annular steam reforming catalyst bed formed by concentric cylinders and having a catalytic combustor located at the center of the innermost cylinder. Fuel is fed into the interior of the catalytic combustor and air is directed at the top of the combustor, creating a catalytic reaction which provides sufficient heat so as to maintain the catalytic reaction in the steam reforming catalyst bed. Alternatively, air is fed into the interior of the catalytic combustor and a fuel mixture is directed at the top. The catalytic combustor provides enhanced radiant and convective heat transfer to the reformer catalyst bed.

  12. Advanced technologies on steam generators

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

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

  13. Advanced numerical description of the behavior of 700 C steam power plant components

    Energy Technology Data Exchange (ETDEWEB)

    Maile, K. [Materialpruefungsanstalt, Univ. Stuttgart (Germany); Schmidt, K.; Roos, E.; Klenk, A.; Speicher, M.

    2009-07-01

    To make full use of the strength potential of new boiler materials like the new 9-11% Cr steels and nickel based alloys, taking into account their specific stress-strain relaxation behavior, new design methods are required in the design of today's power plants. Highly loaded components are approaching more and more the classical design limits with regard to critical wall thicknesses and the related tolerable thermal gradients, due to planed increases of steam parameters like steam pressure and steam temperature. ''Design by analysis'' can be realized by modern state of the art Numerical Finite Element (FE) simulation codes and in some cases by the use of user defined advanced inelastic material laws. These material laws have to be adjusted to specific material behavior of new boiler materials. To model the strain and stress situation in components under high temperature loading, a constitutive equation based on a Graham-Walles approach is used in this paper. Furthermore essential steps and recommendations to implement experimental data in the user defined subroutines and the subsequent integration of the subroutines in modern FE codes like ABAQUS trademark and ANSYS trademark are given. As an example, the results of FE simulations of components like hollow cylinders and waterwall like components made of Alloy 617 or 9-11% Cr steels are discussed and verified with experimental results. In a last step, the successful application of the developed creep equation will be demonstrated by calculating the creep strains and stress relaxation of a P92 steam header under constant loading. (orig.)

  14. 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.

  15. Changes of the more relevant PHTS parameters after the cleaning of the steam generators primary side at Embalse nuclear power plant

    International Nuclear Information System (INIS)

    Moreno, Carlos A.; Coutsiers, Ernesto; Acevedo, Paul; Pomerantz, Marcelo E.

    2003-01-01

    During the operation of the plant magnetite deposition occurs at the inner walls of Primary Heat Transport System (PHTS). This deposition is particularly significant at the U-tubes of steam generators. The consequence of this is the deterioration of heat transfer to the Secondary System. In order to minimize this impact, during the annual outage of 2000, the steam generators primary side cleaning by the SIVABLAST technique was carried out. This technique consists in blasting the inner walls with tiny stainless steel balls propelled by air at high pressure. This paper presents the change of the more relevant parameters of PHTS after that cleaning. The parameters analyzed and the main results are the following: 1) Inlet header temperature dropped 4.7 C degrees at full power; 2) Exit quality at the outlet headers decreased from 3,5% to 1,5%; 3) Global PHTS flow in single phase evaluated from: a) In-site instrumentation increased 4,6%; b) Thermalhydraulic code NUCIRC 1.0 increased 3,2%; c) measured flows at the instrumented fuel channels increased 4.4%. (author)

  16. Neck curve polynomials in neck rupture model

    International Nuclear Information System (INIS)

    Kurniadi, Rizal; Perkasa, Yudha S.; Waris, Abdul

    2012-01-01

    The Neck Rupture Model is a model that explains the scission process which has smallest radius in liquid drop at certain position. Old fashion of rupture position is determined randomly so that has been called as Random Neck Rupture Model (RNRM). The neck curve polynomials have been employed in the Neck Rupture Model for calculation the fission yield of neutron induced fission reaction of 280 X 90 with changing of order of polynomials as well as temperature. The neck curve polynomials approximation shows the important effects in shaping of fission yield curve.

  17. Acute Pectoralis Major Rupture Captured on Video

    Directory of Open Access Journals (Sweden)

    Alejandro Ordas Bayon

    2016-01-01

    Full Text Available Pectoralis major (PM ruptures are uncommon injuries, although they are becoming more frequent. We report a case of a PM rupture in a young male who presented with axillar pain and absence of the anterior axillary fold after he perceived a snap while lifting 200 kg in the bench press. Diagnosis of PM rupture was suspected clinically and confirmed with imaging studies. The patient was treated surgically, reinserting the tendon to the humerus with suture anchors. One-year follow-up showed excellent results. The patient was recording his training on video, so we can observe in detail the most common mechanism of injury of PM rupture.

  18. Traumatic rupture of an intracranial dermoid cyst

    Directory of Open Access Journals (Sweden)

    Raksha Ramlakhan, BMedSc, MBBCh

    2015-01-01

    Full Text Available Intracranial dermoid cysts are congenital tumors of ectodermal origin. Rupture of these cysts can occur spontaneously, but rupture in association with trauma is reported infrequently. The diagnosis of rupture is made by the presence of lipid (cholesterol droplets in the subarachnoid spaces and ventricles. Nonenhanced CT of the head demonstrates multiple foci of low attenuation that correspond with hyperintense signal on T1-weighted MRI. We present a case of an adult patient with rupture of an intracranial dermoid cyst, precipitated by minor trauma.

  19. CT diagnosis of ruptured abdominal aortic aneurysm

    International Nuclear Information System (INIS)

    Sacknoff, R.; Novelline, R.A.; Wittenberg, J.; Waltman, A.C.; De Luca, S.A.; Rhea, J.T.; Lawrason, J.N.

    1986-01-01

    Ruptured abdominal aortic aneurysm (AAA) is a life-threatening condition requiring immediate diagnosis and surgery. In a series of 23 consecutive patients scanned by CT for suspected ruptured AAA, CT proved 100% accurate. In seven patients with surgically or pathologically proved ruptured AAA, CT demonstrated a similar distribution of hemorrhage into the perirenal space and to a lesser degree into the anterior and posterior pararenal spaces. The 16 true-negative examinations included ten in patients with unruptured AAA and six in patients with other diseases. The authors conclude that patients in stable condition with suspected ruptured AAA should be examined by CT

  20. Model of reverse steam generator

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  1. Steam generator thermal-hydraulics

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  2. 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

  3. Research program plan: steam generators

    International Nuclear Information System (INIS)

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

    1985-07-01

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

  4. 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)

  5. A thin-lip rupture of carbon steel superheater boiler tube

    International Nuclear Information System (INIS)

    Khalil, E.O.; Alzoye, K.S.; Elwaer, A.M.

    1993-01-01

    A ruptured A 42 medium carbon steel tube was collected by the engineering department in one of our steam power stations. Inspection of ruptured tube revealed a thin - lip fracture with brownish thin layer of oxide film on inner tube surfaces. There was no evidence of pitting, the outer surfaces of the tube exhibited a general oxidized conditions. A micro section taken near the fracture surface consists of ferrite and martensite, the amount of martensite decreased as we away from the fracture surface. Presence of martensite phase in the microstructure indicates that the tube material has been overheated. An erosion corrosion mechanism in conjunction with overheated. An erosion corrosion mechanism in conjunction with overheating resulted in strength deterioration with consequent premature failure. 4 fig., 1 tab

  6. Predicting creep rupture from early strain data

    International Nuclear Information System (INIS)

    Holmstroem, Stefan; Auerkari, Pertti

    2009-01-01

    To extend creep life modelling from classical rupture modelling, a robust and effective parametric strain model has been developed. The model can reproduce with good accuracy all parts of the creep curve, economically utilising the available rupture models. The resulting combined model can also be used to predict rupture from the available strain data, and to further improve the rupture models. The methodology can utilise unfailed specimen data for life assessment at lower stress levels than what is possible from rupture data alone. Master curves for creep strain and rupture have been produced for oxygen-free phosphorus-doped (OFP) copper with a maximum testing time of 51,000 h. Values of time to specific strain at given stress (40-165 MPa) and temperature (125-350 deg. C) were fitted to the models in the strain range of 0.1-38%. With typical inhomogeneous multi-batch creep data, the combined strain and rupture modelling involves the steps of investigation of the data quality, extraction of elastic and creep strain response, rupture modelling, data set balancing and creep strain modelling. Finally, the master curves for strain and rupture are tested and validated for overall fitting efficiency. With the Wilshire equation as the basis for the rupture model, the strain model applies classical parametric principles with an Arrhenius type of thermal activation and a power law type of stress dependence for the strain rate. The strain model also assumes that the processes of primary and secondary creep can be reasonably correlated. The rupture model represents a clear improvement over previous models in the range of the test data. The creep strain information from interrupted and running tests were assessed together with the rupture data investigating the possibility of rupture model improvement towards lower stress levels by inverse utilisation of the combined rupture based strain model. The developed creep strain model together with the improved rupture model is

  7. Three-dimensional studies of the 700 MWe steam generator design

    International Nuclear Information System (INIS)

    John, B.; Pietralik, J.

    2006-01-01

    The next stage in the Indian nuclear power programme envisions building 700 MWe Indian Pressurized Heavy Water Reactor (IPHWR) units. This involves up-rating of all the plant equipment including the reactor, steam generators (SGs), turbo-generator, major pumps, etc. The SG used in the current generation of 540 MWe IPHWRs, is a mushroom type, inverted U-tube, natural-circulation SG. The 700 MWe SG is of the same type and has the same tube bundle design and the same heat transfer area. The tube diameter, tube pitch, and outer diameter of the SG sections are the same as for the 540 MWe SG. The geometry of the feedwater header, the flow restrictor in the downcomer and the flow distribution plate are different in the two designs. The changes were required due to a 26% increase in steam flow rate while maintaining the same circulation ratio. This paper describes the design of the 700 MWe SG and a thermalhydraulic analysis using a one-dimensional, in-house code and a three-dimensional code called THIRST developed by AECL. The codes were validated against the 540 MWe SG data. The analysis was made for the 700 MWe SG for two versions: with and without integral preheater. The results of the THIRST runs were used for a flow-induced vibration analysis. The results of the flow-induced vibration analysis show that the vibrations are not excessive. (author)

  8. The stress-rupture behavior of tubes made from austenitic stainless steels and Ni-based alloys subjected to internal pressure

    International Nuclear Information System (INIS)

    Schaefer, L.; Kempe, H.

    1983-12-01

    The report outlines the stress-rupture results obtained on tubes tested as possible fuel rod cladding tubes for fast breeder reactors cooled with sodium, steam or gas. For the rupture elongations of some specimens showing a pronounced burst, higher values than in earlier reports are now indicated because of better evaluation techniques. The choice and comparisons of materials are explained, the calculations of stresses and strains are described, and reference is made to the own studies carried out to date of the parameters influencing creep-rupture behaviour. Minor modifications of the composition of an alloy and of the mechanical-thermal treatment of materials, respectively, are seen to produce clearcut changes in the stress-rupture properties. (orig.) [de

  9. 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.

  10. Common and uncommon CT findings in rupture and impending rupture of abdominal aortic aneurysms

    International Nuclear Information System (INIS)

    Ahmed, M.Z.; Ling, L.; Ettles, D.F.

    2013-01-01

    The rapid imaging evaluation and diagnosis of rupture and impending rupture of an abdominal aortic aneurysm (AAA) is imperative. This article describes the imaging findings of rupture, impending rupture, and other abdominal aortic abnormalities. It is important not to overlook AAA as the consequences can be life threatening. All patients who had open or endovascular repair of AAA rupture over 6 years (2008–2012) were identified from our departmental database. The computed tomography (CT) images of 99 patients were reviewed for relevant findings. The mean age of the patients was 65 years and 85% were male

  11. Radionuclide scan findings in delayed splenic rupture

    International Nuclear Information System (INIS)

    Flickinger, F.W.; Jackson, G.L.

    1978-01-01

    An initial liver/spleen scan performed on a patient with blunt abdominal trauma was negative 3 days following the accident. A follow-up scan 7 days later showed definite evidence of splenic rupture, proved surgically. The authors conclude that, in such cases, spleen scans may be negative initially because of delayed splenic ruptures

  12. Rupture of esophagus by compressed air.

    Science.gov (United States)

    Wu, Jie; Tan, Yuyong; Huo, Jirong

    2016-11-01

    Currently, beverages containing compressed air such as cola and champagne are widely used in our daily life. Improper ways to unscrew the bottle, usually by teeth, could lead to an injury, even a rupture of the esophagus. This letter to editor describes a case of esophageal rupture caused by compressed air.

  13. Coiling of ruptured pericallosal artery aneurysms.

    NARCIS (Netherlands)

    Menovsky, T.; Rooij, W.J.J. van; Sluzewski, M.; Wijnalda, D.

    2002-01-01

    OBJECTIVE: To assess the technical feasibility of treating ruptured pericallosal artery aneurysms with detachable coils and to evaluate the anatomic and clinical results. METHODS: Over a period of 27 months, 12 patients with a ruptured pericallosal artery aneurysm were treated with detachable

  14. Treatment Alternative for Irreparable Rotator Cuff Ruptures ...

    African Journals Online (AJOL)

    2016-09-03

    Sep 3, 2016 ... and such ruptures also lead to a pseudo-paralysis.[1,2]. Pain during daily ... 2) repairable rotator cuff rupture, as determined on MRI and during arthroscopy ..... functioning and lead to cosmetic deformities.[4]. Arthroplasty is a ...

  15. Uterine rupture without previous caesarean delivery

    DEFF Research Database (Denmark)

    Thisted, Dorthe L. A.; H. Mortensen, Laust; Krebs, Lone

    2015-01-01

    to uterine rupture when adjusted for parity, epidural analgesia and augmentation by oxytocin. CONCLUSION: Although uterine rupture is rare, its association with epidural analgesia and augmentation of labour with oxytocin in multipara should be considered. Thus, vigilance should be exercised when labour...

  16. Spontaneous rupture of choledochal cyst: case report

    International Nuclear Information System (INIS)

    Shin, Ho Seob; Nam, Kyung Jin; Lee, Jin Hwa; Kim, Chan Sung; Choi, Jong Cheol; Oh, Jong Young

    2002-01-01

    Spontaneous rupture of a choledochal cyst leading to biliary peritonitis is a rare complication which can be fatal if not promptly diagnosed. The authors report the ultrasound and CT findings of two cases of spontaneous choledochal cystic rupture and the biliary peritonitis which ensued

  17. Isolated gallbladder rupture following blunt abdominal injury

    African Journals Online (AJOL)

    2015-05-26

    May 26, 2015 ... Kaohsiung Medical University, Kaohsiung, Taiwan. Abstract. Isolated traumatic gallbladder rupture subsequent to blunt abdominal injury is rare. Most literatures on the subjects consist of case reports. We reported a rare case of isolated gallbladder rupture and discussed the possible predisposing factors to ...

  18. Spontaneous Achilles tendon rupture in alkaptonuria | Mohammed ...

    African Journals Online (AJOL)

    Spontaneous Achilles tendon ruptures are uncommon. We present a 46-year-old man with spontaneous Achilles tendon rupture due to ochronosis. To our knowledge, this has not been previously reported in Sudan literature. The tendon of the reported patient healed well after debridement and primary repairs.

  19. Spontaneous rupture of choledochal cyst: case report

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Ho Seob; Nam, Kyung Jin; Lee, Jin Hwa; Kim, Chan Sung; Choi, Jong Cheol; Oh, Jong Young [Dong-a University College of Medicine, Pusan (Korea, Republic of)

    2002-11-01

    Spontaneous rupture of a choledochal cyst leading to biliary peritonitis is a rare complication which can be fatal if not promptly diagnosed. The authors report the ultrasound and CT findings of two cases of spontaneous choledochal cystic rupture and the biliary peritonitis which ensued.

  20. Spontaneous rupture of adrenal metastasis from hepatocellular carcinoma

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Chae Hun; Kim, Hyun Jin; Park, Soo Youn; Hwang, Seong Su; Choi, Hyun Joo [St. Vincent Hospital, Suwon (Korea, Republic of)

    2007-03-15

    Rupture of adrenal tumor from various primary origins is a rather rare event. We report here on a ruptured adrenal metastasis from hepatocellular carcinoma, and this ruptured metastasis was observed at the time of the initial diagnosis.

  1. Triple Achilles Tendon Rupture: Case Report.

    Science.gov (United States)

    Saxena, Amol; Hofer, Deann

    We present a case report with 1-year follow-up data of a 57-year-old male soccer referee who had sustained an acute triple Achilles tendon rupture injury during a game. His triple Achilles tendon rupture consisted of a rupture of the proximal watershed region, a rupture of the main body (mid-watershed area), and an avulsion-type rupture of insertional calcific tendinosis. The patient was treated surgically with primary repair of the tendon, including tenodesis with anchors. Postoperative treatment included non-weightbearing for 4 weeks and protected weightbearing until 10 weeks postoperative, followed by formal physical therapy, which incorporated an "antigravity" treadmill. The patient was able to return to full activity after 26 weeks, including running and refereeing, without limitations. Copyright © 2017 The American College of Foot and Ankle Surgeons. Published by Elsevier Inc. All rights reserved.

  2. MRI of tibialis anterior tendon rupture

    International Nuclear Information System (INIS)

    Gallo, Robert A.; DeMeo, Patrick J.; Kolman, Brett H.; Daffner, Richard H.; Sciulli, Robert L.; Roberts, Catherine C.

    2004-01-01

    Ruptures of the tibialis anterior tendon are rare. We present the clinical histories and MRI findings of three recent male patients with tibialis anterior tendon rupture aged 58-67 years, all of whom presented with pain over the dorsum of the ankle. Two of the three patients presented with complete rupture showing discontinuity of the tendon, thickening of the retracted portion of the tendon, and excess fluid in the tendon sheath. One patient demonstrated a partial tear showing an attenuated tendon with increased surrounding fluid. Although rupture of the tibialis anterior tendon is a rarely reported entity, MRI is a useful modality in the definitive detection and characterization of tibialis anterior tendon ruptures. (orig.)

  3. Ruptured gastroepiploic artery aneurysm: A case report

    Directory of Open Access Journals (Sweden)

    Ahmad S. Ashrafi

    Full Text Available Introduction: Gastroepiploic artery aneurysms are extremely rare, with few reported cases in the literature. The risk of rupture however, is high and thus warrants attention. Presentation of case: Here we present a rare case of a women who presented to the emergency department in shock and was found to have a ruptured gastroepiploic artery aneurysm during surgical exploration. Suture ligation of the aneurysm was completed. Discussion: Although rare, gastroepiploic artery aneurysms have up to a 90% rate of rupture and therefore require intervention. A laparoscopic approach has been described however, in cases where rupture has occurred, urgent laparotomy and control of hemorrhage is needed. Conclusion: We describe a rare case of a ruptured gastroepiploic aneurysm that was successfully managed with urgent laparotomy and aneurysmal resection. Keywords: Gastroepiploic, Aneurysm, Hemorrhage, Case report

  4. Earthquake lights and rupture processes

    Directory of Open Access Journals (Sweden)

    T. V. Losseva

    2005-01-01

    Full Text Available A physical model of earthquake lights is proposed. It is suggested that the magnetic diffusion from the electric and magnetic fields source region is a dominant process, explaining rather high localization of the light flashes. A 3D numerical code allowing to take into account the arbitrary distribution of currents caused by ground motion, conductivity in the ground and at its surface, including the existence of sea water above the epicenter or (and near the ruptured segments of the fault have been developed. Simulations for the 1995 Kobe earthquake were conducted taking into account the existence of sea water with realistic geometry of shores. The results do not contradict the eyewitness reports and scarce measurements of the electric and magnetic fields at large distances from the epicenter.

  5. Traumatic rupture of the aorta

    International Nuclear Information System (INIS)

    Dorfman, G.S.; Paolella, L.P.; Haas, R.A.; Lambiase, R.E.; Cronan, J.J.

    1988-01-01

    To evaluate the acceptability of digital subtraction angiography (DSA) in the evaluation of traumatic rupture of the aorta (TRA), the authors obtained 56 thoracic aortograms in 55 consecutive trauma patients, using both DSA and cut-film angiography (CFA). Both studies were ranked blindly and assigned scores for quality and diagnosis. Interobserver variance for DSA and CFA quality rankings was insignificant. CFA achieved significantly higher ranking for quality. In the 56 examinations, CFA demonstrated seven abnormalities that demanded intervention for follow-up angiography. DSA demonstrated only five of these and found no additional abnormalities. While this evaluation applies only to the particular digital system that the authors tested, they found that DSA, as compared with CFA, failed to demonstrate significant aortic injury. The confidence of diagnosis was significantly greater with CFA. Similar double-blind evaluation is mandatory at any trauma center prior to converting from CFA to DSA in the diagnosis of this life-threatening condition

  6. Rupture of the meniscofibular ligament

    Directory of Open Access Journals (Sweden)

    Poyanli Oguz

    2010-05-01

    Full Text Available Abstract The meniscofibular ligament is an anatomically defined ligament of the knee in humans. However, there are no data regarding the prognosis following injury to this ligament. Our case was a 42-year-old man who presented at our clinic with pain of the lateral side of his left knee. MRI of his left knee revealed the rupture of the meniscofibular ligament. The mechanism of injury was consistent with anatomical and mechanical studies of the meniscofibular ligament. The patient was treated conservatively for 1 year, but his pain did not resolve completely. A case series of patients with the same injury is required to establish an effective treatment for this rare injury.

  7. Steam generator tube degradation at the Doel 4 plant influence on plant operation and safety

    International Nuclear Information System (INIS)

    Scheveneels, G.

    1997-01-01

    The steam generator tubes of Doel 4 are affected by a multitude of corrosion phenomena. Some of them have been very difficult to manage because of their extremely fast evolution, non linear evolution behavior or difficult detectability and/or measurability. The exceptional corrosion behavior of the steam generator tubes has had its drawbacks on plant operation and safety. Extensive inspection and repair campaigns have been necessary and have largely increased outage times and radiation exposure to personnel. Although considerable effort was invested by the utility to control corrosion problems, non anticipated phenomena and/or evolution have jeopardized plant safety. The extensive plugging and repairs performed on the steam generators have necessitated continual review of the design basis safety studies and the adaptation of the protection system setpoints. The large asymmetric plugging has further complicated these reviews. During the years many preventive and recently also defence measures have been implemented by the utility to manage corrosion and to decrease the probability and consequences of single or multiple tube rupture. The present state of the Doel 4 steam generators remains troublesome and further examinations are performed to evaluate if continued operation until June '96, when the steam generators will be replaced, is justified

  8. A current perspective on the risk significance of steam explosions

    International Nuclear Information System (INIS)

    Snyder, A.W.

    1982-01-01

    The view currently held in the Sandia National Laboratory is that, in the case of a meltdown in the reactor core, the probability of a steam explosion is greater than was estimated in WASH-1400, but that the extent and effect of an explosion will be very much smaller than assumed in WASH-1400. This results in a far smaller total risk with regard to containment. In WASH-1400, a nominal conditional probability of 1% was assumed for a containment rupture in a PWR-type reactor, should a large part of the reactor fuel be subject to meltdown during the course of the accident. The German risk analysis study 'Deutsche Risikostudie Kernkraftwerke' dated 1979 considers an explosion of a size sufficient to represent a threat to containment to be considerably more improbable than was assumed in WASH-1400. (orig./DG) [de

  9. Device for achieving pressure balance in the steam generator of a power plant in case of a main-steam pipe or a feedwater pipe break

    International Nuclear Information System (INIS)

    Wietelmann, F.

    1978-01-01

    In order to increase the safety in the steam generator of a power plant in case of a pipe break, the possibility of a pressure balance between the feedwater inlet and the initial steam outlet chambers is allowed for. According to the invention, the partition wall separating these two chambers will exhibit several overflow openings, each of which will be provided with a closure and half of which may be opened to one side only, care having been taken that in case of an accident on occurrence of a certain differential pressure they will always be opened to the low-pressure side. As closures caps, which may be swing out of the way, or rupture diaphragms are mentioned. (UWI) 891 HP [de

  10. Multi-target Wastage Phenomena on Steam Generator Tubes During an SWR Event

    International Nuclear Information System (INIS)

    Jeong, Ji Young; Kim, Jong Man; Kim, Tae Joon; Eoh, Jae Hyuk; Choi, Jong Hyeun; Lee, Yong Bum

    2011-01-01

    The Korean sodium cooled fast reactor, KALIMER- 600 (Korea Advanced LIquid MEtal Reactor) of which the electric output is 600MWe, was developed. The steam generator (SG) of this system is a shell-and-tube type counter-current flow heat exchanger, which is vertically oriented with fixed tube-sheets. A direct heat exchange occurs between the shell-side sodium and the tube-side water at the SG unit. Feed-water enters the inlet nozzle at the lower part of the unit and it flows upward along the helically coiled heat transfer tubes. The inflow sodium is cooled down at the bundle region and then flows out through the sodium outlet nozzle at the bottom of the unit. The typical configuration of the KALIMER-600 SG is shown in Figure 1. In a steam generator, sodium and water are separated by the heat transfer tube wall and it makes a strong pressure boundary between the shell-side sodium and the tube-side water/steam. For this reason, if there is a small hole or crack, even with a pin hole, on heat transfer tubes, a large amount of water/steam would leak into the liquid sodium due to the high pressure difference more than 150 bars, and an exothermic sodium-water chemical reaction takes place as a result. This type of sodium-water reaction (SWR) has been considered as one of the most important safety issues to be resolved. From previous studies, it was obviously figured out that the number of ruptured tubes during an SWR event is one of the most significant factors to determine the temperature and pressure transient. Any subsequent tube rupture behavior in the vicinity of the initially postulated single ruptured tube should be evaluated by considering the single- and multi-target wastage phenomena. Wastage is defined as damage to the structural material (e.g. heat transfer tubes) due to an impingement of the highly corrosive reaction product. Since the impingement may cause wastage of the neighboring heat transfer tubes, a subsequent tube failure can occur in a very short time

  11. Maintaining steam/condensate lines

    International Nuclear Information System (INIS)

    Russum, S.A.

    1992-01-01

    Steam and condensate systems must be maintained with the same diligence as the boiler itself. Unfortunately, they often are not. The water treatment program, critical to keeping the boiler at peak efficiency and optimizing operating life, should not stop with the boiler. The program must encompass the steam and condensate system as well. A properly maintained condensate system maximizes condensate recovery, which is a cost-free energy source. The fuel needed to turn the boiler feedwater into steam has already been provided. Returning the condensate allows a significant portion of that fuel cost to be recouped. Condensate has a high heat content. Condensate is a readily available, economical feedwater source. Properly treated, it is very pure. Condensate improves feedwater quality and reduces makeup water demand and pretreatment costs. Higher quality feedwater means more reliable boiler operation

  12. 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

  13. Water box for steam generator

    International Nuclear Information System (INIS)

    Lecomte, Robert; Viaud, Michel.

    1975-01-01

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

  14. Study on tube rupture strength evaluation method for rapid overheating

    International Nuclear Information System (INIS)

    Komine, Ryuji; Wada, Yusaku

    1998-08-01

    A sodium-water reaction derived from the single tube break in steam generator might overheat neighbor tubes rapidly under internal pressure loadings. If the temperature of tube wall becomes too high, it has to be evaluated that the stress of tube does not exceed the material strength limit to prevent the propagation of tube rupture. In the present study this phenomenon was recognized as the fracture of cylindrical tube with the large deformation due to overheating, and the evaluation method was investigated based on both of experimental and analytical approaches. The results obtained are as follows. (1) As for the nominal stress estimation, it was clarified through the experimental data and the detailed FEM elasto-plastic large deformation analysis that the formula used in conventional designs can be applied. (2) Within the overheating temperature limits of tubes, the creep effect is dominant, even if the loading time is too short. So the strain rate on the basis of JIS elevated temperature tensile test method for steels and heat-resisting alloys is too late and almost of total strain is composed by creep one. As a result the time dependent effect cannot be evaluated under JIS strain rate condition. (3) Creep tests in shorter time condition than a few minutes and tensile tests in higher strain rate condition than 10%/min of JIS are carried out for 2 1/4Cr-1Mo(NT) steel, and the standard values for tube rupture strength evaluation are formulated. (4) The above evaluation method based on both of the stress estimation and the strength standard values application is justified by using the tube burst test data under internal pressure. (5) The strength standard values on Type 321 ss is formulated in accordance with the procedure applied for 2 1/4Cr-1Mo(NT) steel. (author)

  15. 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)

  16. Nuclear reactor steam depressurization valve

    International Nuclear Information System (INIS)

    Moore, G.L.

    1991-01-01

    This patent describes improvement in a nuclear reactor plant, an improved steam depressurization valve positioned intermediate along a steam discharge pipe for controlling the venting of steam pressure from the reactor through the pipe. The improvement comprises: a housing including a domed cover forming a chamber and having a partition plate dividing the chamber into a fluid pressure activation compartment and a steam flow control compartment, the valve housing being provided with an inlet connection and an outlet connection in the steam flow control compartment, and a fluid duct in communication with a source of fluid pressure for operating the valve; a valve set mounted within the fluid flow control compartment comprising a cylindrical section surrounding the inlet connection with one end adjoining the connection and having a radially projecting flange at the other end with a contoured extended valve sealing flange provided with an annular valve sealing member, and a valve cylinder traversing the partition plate and reciprocally movable within an opening in the partition plate with one terminal and extending into the fluid pressure activation compartment and the other terminal end extending into the steam flow control compartment coaxially aligned with the valve seat surrounding the inlet connection, the valve cylinder being surrounded by two bellow fluid seals and provided with guides to inhibit lateral movement, an end of the valve cylinder extending into the fluid flow control compartment having a radially projecting flange substantially conterminous with the valve seat flange and having a contoured surface facing and complimentary to the contoured valve seating surface whereby the two contoured valve surfaces can meet in matching relationship, thus providing a pressure actuated reciprocatable valve member for making closing contact with the valve seat and withdrawing therefrom for opening fluid flow through the valve

  17. Metrics for comparing dynamic earthquake rupture simulations

    Science.gov (United States)

    Barall, Michael; Harris, Ruth A.

    2014-01-01

    Earthquakes are complex events that involve a myriad of interactions among multiple geologic features and processes. One of the tools that is available to assist with their study is computer simulation, particularly dynamic rupture simulation. A dynamic rupture simulation is a numerical model of the physical processes that occur during an earthquake. Starting with the fault geometry, friction constitutive law, initial stress conditions, and assumptions about the condition and response of the near‐fault rocks, a dynamic earthquake rupture simulation calculates the evolution of fault slip and stress over time as part of the elastodynamic numerical solution (Ⓔ see the simulation description in the electronic supplement to this article). The complexity of the computations in a dynamic rupture simulation make it challenging to verify that the computer code is operating as intended, because there are no exact analytic solutions against which these codes’ results can be directly compared. One approach for checking if dynamic rupture computer codes are working satisfactorily is to compare each code’s results with the results of other dynamic rupture codes running the same earthquake simulation benchmark. To perform such a comparison consistently, it is necessary to have quantitative metrics. In this paper, we present a new method for quantitatively comparing the results of dynamic earthquake rupture computer simulation codes.

  18. Hepatic Rupture Induced by Spontaneous Intrahepatic Hematoma

    Directory of Open Access Journals (Sweden)

    Jin-bao Zhou

    2018-01-01

    Full Text Available The etiology of hepatic rupture is usually secondary to trauma, and hepatic rupture induced by spontaneous intrahepatic hematoma is clinically rare. We describe here a 61-year-old female patient who was transferred to our hospital with hepatic rupture induced by spontaneous intrahepatic hematoma. The patient had no history of trauma and had a history of systemic lupus erythematosus for five years, taking a daily dose of 5 mg prednisone for treatment. The patients experienced durative blunt acute right upper abdominal pain one day after satiation, which aggravated in two hours, accompanied by dizziness and sweating. Preoperative diagnosis was rupture of the liver mass. Laparotomy revealed 2500 mL fluid consisting of a mixture of blood and clot in the peritoneal cavity. A 3.5 cm × 2.5 cm rupture was discovered on the hepatic caudate lobe near the vena cava with active arterial bleeding, and a 5  × 6 cm hematoma was reached on the right posterior lobe of the liver. Abdominal computed tomography (CT and laparotomy revealed spontaneous rupture of intrahepatic hematoma with hemorrhagic shock. The patient was successfully managed by suturing the rupture of the hepatic caudate lobe and clearing part of the hematoma. The postoperative course was uneventful, and the patient was discharged after two weeks of hospitalization.

  19. Nuclear steam generator tubesheet shield

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  20. 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)

  1. Double wall steam generator tubing

    International Nuclear Information System (INIS)

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

    1983-01-01

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

  2. Steam reforming of light oxygenates

    DEFF Research Database (Denmark)

    Trane-Restrup, Rasmus; Resasco, Daniel E; Jensen, Anker Degn

    2013-01-01

    Steam reforming (SR) of ethanol, acetic acid, acetone, acetol, 1-propanol, and propanal has been investigated over Ni/MgAl2O4 at temperatures between 400 and 700 degrees C and at a steam-to-carbon-ratio (S/C) of 6. The yield of H-2 and conversion increased with temperature, while the yield of by-...... of CH4. Significant deactivation of the catalyst was observed for all of the compounds and was mainly due to carbon formation. The carbon formation was highest for alcohols due to a high formation of olefins, which are potent coke precursors....

  3. Steam Digest 2001: Office of Industrial Technologies

    Energy Technology Data Exchange (ETDEWEB)

    None, None

    2002-01-01

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

  4. Efficiency of defect specific maintenance od steam generator tubes: the case of ODSCC

    International Nuclear Information System (INIS)

    Cizelj, L.; Dvorsek, T.

    1996-01-01

    The outside diameter stress corrosion cracking at tube support plates became the dominating ageing mechanism in steam generators tubes made of Inconel 600. A variety of maintenance approaches were developed and implemented worldwide to deal with this mechanism. Despite different philosophical and physical backgrounds implemented, all of the applied approaches satisfy the relevant regulatory requirements. For our purpose, the maintenance approach consist of: (1) inspection of tubes, (2) accepting or rejecting the defective tube and (3) plugging of rejected tubes. The problem of selecting an optimal maintenance approach is raised in the paper. Consequently, a method comparing the efficiency of applicable maintenance approaches is proposed. The efficiency is defined by three parameters: (a) number of plugged tubes, (b) probability of steam generator tube rupture and (c) predicted accidental leak rates through the defects. An original probabilistic model is proposed to quantify the probability of tube rupture, while procedures available in literature were used to define the accidental leak rates. The numerical example considers the data from Krsko NPP (Westinghouse 632 MWe). The maintenance approaches analyzed include: (i) no repair at all, (ii) traditional defect depth (40%) based maintenance, (iii) alternate plugging criterion (bobbin coil voltage as defined by EPRI and U.S. NRC) and (iv) combined traditional and alternate approach. Advantages of the defect specific approaches (iii) and (iv) over the traditional one (defect depth) are clearly shown. A brief discussion on the optimization of safe life of steam generator is given. (author)

  5. Misdiagnosed Chest Pain: Spontaneous Esophageal Rupture

    Science.gov (United States)

    Inci, Sinan; Gundogdu, Fuat; Gungor, Hasan; Arslan, Sakir; Turkyilmaz, Atila; Eroglu, Atila

    2013-01-01

    Chest pain is one of themost common complaints expressed by patients presenting to the emergency department, and any initial evaluation should always consider life-threatening causes. Esophageal rupture is a serious condition with a highmortality rate. If diagnosed, successful therapy depends on the size of the rupture and the time elapsed between rupture and diagnosis.We report on a 41-year-old woman who presented to the emergency department complaining of left-sided chest pain for two hours. PMID:27122690

  6. Eccentric pressurized tube for measuring creep rupture

    International Nuclear Information System (INIS)

    Schwab, P.R.

    1981-01-01

    Creep rupture is a long term failure mode in structural materials that occurs at high temperatures and moderate stress levels. The deterioration of the material preceding rupture, termed creep damage, manifests itself in the formation of small cavities on grain boundaries. To measure creep damage, sometimes uniaxial tests are performed, sometimes density measurements are made, and sometimes the grain boundary cavities are measured by microscopy techniques. The purpose of the present research is to explore a new method of measuring creep rupture, which involves measuring the curvature of eccentric pressurized tubes. Theoretical investigations as well as the design, construction, and operation of an experimental apparatus are included in this research

  7. Spontaneous distal rupture of the plantar fascia.

    Science.gov (United States)

    Gitto, Salvatore; Draghi, Ferdinando

    2018-07-01

    Spontaneous ruptures of the plantar fascia are uncommon injuries. They typically occur at its calcaneal insertion and usually represent a complication of plantar fasciitis and local treatment with steroid injections. In contrast, distal ruptures commonly result from traumatic injuries. We describe the case of a spontaneous distal rupture of the plantar fascia in a 48-year-old woman with a low level of physical activity and no history of direct injury to the foot, plantar fasciitis, or steroid injections. © 2017 Wiley Periodicals, Inc.

  8. 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

  9. 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

  10. First application of a partially automated road header at Prosper-Haniel colliery; Ersteinsatz einer teilautomatisierten Teilschnittmaschine auf dem Bergwerk Prosper-Haniel

    Energy Technology Data Exchange (ETDEWEB)

    Reinewardt, Klaus-Juergen [Bergwerk Prosper-Haniel, Bottrop (Germany). Betrieb Produktion; Achilles, Peter [RAG Deutsche Steinkohle AG, Herne (Germany). Abt. PPE-V Vorleistungstechnik

    2010-09-15

    Mechanical road heading in the RAG Deutsche Steinkohle mines makes use of AM 105 road headers. Within the scope of an EU-subsidised R and D project the machine has been subjected to an automation of its control features and an integration of sensors for seam position identification and for navigation. The focal points of the automation are: - the scheduled performance of automated cutting operations, - the adherence to a defined loading height due to seam position identification and - the incorporation of first auxiliary functions for navigation (position sensing). In addition, the machine is expected to determine its respective functional state and recognised potential functional faults by itself, and - subject to its present load condition - it is to deliver the basis for a maintenance scheme geared to its current condition. This paper describes above-mentioned development steps and reports on the experience gathered in the underground use of that machine in the Prosper-Haniel colliery. (orig.)

  11. 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

  12. 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

  13. Parametric study for horizontal steam generator modelling

    Energy Technology Data Exchange (ETDEWEB)

    Ovtcharova, I. [Energoproekt, Sofia (Bulgaria)

    1995-12-31

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

  14. 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.

  15. Parametric study for horizontal steam generator modelling

    Energy Technology Data Exchange (ETDEWEB)

    Ovtcharova, I [Energoproekt, Sofia (Bulgaria)

    1996-12-31

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

  16. 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.)

  17. BWR Steam Dryer Alternating Stress Assessment Procedures

    Energy Technology Data Exchange (ETDEWEB)

    Morante, R. J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Hambric, S. A. [Brookhaven National Lab. (BNL), Upton, NY (United States); Ziada, S. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2016-12-01

    This report presents an overview of Boiling Water Reactor (BWR) steam dryer design; the fatigue cracking failures that occurred at the Quad Cities (QC) plants and their root causes; a history of BWR Extended Power Uprates (EPUs) in the USA; and a discussion of steam dryer modifications/replacements, alternating stress mechanisms on steam dryers, and structural integrity evaluations (static and alternating stress).

  18. Real-Time Detection of Rupture Development: Earthquake Early Warning Using P Waves From Growing Ruptures

    Science.gov (United States)

    Kodera, Yuki

    2018-01-01

    Large earthquakes with long rupture durations emit P wave energy throughout the rupture period. Incorporating late-onset P waves into earthquake early warning (EEW) algorithms could contribute to robust predictions of strong ground motion. Here I describe a technique to detect in real time P waves from growing ruptures to improve the timeliness of an EEW algorithm based on seismic wavefield estimation. The proposed P wave detector, which employs a simple polarization analysis, successfully detected P waves from strong motion generation areas of the 2011 Mw 9.0 Tohoku-oki earthquake rupture. An analysis using 23 large (M ≥ 7) events from Japan confirmed that seismic intensity predictions based on the P wave detector significantly increased lead times without appreciably decreasing the prediction accuracy. P waves from growing ruptures, being one of the fastest carriers of information on ongoing rupture development, have the potential to improve the performance of EEW systems.

  19. Analysis of independent failure assumptions on postulated secondary high energy line ruptures

    International Nuclear Information System (INIS)

    Hollingsworth, S.D.

    1977-01-01

    Postulated ruptures of the main steam piping in pressurized water reactors result in large amounts of steam being removed from the secondary system. Since the energy removal rate could be many times that of nominal design power, there may be a rapid cooldown of the primary coolant system and a positive addition of reactivity to the reactor core. The Westinghouse protection system design concept incorporates features that trip the reactor, isolate the main steamlines and provide for automatic alternate shutdown capability in the form of boric acid solution injection into the primary coolant system. At the most limiting time in life (end of life) the reactivity calculated to be inserted by the cooldown is sufficient to overcome the shutdown margin predicted to be available from control rods with the most reactive rod in the fully withdrawn position. Because the boron injected into the core may be delayed due to system responses, there is potential that the reactor core could return critical and return to power. The extremely adverse radial power distributions caused by the fully withdrawn control rod causes localized high power densities that could lead to reduced heat transfer capability (DNB). Because of the large amount of stored energy in the reactor coolant system at full power, the cooldown and subsequent return to power is more severe when calculated from a shutdown, hot zero power condition. It is shown that the protection system design has large margins to protect against adverse core effects following a steamline rupture

  20. Spontaneous rupture of ovarian cystadenocarcinoma: pre- and post-rupture computed tomography evaluation

    Energy Technology Data Exchange (ETDEWEB)

    Salvadori, Priscila Silveira; Atzingen, Augusto Castelli von; D' Ippolito, Giuseppe [Universidade Federal de Sao Paulo (EPM/UNIFESP), Sao Paulo, SP (Brazil). Escola Paulista de Medicina; Bomfim, Lucas Novais [Universidade Tiradentes (UNIT), Maceio, AL, (Brazil)

    2015-09-15

    Epithelial ovarian tumors are the most common malignant ovarian neoplasms and, in most cases, eventual rupture of such tumors is associated with a surgical procedure. The authors report the case of a 54-year-old woman who presented with spontaneous rupture of ovarian cystadenocarcinoma documented by computed tomography, both before and after the event. In such cases, a post-rupture staging tends to be less favorable, compromising the prognosis. (author)

  1. Global catalog of earthquake rupture velocities shows anticorrelation between stress drop and rupture velocity

    Science.gov (United States)

    Chounet, Agnès; Vallée, Martin; Causse, Mathieu; Courboulex, Françoise

    2018-05-01

    Application of the SCARDEC method provides the apparent source time functions together with seismic moment, depth, and focal mechanism, for most of the recent earthquakes with magnitude larger than 5.6-6. Using this large dataset, we have developed a method to systematically invert for the rupture direction and average rupture velocity Vr, when unilateral rupture propagation dominates. The approach is applied to all the shallow (z earthquakes of the catalog over the 1992-2015 time period. After a careful validation process, rupture properties for a catalog of 96 earthquakes are obtained. The subsequent analysis of this catalog provides several insights about the seismic rupture process. We first report that up-dip ruptures are more abundant than down-dip ruptures for shallow subduction interface earthquakes, which can be understood as a consequence of the material contrast between the slab and the overriding crust. Rupture velocities, which are searched without any a-priori up to the maximal P wave velocity (6000-8000 m/s), are found between 1200 m/s and 4500 m/s. This observation indicates that no earthquakes propagate over long distances with rupture velocity approaching the P wave velocity. Among the 23 ruptures faster than 3100 m/s, we observe both documented supershear ruptures (e.g. the 2001 Kunlun earthquake), and undocumented ruptures that very likely include a supershear phase. We also find that the correlation of Vr with the source duration scaled to the seismic moment (Ts) is very weak. This directly implies that both Ts and Vr are anticorrelated with the stress drop Δσ. This result has implications for the assessment of the peak ground acceleration (PGA) variability. As shown by Causse and Song (2015), an anticorrelation between Δσ and Vr significantly reduces the predicted PGA variability, and brings it closer to the observed variability.

  2. Materials Performance in USC Steam

    Energy Technology Data Exchange (ETDEWEB)

    G. R. Holcomb, P. Wang, P. D. Jablonski, and J. A. Hawk

    2010-05-01

    The proposed steam inlet temperature in the Advanced Ultra Supercritical (A-USC) steam turbine is high enough (760 °C) that traditional turbine casing and valve body materials such as ferritic/martensitic steels will not suffice due to temperature limitations of this class of materials. Cast versions of several traditionally wrought Ni-based superalloys were evaluated for use as casing or valve components for the next generation of industrial steam turbines. The full size castings are substantial: 2-5,000 kg each half and on the order of 100 cm thick. Experimental castings were quite a bit smaller, but section size was retained and cooling rate controlled to produce equivalent microstructures. A multi-step homogenization heat treatment was developed to better deploy the alloy constituents. The most successful of these cast alloys in terms of creep strength (Haynes 263, Haynes 282, and Nimonic 105) were subsequently evaluated by characterizing their microstructure as well as their steam oxidation resistance (at 760 and 800 °C).

  3. The STEAM behind the Scenes

    Science.gov (United States)

    Smith, Carmen Petrick; King, Barbara; González, Diana

    2015-01-01

    There is a growing need for STEAM-based (Science, Technology, Engineering, Arts, and Mathematics) knowledge and skills across a wide range of professions (Brazell 2013). Yet students often fail to see the usefulness of mathematics beyond the classroom (Kloosterman, Raymond, and Emenaker 1996), and they do not regularly make connections between…

  4. 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)

  5. Severity parameters for steam cracking

    NARCIS (Netherlands)

    Golombok, M.; Bijl, J.L.M.; Kornegoor, M.

    2001-01-01

    There are several ways to measure severity in steam cracking which are all a function of residence time, temperature, and pressure. Many measures of severity are not practicable for experimental purposes. Our experimental study shows that methane make is the best measure of severity because it is an

  6. Nuclear process steam for industry

    International Nuclear Information System (INIS)

    Seddon, W.A.

    1981-11-01

    A joint industrial survey funded by the Bruce County Council, the Ontario Energy Corporation and Atomic Energy of Canada Limited was carried out with the cooperation of Ontario Hydro and the Ontario Ministry of Industry and Tourism. Its objective was to identify and assess the future needs and interest of energy-intensive industries in an Industrial Energy Park adjacent to the Bruce Nuclear Power Development. The Energy Park would capitalize on the infrastructure of the existing CANDU reactors and Ontario Hydro's proven and unique capability to produce steam, as well as electricity, at a cost currently about half that from a comparable coal-fired station. Four industries with an integrated steam demand of some 1 x 10 6 lb/h were found to be prepared to consider seriously the use of nuclear steam. Their combined plants would involve a capital investment of over $200 million and provide jobs for 350-400 people. The high costs of transportation and the lack of docking facilities were considered to be the major drawbacks of the Bruce location. An indication of steam prices would be required for an over-all economic assessment

  7. Steam initiated hydrotalcite conversion coatings

    DEFF Research Database (Denmark)

    Zhou, Lingli; Friis, Henrik; Roefzaad, Melanie

    2018-01-01

    A facile process of exploiting high-temperature steam to deposit nvironmentally friendly hydrotalcite (HT) coatings on Al alloy 6060 was developed in a spray system. Scanning electron microscopy showed the formationf a continuous and conformal coating comprised of a compact mass of crystallites. ...

  8. 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.

  9. Acute Iliac Artery Rupture: Endovascular Treatment

    International Nuclear Information System (INIS)

    Chatziioannou, A.; Mourikis, D.; Katsimilis, J.; Skiadas, V.; Koutoulidis, V.; Katsenis, K.; Vlahos, L.

    2007-01-01

    The authors present 7 patients who suffered iliac artery rupture over a 2 year period. In 5 patients, the rupture was iatrogenic: 4 cases were secondary to balloon angioplasty for iliac artery stenosis and 1 occurred during coronary angioplasty. In the last 2 patients, the rupture was secondary to iliac artery mycotic aneurysm. Direct placement of a stent-graft was performed in all cases, which was dilated until extravasation was controlled. Placement of the stent-graft was successful in all the cases, without any complications. The techniques used, results, and mid-term follow-up are presented. In conclusion, endovascular placement of a stent-graft is a quick, minimally invasive, efficient, and safe method for emergency treatment of acute iliac artery rupture, with satisfactory short- and mid-term results

  10. Arthrographic diagnosis of ruptured calcaneofibular ligament. I

    International Nuclear Information System (INIS)

    Vuust, M.

    1980-01-01

    A new projection, oblique axial, is recommended for the arthrography of the acute sprained ankle for the correct diagnosis of a ruptured calcaneofibular ligament. Its value is experimentally confirmed. (Auth.)

  11. Right Diaphragm Spontaneous Rupture: A Surgical Approach

    Directory of Open Access Journals (Sweden)

    Duilio Divisi

    2011-01-01

    Full Text Available We present a case of spontaneous rupture of the diaphragm, characterized by nonspecific symptoms. The rapid diagnosis and appropriate surgical approach led to a positive resolution of the pathology.

  12. Traumatic Gallbladder Rupture Treated by Laparoscopic Cholecystectomy

    Science.gov (United States)

    Egawa, Noriyuki; Ueda, Junji; Hiraki, Masatsugu; Ide, Takao; Inoue, Satoshi; Sakamoto, Yuichiro; Noshiro, Hirokazu

    2016-01-01

    Abstract Gallbladder rupture due to blunt abdominal injury is rare. There are few reports of traumatic gallbladder injury, and it is commonly associated with other concomitant visceral injuries. Therefore, it is difficult to diagnose traumatic gallbladder rupture preoperatively when it is caused by blunt abdominal injury. We report a patient who underwent laparoscopic cholecystectomy after an exact preoperative diagnosis of traumatic gallbladder rupture. A 43-year-old man was admitted to our hospital due to blunt abdominal trauma. The day after admission, abdominal pain and ascites increased and a muscular defense sign appeared. Percutaneous drainage of the ascites was performed, and the aspirated fluid was bloody and almost pure bile. He was diagnosed with gallbladder rupture by the cholangiography using the endoscopic retrograde cholangiopancreatography technique. Laparoscopic cholecystectomy was performed safely, and he promptly recovered. If accumulated fluids contain bile, endoscopic cholangiography is useful not only to diagnose gallbladder injury but also to determine the therapeutic strategy. PMID:27462188

  13. Traumatic Gallbladder Rupture Treated by Laparoscopic Cholecystectomy

    Directory of Open Access Journals (Sweden)

    Noriyuki Egawa

    2016-05-01

    Full Text Available Gallbladder rupture due to blunt abdominal injury is rare. There are few reports of traumatic gallbladder injury, and it is commonly associated with other concomitant visceral injuries. Therefore, it is difficult to diagnose traumatic gallbladder rupture preoperatively when it is caused by blunt abdominal injury. We report a patient who underwent laparoscopic cholecystectomy after an exact preoperative diagnosis of traumatic gallbladder rupture. A 43-year-old man was admitted to our hospital due to blunt abdominal trauma. The day after admission, abdominal pain and ascites increased and a muscular defense sign appeared. Percutaneous drainage of the ascites was performed, and the aspirated fluid was bloody and almost pure bile. He was diagnosed with gallbladder rupture by the cholangiography using the endoscopic retrograde cholangiopancreatography technique. Laparoscopic cholecystectomy was performed safely, and he promptly recovered. If accumulated fluids contain bile, endoscopic cholangiography is useful not only to diagnose gallbladder injury but also to determine the therapeutic strategy.

  14. Cognitive frames in psychology: demarcations and ruptures.

    Science.gov (United States)

    Yurevich, Andrey V

    2009-06-01

    As there seems to be a recurrent feeling of crisis in psychology, its present state is analyzed in this article. The author believes that in addition to the traditional manifestations that have dogged psychology since it emerged as an independent science some new features of the crisis have emerged. Three fundamental "ruptures" are identified: the "horizontal" rupture between various schools and trends, the "vertical" rupture between natural science and humanitarian psychology, and the "diagonal" rupture between academic research and applied practice of psychology. These manifestations of the crisis of psychology have recently been compounded by the crisis of its rationalistic foundations. This situation is described in terms of the cognitive systems in psychology which include meta-theories, paradigms, sociodigms and metadigms.

  15. Analyses of steam generator collector rupture for WWER-1000 using Relap5 code

    Energy Technology Data Exchange (ETDEWEB)

    Balabanov, E.; Ivanova, A. [Energoproekt, Sofia (Bulgaria)

    1995-12-31

    The paper presents some of the results of analyses of an accident with a LOCA from the primary to the secondary side of a WWER-1000/320 unit. The objective of the analyses is to estimate the primary coolant to the atmosphere, to point out the necessity of a well defined operator strategy for this type of accident as well as to evaluate the possibility to diagnose the accident and to minimize the radiological impact on the environment.

  16. Analysis of reactor installation accident WWER-1000 caused by collector head rupture of steam generator

    International Nuclear Information System (INIS)

    Sviridenko, I.I.

    2005-01-01

    System classification of passive heat defence of nuclear dangerous objects with usage of heat pipes. The article is continuation of series of works devoted to research and development of promising passive cooling system of nuclear dangerous objects which use evaporating-condensing heat transfer devices of closed type - heat pipes and two-phase thermosyphons. Classification of autonomous systems of passive heat defence with usage low temperature heat tubes is given. Base classificational lectures are observed. Advantages n disadvantages of various circuit (scheme) solutions are analyzed

  17. Analyses of steam generator collector rupture for WWER-1000 using Relap5 code

    Energy Technology Data Exchange (ETDEWEB)

    Balabanov, E; Ivanova, A [Energoproekt, Sofia (Bulgaria)

    1996-12-31

    The paper presents some of the results of analyses of an accident with a LOCA from the primary to the secondary side of a WWER-1000/320 unit. The objective of the analyses is to estimate the primary coolant to the atmosphere, to point out the necessity of a well defined operator strategy for this type of accident as well as to evaluate the possibility to diagnose the accident and to minimize the radiological impact on the environment.

  18. Analyses of steam generator collector rupture for WWER-1000 using Relap5 code

    International Nuclear Information System (INIS)

    Balabanov, E.; Ivanova, A.

    1995-01-01

    The paper presents some of the results of analyses of an accident with a LOCA from the primary to the secondary side of a WWER-1000/320 unit. The objective of the analyses is to estimate the primary coolant to the atmosphere, to point out the necessity of a well defined operator strategy for this type of accident as well as to evaluate the possibility to diagnose the accident and to minimize the radiological impact on the environment

  19. Assessment on Event Classification of One Steam Generator Tube Rupture in EU-APR

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ji Hwan; Kim, Yong Soo [KHNP CRI, Daejeon (Korea, Republic of)

    2016-10-15

    The Fukushima Daiichi nuclear power plant accident showed the vulnerability of coping strategy to beyond design natural disaster such as beyond design earthquake and tsunami. In Korea, the government and industry performed comprehensive safety inspection on all domestic nuclear power plants against beyond design basis external events and fifty action items have been issued. In addition to post- Fukushima action items, the stress tests for all domestic nuclear power plants are on the way to enhance the safety of domestic nuclear power plants through finding the vulnerabilities in intentional stress conditions initiated by beyond design natural disaster. Recently, the stress tests for WOLSONG Unit 1 and KORI Unit 1 have been performed and their assessment results have been reviewed by Korean regulatory body. The assessment of the coping capability of KORI Unit 1 has been performed under simultaneous the extended loss of AC power and loss of ultimate heat sink initiated by beyond design natural disaster. It is concluded that KORI Unit 1 has the capability, in the event of loss of safety functions by beyond design natural disaster, to sufficiently cool down the reactor core without fuel damage, to keep pressure boundaries of the reactor coolant system in transient condition and to control containment and temperature to maintain the integrity of the containment buildings. The several additional items for safety improvement has been drawn to enhance the coping capability for loss of safety functions under beyond design natural disaster in addition to post Fukushima action items.

  20. Assessment on Event Classification of One Steam Generator Tube Rupture in EU-APR

    International Nuclear Information System (INIS)

    Kim, Ji Hwan; Kim, Yong Soo

    2016-01-01

    The Fukushima Daiichi nuclear power plant accident showed the vulnerability of coping strategy to beyond design natural disaster such as beyond design earthquake and tsunami. In Korea, the government and industry performed comprehensive safety inspection on all domestic nuclear power plants against beyond design basis external events and fifty action items have been issued. In addition to post- Fukushima action items, the stress tests for all domestic nuclear power plants are on the way to enhance the safety of domestic nuclear power plants through finding the vulnerabilities in intentional stress conditions initiated by beyond design natural disaster. Recently, the stress tests for WOLSONG Unit 1 and KORI Unit 1 have been performed and their assessment results have been reviewed by Korean regulatory body. The assessment of the coping capability of KORI Unit 1 has been performed under simultaneous the extended loss of AC power and loss of ultimate heat sink initiated by beyond design natural disaster. It is concluded that KORI Unit 1 has the capability, in the event of loss of safety functions by beyond design natural disaster, to sufficiently cool down the reactor core without fuel damage, to keep pressure boundaries of the reactor coolant system in transient condition and to control containment and temperature to maintain the integrity of the containment buildings. The several additional items for safety improvement has been drawn to enhance the coping capability for loss of safety functions under beyond design natural disaster in addition to post Fukushima action items

  1. Aortic ruptures in seat belt wearers.

    Science.gov (United States)

    Arajärvi, E; Santavirta, S; Tolonen, J

    1989-09-01

    Several investigations have indicated that rupture of the thoracic aorta is one of the leading causes of immediate death in victims of road traffic accidents. In Finland in 1983, 92% of front-seat passengers were seat belt wearers on highways and 82% in build-up areas. The mechanisms of rupture of the aorta have been intensively investigated, but the relationship between seat belt wearing and injury mechanisms leading to aortic rupture is still largely unknown. This study comprises 4169 fatally injured victims investigated by the Boards of Traffic Accident Investigation of Insurance Companies during the period 1972 to 1985. Chest injuries were recorded as the main cause of death in 1121 (26.9%) victims, 207 (5.0%) of those victims having worn a seat belt. Aortic ruptures were found at autopsy in 98 victims and the exact information of the location of the aortic tears was available in 68. For a control group, we analyzed 72 randomly chosen unbelted victims who had a fatal aortic rupture in similar accidents. The location of the aortic rupture in unbelted victims was more often in the ascending aorta, especially in drivers, whereas in seat belt wearers the distal descending aorta was statistically more often ruptured, especially in right-front passengers (p less than 0.05). The steering wheel predominated statistically as the part of the car estimated to have caused the injury in unbelted victims (37/72), and some interior part of the car was the most common cause of fatal thoracic impacts in seat belt wearers (48/68) (p less than 0.001). The mechanism of rupture of the aorta in the classic site just distal to the subclavian artery seems to be rapid deceleration, although complex body movements are also responsible in side impact collisions. The main mechanism leading to rupture of the ascending aorta seems to be severe blow to the bony thorax. This also often causes associated thoracic injuries, such as heart rupture and sternal fracture. Injuries in the ascending

  2. Traumatic Fundal Rupture of unscarred Uterus in a Primigravida ...

    African Journals Online (AJOL)

    Background: Uterine rupture is an infrequent but life threatening obstetric emergency. Rupture of previously scarred uterus is often encountered especially in multiparous women, but the traumatic rupture of an unscarred primigravid uterus as presented here is a relatively rare event. We report a case of rupture of an ...

  3. Describing Soils: Calibration Tool for Teaching Soil Rupture Resistance

    Science.gov (United States)

    Seybold, C. A.; Harms, D. S.; Grossman, R. B.

    2009-01-01

    Rupture resistance is a measure of the strength of a soil to withstand an applied stress or resist deformation. In soil survey, during routine soil descriptions, rupture resistance is described for each horizon or layer in the soil profile. The lower portion of the rupture resistance classes are assigned based on rupture between thumb and…

  4. Micromechanical properties and collagen composition of ruptured human achilles tendon

    DEFF Research Database (Denmark)

    Hansen, Philip; Kovanen, Vuokko; Hölmich, Per

    2013-01-01

    The Achilles tendon is one of the strongest tendons in the human body, and yet it frequently ruptures, which is a substantial clinical problem. However, the cause of ruptures remains elusive.......The Achilles tendon is one of the strongest tendons in the human body, and yet it frequently ruptures, which is a substantial clinical problem. However, the cause of ruptures remains elusive....

  5. Bladder And Uterine Rupture At Vaginal Birth After Cesarean Section

    African Journals Online (AJOL)

    Rupture of the uterus is still a major obstetric problem in developing countries. The urinary bladder is often affected especially when a lower segment Caesarean scar ruptures. We present a case of bladder rupture associated with uterine rupture in a patient who was attempting vaginal delivery after two previous Caesarean ...

  6. Biomechanical rupture risk assessment of abdominal aortic aneurysms based on a novel probabilistic rupture risk index.

    Science.gov (United States)

    Polzer, Stanislav; Gasser, T Christian

    2015-12-06

    A rupture risk assessment is critical to the clinical treatment of abdominal aortic aneurysm (AAA) patients. The biomechanical AAA rupture risk assessment quantitatively integrates many known AAA rupture risk factors but the variability of risk predictions due to model input uncertainties remains a challenging limitation. This study derives a probabilistic rupture risk index (PRRI). Specifically, the uncertainties in AAA wall thickness and wall strength were considered, and wall stress was predicted with a state-of-the-art deterministic biomechanical model. The discriminative power of PRRI was tested in a diameter-matched cohort of ruptured (n = 7) and intact (n = 7) AAAs and compared to alternative risk assessment methods. Computed PRRI at 1.5 mean arterial pressure was significantly (p = 0.041) higher in ruptured AAAs (20.21(s.d. 14.15%)) than in intact AAAs (3.71(s.d. 5.77)%). PRRI showed a high sensitivity and specificity (discriminative power of 0.837) to discriminate between ruptured and intact AAA cases. The underlying statistical representation of stochastic data of wall thickness, wall strength and peak wall stress had only negligible effects on PRRI computations. Uncertainties in AAA wall stress predictions, the wide range of reported wall strength and the stochastic nature of failure motivate a probabilistic rupture risk assessment. Advanced AAA biomechanical modelling paired with a probabilistic rupture index definition as known from engineering risk assessment seems to be superior to a purely deterministic approach. © 2015 The Author(s).

  7. Blunt cardiac rupture in a toddler

    Directory of Open Access Journals (Sweden)

    Peep Talving

    2016-08-01

    Full Text Available Blunt cardiac rupture is typically a fatal injury with overall mortality exceeding 90%. Most of the patients never reach the hospital alive. In pediatric patients, only 0.03% of cases following blunt trauma admissions have a cardiac injury. This report presents a rare survivor of 16-months old toddler injured in a domestic accident suffering a right atrial rupture repaired through a median sternotomy. To the best of our knowledge this is the youngest case reported in the literature.

  8. Treatment strategy for ruptured abdominal aortic aneurysms.

    Science.gov (United States)

    Davidovic, L

    2014-07-01

    Rupture is the most serious and lethal complication of the abdominal aortic aneurysm. Despite all improvements during the past 50 years, ruptured abdominal aortic aneurysms are still associated with very high mortality. Namely, including patients who die before reaching the hospital, the mortality rate due to abdominal aortic aneurysm rupture is 90%. On the other hand, during the last twenty years, the number of abdominal aortic aneurysms significantly increased. One of the reasons is the fact that in majority of countries the general population is older nowadays. Due to this, the number of degenerative AAA is increasing. This is also the case for patients with abdominal aortic aneurysm rupture. Age must not be the reason of a treatment refusal. Optimal therapeutic option ought to be found. The following article is based on literature analysis including current guidelines but also on my Clinics significant experience. Furthermore, this article show cases options for vascular medicine in undeveloped countries that can not apply endovascular procedures at a sufficient level and to a sufficient extent. At this moment the following is evident. Thirty-day-mortality after repair of ruptured abdominal aortic aneurysms is significantly lower in high-volume hospitals. Due to different reasons all ruptured abdominal aortic aneurysms are not suitable for EVAR. Open repair of ruptured abdominal aortic aneurysm should be performed by experienced open vascular surgeons. This could also be said for the treatment of endovascular complications that require open surgical conversion. There is no ideal procedure for the treatment of AAA. Each has its own advantages and disadvantages, its own limits and complications, as well as indications and contraindications. Future reductions in mortality of ruptured abdominal aortic aneurysms will depend on implementation of population-based screening; on strategies to prevent postoperative organ injury and also on new medical technology

  9. Endovascular treatment of ruptured splenic artery aneurysm

    DEFF Research Database (Denmark)

    Bjerring, Ole Steen

    2008-01-01

    Splenic artery aneurysms (SAA) are traditionally treated surgically, but endovascular techniques are becoming increasingly popular. A 64 year-old male with chest pain and low blood pressure was admitted under suspicion of AMI. A CT scan showed a 56 mm SAA with signs of rupture. The patient...... was treated with endovascular embolisation of the SAA with coils. Blood pressure and haemoglobin levels were stabilized and the patient was discharged. In the case of rupture the treatment of choice seems to be endovascular....

  10. Splenic artery aneurysm rupture in pregnancy

    International Nuclear Information System (INIS)

    Khan, H.R.; Low, S.; Selinger, M.; Nelson, N.

    2004-01-01

    Rupture of a splenic artery aneurysm, commonly associated with pregnancy is a rare and catastrophic event. We report here a case of a patient in her second pregnancy who presented with a short history of left hypochondriac and epigastric pain, followed by collapse at 32 weeks gestation. Sudden fetal distress lead to emergency caesarean delivery when splenic artery aneurysm rupture was diagnosed. With timely involvement of multidisciplinary personnel both mother and baby survived and had an uneventful recovery. (author)

  11. Plantar Fascia Rupture: Ultrasound to Facilitate Recognition.

    Science.gov (United States)

    Servey, Jessica T; Jonas, Christopher

    2018-01-01

    Plantar fascia rupture in the absence of previous diagnosis of plantar fasciitis, corticosteroid injection, or injury is a rare occurrence with only 7 case reports in the literature since 1978. This is a case of spontaneous plantar fascia rupture in a 38-year-old active-duty US military member with current considerations in musculoskeletal ultrasound, other radiologic imaging, treatment, and followup of this diagnosis. © Copyright 2018 by the American Board of Family Medicine.

  12. MRI findings of achilles tendon rupture

    International Nuclear Information System (INIS)

    Zhang Xuezhe

    2009-01-01

    Objective: To evaluate the MRI findings of achilles tendon rupture. Methods: The MRI data of 7 patients with achilles tendon rupture were retrospectively analysed. All 7 patients were male with the age ranging from 34 to 71 years. Routine MR scanning was performed in axial and sagittal planes, including T 1 WI, T 2 WI and a fat suppression MRI (SPIR). Results: Among 7 patients, complete achilles tendon rupture was seen in 6 cases, partial achilles tendon rupture 1 case. The site of tendon disruption were 2.6-11.0 cm( mean 5.4 cm) proximal to the insertion in the calcaneus. The MRI findings of a partial or complete rupture of the achilles tendon included enlarged and thickened achilles tendon (7 cases), wavy lax achilles tendon (2 cases), discontinuity of some or all of its fibers and intratendinous regions of increased signal intensity (7 cases). In the cases of complete tendon rupture, the size of the tendinous gap varied from 3.0-8.0 mm, which was filled with blood and appeared as edema of increase signal intensity on T 2 WI and SPIR. In all 7 patients, MR scanning showed medium signal intensity (7 cases) on T 1 WI, or medium signal intensity (1 cases), medium-high signal intensity (3 cases ), high signal intensity (3 cases) on T 2 WI, and medium-high signal intensity (2 cases), high signal intensity (5 cases) on fat suppression MRI. The preachilles fat pad showed obscure in 6 cases of complete achilles tendon rupture. Conclusion: MRI is an excellent method for revealing achilles tendon rupture and confirming the diagnosis. (authors)

  13. Endovascular therapeutic strategies in ruptured intracranial aneurysms

    International Nuclear Information System (INIS)

    Machi, Paolo; Lobotesis, Kyriakos; Vendrell, Jean Francoise; Riquelme, Carlos; Eker, Omer; Costalat, Vincent; Bonafe, Alain

    2013-01-01

    The aim of the present study was to evaluate endovascular techniques used currently which were not available at the time of ISAT inclusion period, such as balloon remodelling and flow-divertion, in order to assess whether these new technologies have improved the endovascular approach outcomes. We present a review of articles, published in major journals, with the aim to evaluate the efficacy and the safety of coiling with balloon remodelling for the treatment of ruptured aneurysms in comparison to coiling performed without such coadjutant techniques. Furthermore, we reviewed publications reporting on the treatment of ruptured aneurysms in the acute phase with the one of the most recent technologies available nowadays: the flow diverting stent. Looking at the recent literature the results regarding ruptured aneurysms treated with balloon assisted coiling (BAC) have shown an improvement in terms of anatomical results and morbi-mortality rates. Case series of ruptured middle cerebral artery (MCA) aneurysms treated by EVT report results similar to those obtained by surgical clipping. Several articles recently report encouraging results in treating ruptured dissecting and blister aneurysms with flow diverters. Questions regarding the best treatment available for ruptured aneurysms are yet to be answered. Hence there is a need for a subsequent trial aiming to answer these unresolved issues

  14. 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.)

  15. 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

  16. 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

  17. 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.

  18. Modeling of Spray System Operation under Hydrogen and Steam Emissions in NPP Containment during Severe Accident

    Directory of Open Access Journals (Sweden)

    Vadim E. Seleznev

    2011-01-01

    Full Text Available The paper describes one of the variants of mathematical models of a fluid dynamics process inside the containment, which occurs in the conditions of operation of spray systems in severe accidents at nuclear power plant. The source of emergency emissions in this case is the leak of the coolant or rupture at full cross-section of the main circulating pipeline in a reactor building. Leak or rupture characteristics define the localization and the temporal law of functioning of a source of emergency emission (or accrued operating of warmed up hydrogen and steam in the containment. Operation of this source at the course of analyzed accident models should be described by the assignment of the relevant Dirichlet boundary conditions. Functioning of the passive autocatalytic recombiners of hydrogen is described in the form of the complex Newton boundary conditions.

  19. RESEARCH ON REDUCING PREMATURITY RUPTURE OF MEMBRANE

    Directory of Open Access Journals (Sweden)

    Maria URSACHI (BOLOTA

    2016-12-01

    Full Text Available The membranes surrounding the amniotic cavity are composed from amnion and chorion, tightly adherent layers which are composed of several cell types, including epithelial cells, trophoblasts cells and mesenchyme cells, embedded in a collagenous matrix. They retain amniotic fluid, secret substances into the amniotic fluid, as well as to the uterus and protect the fetus against upward infections from urogenital tract. Normally, the membranes it breaks during labor. Premature rupture of the amniotic sac (PRAS is defined as rupture of membranes before the onset of labor. Premature rupture of the fetal membrane, which occurs before 37 weeks of gestation, usually, refers to preterm premature rupture of membranes. Despite advances in the care period, premature rupture of membranes and premature rupture of membranes preterm continue to be regarded as serious obstetric complications. On the term 8% - 10% of pregnant women have premature rupture of membranes; these women are at increased risk of intrauterine infections, where the interval between membrane rupture and expulsion is rolled-over. Premature rupture of membranes preterm occurs in approximately 1% of all pregnancies and is associated with 30% -40% of preterm births. Thus, it is important to identify the cause of pre-term birth (after less than 37 completed weeks of "gestation" and its complications, including respiratory distress syndrome, neonatal infection and intraventricular hemorrhage. Objectives: the development of the protocol of the clinical trial on patients with impending preterm birth, study clinical and statistical on the socio-demographic characteristics of patients with imminent preterm birth; clinical condition of patients and selection of cases that could benefit from the application of interventional therapy; preclinical investigation (biological and imaging of patients with imminent preterm birth; the modality therapy; clinical investigation of the effectiveness of short

  20. 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.

  1. 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.

  2. Analysis code for large rupture accidents in ATR. SENHOR/FLOOD/HEATUP

    International Nuclear Information System (INIS)

    1997-08-01

    In the evaluation of thermo-hydraulic transient change, the behavior of core reflooding and the transient change of fuel temperature in the events which are classified in large rupture accidents of reactor coolant loss, that is the safety evaluation event of the ATR, the analysis codes for thermo-hydraulic transient change at the time of large rupture SENHOR, for core reflooding characteristics FLOOD and for fuel temperature HEATUP are used, respectively. The analysis code system for loss of coolant accident comprises the analysis code for thermo-hydraulic transient change at the time of medium and small ruptures LOTRAC in addition to the above three codes. Based on the changes with time lapse of reactor thermal output and steam drum pressure obtained by the SENHOR, average reflooding rate is analyzed by the FLOOD, and the time of starting the turnaround of fuel cladding tube temperature and the heat transfer rate after the turnaround are determined. Based on these data, the detailed temperature change of fuel elements is analyzed by the HEATUP, and the highest temperature and the amount of oxidation of fuel cladding tubes are determined. The SENHOR code, the FLOOD code and the HEATUP code and various models for these codes are explained. The example of evaluation and the sensitivity analysis of the ATR plant are reported in the Appendix. (K.I.)

  3. Unsteady coupling effects of wet steam in steam turbines flows

    International Nuclear Information System (INIS)

    Blondel, Frederic

    2014-01-01

    In addition to conventional turbomachinery problems, both the behavior and performances of steam turbines are highly dependent on the vapour thermodynamic state and the presence of a liquid phase. EDF, the main French electricity producer, is interested in further developing its' modelling capabilities and expertise in this area to allow for operational studies and long-term planning. This PhD thesis explores the modelling of wetness formation and growth in a steam turbine and an analysis of the coupling between the liquid phase and the main flow unsteadiness. To this end, the work in this thesis took the following approach. Wetness was accounted for using a homogeneous model coupled with transport equations to take into account the effects of non-equilibrium phenomena, such as the growth of the liquid phase and nucleation. The real gas attributes of the problem demanded adapted numerical methods. Before their implementation in the 3D elsA solver, the accuracy of the chosen models was tested using a developed one-dimensional nozzle code. In this manner, various condensation models were considered, including both poly-dispersed and monodispersed behaviours of the steam. Finally, unsteady coupling effects were observed from several perspectives (1D, 1D - 3D, 3D), demonstrating the ability of the method of moments to sustain unsteady phenomena which were not apparent in a simple monodispersed model. (author)

  4. Calculation of the stationary mass velocity of steam mixtures and of the recoil forces occurring

    International Nuclear Information System (INIS)

    Pana, P.

    1976-11-01

    The best known theories for steam flow (e.g. after pipe rupture within the primary coolant loop of a nuclear power plant) are deeply discussed, the theory of the modified-Bernoulli-equation for the subcooled region, the Moody theory and the homogenious euquilibrium theory for the steam-water region, and the theory of the perfect gas for the superheated region. The calculated mass velocity and thrust coefficient is shown for the whole h-s chart, including various initial pressures and Zeta values as parameter. The comparison of the results leads to important conclusions, concerning conservatism and appropreateness of the considered theories, for friction and frictionless flow. (orig./HP) [de

  5. Predictions of structural integrity of steam generator tubes under normal operating, accident, and severe accident conditions

    International Nuclear Information System (INIS)

    Majumdar, S.

    1996-09-01

    Available models for predicting failure of flawed and unflawed steam generator tubes under normal operating, accident, and severe accident conditions are reviewed. Tests conducted in the past, though limited, tended to show that the earlier flow-stress model for part-through-wall axial cracks overestimated the damaging influence of deep cracks. This observation is confirmed by further tests at high temperatures as well as by finite element analysis. A modified correlation for deep cracks can correct this shortcoming of the model. Recent tests have shown that lateral restraint can significantly increase the failure pressure of tubes with unsymmetrical circumferential cracks. This observation is confirmed by finite element analysis. The rate-independent flow stress models that are successful at low temperatures cannot predict the rate sensitive failure behavior of steam generator tubes at high temperatures. Therefore, a creep rupture model for predicting failure is developed and validated by tests under varying temperature and pressure loading expected during severe accidents

  6. VAMCIS, a new measuring channel for continuous monitoring of leak rates inside PWR steam generators

    International Nuclear Information System (INIS)

    Champion, G.; Dubail, A.; Lefevre, F.

    1988-01-01

    In order to assess the primary to secondary leakage, radioactive isotopes, formed in the primary coolant as a result of fission or neutron capture, are usually monitored in the pressurized water reactor (PWR) secondary coolant. Conventional methods mainly based on the detection of 133 Xe, tritium, and 41 Ar are widely used on French Electricite de France (EdF) PWRs. Some years ago, it appeared necessary to improve both leak rate assessments and steam generator tube rupture (SGTR) detection. A volumetric activity measuring channel inside steam (VAMCIS) has been developed for this purpose. The SGTR that occurred at the North Anna PWR has focused the attention of safety authorities on this new measuring channel. It is planned to implement VAMCIS at North Anna in order to check the leak rate variations more accurately

  7. Method for estimating steam hammer effects on swing-check valves during closure

    International Nuclear Information System (INIS)

    Uram, E.M.

    1976-01-01

    Relationships are developed for estimating the disk impact velocity resulting from a free swing closure of swing-check valves in normal flow and for pipe rupture. They derive from a phase-plane solution of the differential equation for the disk motion that accounts for the nature of the valve pressure drop variation due to steam-hammer effects during closure. For closure in normal flow, the method presented has a more correct foundation than that given in reference where a constant, average valve pressure differential based upon the steady-state pressure drop for the total piping system (which has no real relationship to the steam-hammer-induced valve pressure changes during the closure transient) is used in the valve disk motion equation

  8. The deformation of zircaloy PWR cladding with low internal pressures, under mainly convective cooling by steam

    International Nuclear Information System (INIS)

    Hindle, E.D.; Mann, C.A.; Reynolds, A.E.

    1981-01-01

    The deformation behaviour is reported of specimens of Zircaloy PWR fuel cladding when directly heated in flowing steam. The range of internal pressures studied was 0.69-2.07 MPa; this extended earlier studies using higher pressures. The specimens were ramped and then held at a steady test temperature until rupture or until 600 seconds had elapsed. Under these conditions it was found that extended deformation occurred with pressures down to 1 MPa at temperatures up to 900 deg C. At lower pressures and higher temperatures there was no large extended deformation; this is believed to result from the effects of oxidation

  9. Application of probabilistic fracture mechanics to optimize the maintenance of PWR steam generator tubes

    International Nuclear Information System (INIS)

    Pitner, P.; Riffard, T.

    1993-09-01

    This paper describes the COMPROMIS code developed by Electricite de France (EDF) to optimize the tube bundle maintenance of steam generators (SG). The model, based on probabilistic fracture mechanics, makes it possible to quantify the influence of in-service inspections and maintenance work on the risk of an SG tube rupture, taking all significant parameters into account as random variables (initial defect size distribution, reliability of nondestructive detection and sizing, crack initiation and propagation, critical sizes, leak before risk of break, etc). (authors). 14 figs., 4 tabs., 12 refs

  10. 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)

  11. Modelling of WWER-1000 steam generators by REALP5/MOD3.2 code

    Energy Technology Data Exchange (ETDEWEB)

    D`Auria, F.; Galassi, G.M. [Univ. of Pisa (Italy); Frogheri, M. [Univ. of Genova (Italy)

    1997-12-31

    The presentation summarises the results of best estimate calculations carried out with reference to the WWER-1000 Nuclear Power Plant, utilizing a qualified nodalization set-up for the Relap5/Mod3.2 code. The nodalization development has been based on the data of the Kozloduy Bulgarian Plant. The geometry of the steam generator imposed drastic changes in noding philosophy with respect to what is suitable for the U-tubes steam generators. For the secondary side a symmetry axis was chosen to separate (in the nodalization) the hot and the cold sides of the tubes. In this way the secondary side of the steam generators was divided into three zones: (a) the hot zone including the hot collector and the hot l/2 parts of the tubes; (b) the cold zone including the cold collector and the cold 1/2 parts of the tubes; (c) the downcomer region, where down flow is assumed. As a consequence of above in the primary side more nodes are placed on the hot side of the tubes. Steady state and transient qualification has been achieved, considering the criteria proposed at the University of Pisa, utilizing plant transient data from the Kozloduy and the Ukrainian Zaporosche Plants. The results of the application of the qualified WWER-1000 Relap5/Mod3.2 nodalization to various transients including large break LOCA, small break LOCA and steam generator tube rupture, together with a sensitivity analysis on the steam generators, are reported in the presentation. Emphasis is given to the prediction of the steam generators performances. 23 refs.

  12. Modelling of WWER-1000 steam generators by REALP5/MOD3.2 code

    Energy Technology Data Exchange (ETDEWEB)

    D` Auria, F; Galassi, G M [Univ. of Pisa (Italy); Frogheri, M [Univ. of Genova (Italy)

    1998-12-31

    The presentation summarises the results of best estimate calculations carried out with reference to the WWER-1000 Nuclear Power Plant, utilizing a qualified nodalization set-up for the Relap5/Mod3.2 code. The nodalization development has been based on the data of the Kozloduy Bulgarian Plant. The geometry of the steam generator imposed drastic changes in noding philosophy with respect to what is suitable for the U-tubes steam generators. For the secondary side a symmetry axis was chosen to separate (in the nodalization) the hot and the cold sides of the tubes. In this way the secondary side of the steam generators was divided into three zones: (a) the hot zone including the hot collector and the hot l/2 parts of the tubes; (b) the cold zone including the cold collector and the cold 1/2 parts of the tubes; (c) the downcomer region, where down flow is assumed. As a consequence of above in the primary side more nodes are placed on the hot side of the tubes. Steady state and transient qualification has been achieved, considering the criteria proposed at the University of Pisa, utilizing plant transient data from the Kozloduy and the Ukrainian Zaporosche Plants. The results of the application of the qualified WWER-1000 Relap5/Mod3.2 nodalization to various transients including large break LOCA, small break LOCA and steam generator tube rupture, together with a sensitivity analysis on the steam generators, are reported in the presentation. Emphasis is given to the prediction of the steam generators performances. 23 refs.

  13. The roentgenographic findings of achilles tendon rupture

    Energy Technology Data Exchange (ETDEWEB)

    Seouk, Kang Hyo; Keun, Rho Yong [Shilla General Hospital, Seoul (Korea, Republic of)

    1999-03-01

    To evaluate the diagnostic value of a lateral view of the ankles in Achilles tendon rupture. We performed a retrospective analysis of the roentgenographic findings of 15 patients with surgically proven Achilles tendon rupture. Four groups of 15 patients(normal, ankle sprain, medial lateral malleolar fracture, and calcaneal fracture) were analysed as reference groups. Plain radiographs were reviewed with regard to Kager's triangle, Arner's sign, Toygar's angle, ill defined radiolucent shadow through the Achilles tendon, sharpness of the anterior margin of Achilles tendon, and meniscoid smooth margin of the posterior skin surface of the ankle. Kager's triangle was deformed and disappeared after rupture of the Achilles tendon in nine patients(60%) with operative verification of the rupture, six patients(40%) had a positive Arner's sign, while none had a diminished Toygars angle. In 13 patients(87%) with a ruptured Achilles tendon, the thickness of this was nonuniform compared with the reference group. The anterior margin of the Achilles tendon became serrated and indistinct in 14 patients(93%) in whom this was ruptured. An abnormal ill defined radiolucent shadow through the Achilles tendon was noted in nine patient(60%), and nonparallelism between the anterior margin of the Achilles tendon and posterior skin surface of the ankle was detected in 11 patients(73%). The posterior skin surface of the ankle had a nodular surface margin in 13 patients(87%). A deformed Kager's triangle and Achilles tendon, and an abnormal ill defined radiolucent shadow through the Achilles tendon in a lateral view of the ankles are important findings for the diagnesis of in diagnosing achilles tendon rupture.

  14. The roentgenographic findings of achilles tendon rupture

    International Nuclear Information System (INIS)

    Seouk, Kang Hyo; Keun, Rho Yong

    1999-01-01

    To evaluate the diagnostic value of a lateral view of the ankles in Achilles tendon rupture. We performed a retrospective analysis of the roentgenographic findings of 15 patients with surgically proven Achilles tendon rupture. Four groups of 15 patients(normal, ankle sprain, medial lateral malleolar fracture, and calcaneal fracture) were analysed as reference groups. Plain radiographs were reviewed with regard to Kager's triangle, Arner's sign, Toygar's angle, ill defined radiolucent shadow through the Achilles tendon, sharpness of the anterior margin of Achilles tendon, and meniscoid smooth margin of the posterior skin surface of the ankle. Kager's triangle was deformed and disappeared after rupture of the Achilles tendon in nine patients(60%) with operative verification of the rupture, six patients(40%) had a positive Arner's sign, while none had a diminished Toygars angle. In 13 patients(87%) with a ruptured Achilles tendon, the thickness of this was nonuniform compared with the reference group. The anterior margin of the Achilles tendon became serrated and indistinct in 14 patients(93%) in whom this was ruptured. An abnormal ill defined radiolucent shadow through the Achilles tendon was noted in nine patient(60%), and nonparallelism between the anterior margin of the Achilles tendon and posterior skin surface of the ankle was detected in 11 patients(73%). The posterior skin surface of the ankle had a nodular surface margin in 13 patients(87%). A deformed Kager's triangle and Achilles tendon, and an abnormal ill defined radiolucent shadow through the Achilles tendon in a lateral view of the ankles are important findings for the diagnesis of in diagnosing achilles tendon rupture

  15. Arthroscintigraphy in suspected rotator cuff rupture

    International Nuclear Information System (INIS)

    Gratz, S.; Behr, T.; Becker, W.; Koester, G.; Vosshenrich, R.; Grabbe, E.

    1998-01-01

    Aim: In order to evaluate the diagnostic efficiency of arthroscintigraphy in suspected rotator cuff ruptures this new imaging procedure was performed 20 times in 17 patients with clinical signs of a rotator cuff lesion. The scintigraphic results were compared with sonography (n=20), contrast arthrography (n=20) and arthroscopy (n=10) of the shoulder joint. Methods: After performing a standard bone scintigraphy with intravenous application of 300 MBq 99m-Tc-methylene diphosphonate (MDP) for landmarking of the shoulder region arthroscintigraphy was performed after an intraarticular injection of 99m-Tc microcolloid (ALBU-RES 400 μCi/5 ml). The application was performed either in direct combination with contrast arthrography (n=10) or ultrasound conducted mixed with a local anesthetic (n=10). Findings at arthroscopical surgery (n=10) were used as the gold standard. Results: In case of complete rotator cuff rupture (n=5), arthroscintigraphy and radiographic arthrography were identical in 5/5. In one patient with advanced degenerative alterations of the shoulder joint radiographic arthrography incorrectly showed a complete rupture which was not seen by arthroscintigraphy and endoscopy. In 3 patients with incomplete rupture, 2/3 results were consistant. A difference was seen in one patient with a rotator cuff, that has been already revised in the past and that suffered of capsulitis and calcification. Conclusion: Arthroscinitgraphy is a sensitive technique for detection of rotator cuff ruptures. Because of the lower viscosity of the active compound, small ruptures can be easily detected, offering additional value over radiographic arthrography and ultrasound, especially for evaluation of incomplete cuff ruptures. (orig.) [de

  16. Spontaneous rupture of the esophagus associated with intramural rupture caused by ingestion of weeding medicine (Lasso)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jong Keon; Park, Heung Il; Kwun, Chung Sik [Chun Nam University College of Medicine, Kwangju (Korea, Republic of)

    1975-06-15

    This is a report of a case of spontaneous rupture of the esophagus associated with intramural rupture caused by ingestion of weeding medicine for the purpose of suicide in a 27 year old Korean male whose chief complaints were dyspnea, epigastric pain, swallowing disturbance, and hoarseness for 3 days prior to admission. A review of literature is submitted.

  17. Depth-Sizing Technique for Crack Indications in Steam Generator Tubing

    International Nuclear Information System (INIS)

    Cho, Chan Hee; Lee, Hee Jeong; Kim, Hong Deok

    2009-01-01

    The nuclear power plants have been safely operated by plugging the steam generator tubes which have the crack indications. Tube rupture events can occur if analysts fail to detect crack indications during in-service inspection. There are various types of crack indication in steam generator tubes and they have been detected by the eddy current test. The integrity assessment should be performed using the crack-sizing results from eddy current data when the crack indication is detected. However, it is not easy to evaluate the crack-depth precisely and consistently due to the complexity of the methods. The current crack-sizing methods were reviewed in this paper and the suitable ones were selected through the laboratory tests. The retired steam generators of Kori Unit 1 were used for this study. The round robin tests by the domestic qualified analysts were carried out and the statistical models were introduced to establish the appropriate depth-sizing techniques. It is expected that the proposed techniques in this study can be utilized in the Steam Generator Management Program

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-15

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

  20. Key Findings from the Artist Project on Aerosol Retention in a Dry Steam Generator

    Directory of Open Access Journals (Sweden)

    Abdelouahab Dehbi

    2016-08-01

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

  1. Ultrasonic inspection for wastage in the LMFBR steam generator due to sodium--water reactions

    International Nuclear Information System (INIS)

    Neely, H.H.; Renger, L.

    1977-01-01

    As part of a program to study the results of large sodium-water reactions in the LMFBR Steam Generator, a boreside ultrasonic inspection device was developed to measure the wall thickness and diameter of the 2- 1 / 4 Cr-1 Mo, 0.397 in. I.D. steam tubes. The reaction was created in a near prototype steam generator by guillotine-type rupture of a steam tube, while the generator was at operating conditions. Wastage occurred on the surrounding tubes due to the high temperature reaction. The UT test instrument was designed to operate with a 15 MHz transducer in the pulse-echo shear-wave mode, with a sampling rate of 10 4 /sec. System outputs are diameter, wall thickness, attitude and axial position of the transducer. All are displayed digitally and may be recorded. Measurements are fed into a computer for later retrieval, and/or cascaded outputs into an x-y recorded displaying either out-of-limit or thickness data. The UT data taken in this experiment were consistent with physical measurements on a tube which was removed from the generator after the test. A machined flat 1 / 8 -inch long and 0.002-inch deep could readily be detected

  2. 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)

  3. Steam coal mines of tomorrow

    Energy Technology Data Exchange (ETDEWEB)

    McCloskey, G

    1986-07-01

    A comprehensive review of new steam coal mines being planned or developed worldwide. It shows that at least 20 major mines with a combined annual output of 110 million tonnes per annum, could add their coal to world markets in the next 10 years. The review highlights: substantial activity in Australia with at least four major mines at advanced planning stages; a strengthening of the South American export industry with 4 major mines operating in 10 years compared with just one today; no major export mines being developed in the traditional US mining areas; and the emergence of Indonesia as a major steam coal producer/exporter. The review also shows a reduction in cost/output ratios, and also the proximity of the new mines to existing infrastructure (e.g. export terminals, rail links).

  4. Organic evaporator steam valve failure

    International Nuclear Information System (INIS)

    Jacobs, R.A.

    1992-01-01

    Defense Waste Processing Facility (DWPF) Technical has requested an analysis of the capacity of the Organic Evaporator (OE) condenser (OEC) be performed to determine its capability in the case where the OE steam flow control valve fails open. Calculations of the OE boilup and the OEC heat transfer coefficient indicate the OEC will have more than enough capacity to remove the heat at maximum OE boilup. In fact, the Salt Cell Vent Condenser (SCVC) should also have sufficient capacity to handle the maximum OE boilup. Therefore, it would require simultaneous loss of OEC and/or SCVC condensing capacity for the steam valve failure to cause high benzene in the Process Vessel Vent System (PVVS)

  5. Closed loop steam cooled airfoil

    Science.gov (United States)

    Widrig, Scott M.; Rudolph, Ronald J.; Wagner, Gregg P.

    2006-04-18

    An airfoil, a method of manufacturing an airfoil, and a system for cooling an airfoil is provided. The cooling system can be used with an airfoil located in the first stages of a combustion turbine within a combined cycle power generation plant and involves flowing closed loop steam through a pin array set within an airfoil. The airfoil can comprise a cavity having a cooling chamber bounded by an interior wall and an exterior wall so that steam can enter the cavity, pass through the pin array, and then return to the cavity to thereby cool the airfoil. The method of manufacturing an airfoil can include a type of lost wax investment casting process in which a pin array is cast into an airfoil to form a cooling chamber.

  6. Steam generator waterlancing at DNGS

    International Nuclear Information System (INIS)

    Seppala, D.; Malaugh, J.

    1995-01-01

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

  7. Splenic rupture following idiopathic rupture of the urinary bladder presenting as acute abdomen

    Directory of Open Access Journals (Sweden)

    Jurisic D

    2007-01-01

    Full Text Available Idiopathic rupture of the urinary bladder is an uncommon condition and represents less than 1% of all bladder rupture cases. In most of the cases the main etiological factor was heavy alcohol ingestion. A combined injury of the spleen and bladder is a very rare condition that is almost often associated with trauma and foreign bodies. In this paper we present the extremely rare clinical course of acute abdomen caused by a combined spontaneous intraperitoneal injury; spontaneous rupture of the urinary bladder and spleen. According to our opinion, spontaneous bladder rupture caused by bladder distension due to alcohol ingestion led to urinary ascites and abdominal distension. Finally, repeated minor abdominal blunt trauma during everyday life, to a moderately distended abdomen caused a spontaneous splenic rupture in the patient with abnormal coagulation studies.

  8. 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)

  9. Large nuclear steam turbine plants

    International Nuclear Information System (INIS)

    Urushidani, Haruo; Moriya, Shin-ichi; Tsuji, Kunio; Fujita, Isao; Ebata, Sakae; Nagai, Yoji.

    1986-01-01

    The technical development of the large capacity steam turbines for ABWR plants was partially completed, and that in progress is expected to be completed soon. In this report, the outline of those new technologies is described. As the technologies for increasing the capacity and heightening the efficiency, 52 in long blades and moisture separating heaters are explained. Besides, in the large bore butterfly valves developed for making the layout compact, the effect of thermal efficiency rise due to the reduction of pressure loss can be expected. As the new technology on the system side, the simplification of the turbine system and the effect of heightening the thermal efficiency by high pressure and low pressure drain pumping-up method based on the recent improvement of feed water quality are discussed. As for nuclear steam turbines, the actual records of performance of 1100 MW class, the largest output at present, have been obtained, and as a next large capacity machine, the development of a steam turbine of 1300 MWe class for an ABWR plant is in progress. It can be expected that by the introduction of those new technologies, the plants having high economical efficiency are realized. (Kako, I.)

  10. 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)

  11. Ultrasonogram of tubal pregnancy: Unruptured or ruptured?

    International Nuclear Information System (INIS)

    Lee, Hee Jung; Kim, Jung Sik; Lee, Sung Moon; Kim, Hong; Woo, Seong Ku; Yeun, Seong Do; Lee, Doo Ryong

    1993-01-01

    To evaluate the usefulness of ultrasonography in the differentiation of the unruptured ones, we performed prospective ultrasonographic study in 56 patients with tubal pregnancy and subsequently correlated with the surgical (45 cases) and laparoscopic (11 cases) findings. The criterion for unruptured tubal pregnancy was the visualization of extra uterine gestational sac or a mass that was separated from the ipsilateral ovary. All cases were confirmed surgically or laparoscopically within 48 hours after sonographic examination. Forty five (80%)cases were proved as unruptured, tubal pregnancy, while 11 (20%) as ruptured one. Extra uterine gestational sac was seen in 31/45 (69%) in unruptured group and 2/11 (18%) in ruptured group. A mass separated from the ipsilateral ovary was seen in 7 unruptured tubal pregnancies and 2 ruptured pregnancies. A mass that was not separated from the ipsilateral ovary was seen in 13 cases. Seven of these were ruptured pregnancies, while 6 were unrupture dones. Using our criteria, the unruptured tubal pregnancy was diagnosed with a sensitivity of 84%, a specificity of 64%, an accuracy of 80%, a positive predictive value of 90% and a negative predictive value of 50%. We conclude that ultrasonography is useful in the differentiation of the unruptured tubal pregnancy from the ruptured one

  12. WWER steam generator tube structural and leakage integrity

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  13. Creep performance of oxide ceramic fiber materials at elevated temperature in air and in steam

    Science.gov (United States)

    Armani, Clinton J.

    comparisons with experimental results. Additionally, the utility of the Monkman-Grant relationship to predicting creep-rupture life of the fiber tows at elevated temperature in air and in steam was demonstrated. Furthermore, the effects of steam on the compressive creep performance of bulk ceramic materials were also studied. Performance of fine grained, polycrystalline alumina (Al2O3) was investigated at 1100 and 1300°C in air and in steam. To evaluate the effect of silica doping during material processing both undoped and silica doped polycrystalline alumina specimens were tested. Finally, compressive creep performance of yttrium aluminum garnet (YAG, Y3Al5O12) was evaluated at 1300°C in air and in steam. Both undoped and silica doped YAG specimens were included in the study. YAG is being considered as the next-generation oxide fiber material. However, before considerable funding and effort are invested in a fiber development program, it is necessary to evaluate the creep performance of YAG at elevated temperature in steam. Results of this research demonstrated that both the undoped YAG and the silica doped YAG exhibited exceptional creep resistance at 1300°C in steam for grain sizes ˜1 microm. These results supplement the other promising features of YAG that make it a strong candidate material for the next generation ceramic fiber.

  14. A study of the effect of maintenance on the safety of a mechanical system subject to aging and its application to steam generator tube degradation

    International Nuclear Information System (INIS)

    Dussarte, D.

    1991-11-01

    The different degradation mechanisms to which pressurized water reactor steam generator tubes are observed to be subject may result in the risk of their rupture being greater than anticipated. Prevention of tube rupture essentially consists of inspections during outages of the units and applying appropriate criteria for the withdrawal of defective tubes from service. Planning such measures implies being able to gauge the effectiveness of the action taken. This document describes a proposed technique for quantifying the effects of the preventive maintenance we have had to develop to address this problem and, hence, to obtain material for assessing the action taken by the utility. (author)

  15. Creep-rupture behavior of 2-1/4 Cr-1 Mo steel, Alloy 800H and Hastelloy Alloy X in a simulated HTGR helium environment

    International Nuclear Information System (INIS)

    Lai, G.Y.; Wolwowicz, R.J.

    1979-12-01

    Creep-rupture testing was conducted on 1 1/4 Cr-1 Mo steel, Alloy 800H and Hastelloy Alloy X in flowing helium containing nominal concentration of following gases: 1500 μatm H 2 , 450 μatm CO, 50 μatm CH 4 , 50 μatm H 2 O and 5 μatm CO 2 . This environment is believed to represent maximum permissible levels of impurities in the primary coolant for the steam-cycle system of a high-temperature gas-cooled reactor (HTGR) when it is operating continuously with a water and/or steam leak at technical specification limits. Two or three heats of material for each alloy were investigated. Tests were conducted at 482 0 C and 760 0 C (1200 0 F and 1400 0 F) for Alloy 800H, and at 760 0 C and 871 0 C (1400 0 F and 1600 0 F) for Hastelloy Alloy X for times up to 10,000 h. Selected tests were performed on same heat of material in both air and helium environments to make a direct comparison of creep-rupture behaviors between two environments. Metallurgical evaluation was performed on selected post test specimens with respect to gas-metal interactions which included oxidation, carburization and/or decarburization. Correlation between gaseous corrosion and creep-rupture behavior was attempted. Limited tests were also performed to investigate the specimen size effects on creep-rupture behavior in the helium environment

  16. Pressure tube rupture in a closed tank

    International Nuclear Information System (INIS)

    Khater, H.A.; Hadaller, G.I.; Stern, F.

    1985-06-01

    A study has been prepared on the feasibility of conducting pressure tube/calandria tube rupture tests in a closed tank, simulating a scaled-down calandria vessel. The study includes: i) a review of previous work, ii) an analytical investigation of the scaling problem of the calandria vessel and relevant in-core structures, iii) selection of a method for initiating pressure tube/calandria tube rupture, iv) a set of specifications for the test assembly, v) general arrangement drawings, vi) a proposal for a test matrix, vii) a survey and evaluation of existing facilities which could provide the required high pressure, temperature and fluid inventory, and viii) a cost estimate for the detailed design and construction, instrumentation, data acquisition and reduction, testing and reporting. The study concludes that it is both technically and practically feasible to conduct pressure tube rupture tests in a closed tank

  17. Bladder rupture caused by postpartum urinary retention.

    Science.gov (United States)

    Dueñas-García, Omar Felipe; Rico, Hugo; Gorbea-Sanchez, Viridiana; Herrerias-Canedo, Tomas

    2008-08-01

    Postpartum bladder rupture is an uncommon surgical emergency and a diagnostic challenge. A primigravida delivered a healthy newborn without complications at 39.4 weeks of gestation. The patient was admitted 80 hours postpartum with abdominal pain, oliguria, hematuria, and pain that worsened during the previous 4 hours. An inserted Foley catheter drained only a small amount of urine, and serum creatinine was elevated (3.5 mg/dL). A laparotomy was performed and revealed a 10-cm hole in the urinary bladder. The bladder was repaired and the patient was discharged 15 days after surgery. The follow-up cystoscopy revealed adequate healing of the bladder. Urinary retention can lead to serious complications, including bladder rupture. Postpartum bladder rupture due to urinary retention should be ruled out if there is a history of abdominal pain, oliguria, and elevated of serum creatinine.

  18. Component external leakage and rupture frequency estimates

    International Nuclear Information System (INIS)

    Eide, S.A.; Khericha, S.T.; Calley, M.B.; Johnson, D.A.; Marteeny, M.L.

    1991-11-01

    In order to perform detailed internal flooding risk analyses of nuclear power plants, external leakage and rupture frequencies are needed for various types of components - piping, valves, pumps, flanges, and others. However, there appears to be no up-to-date, comprehensive source for such frequency estimates. This report attempts to fill that void. Based on a comprehensive search of Licensee Event Reports (LERs) contained in Nuclear Power Experience (NPE), and estimates of component populations and exposure times, component external leakage and rupture frequencies were generated. The remainder of this report covers the specifies of the NPE search for external leakage and rupture events, analysis of the data, a comparison with frequency estimates from other sources, and a discussion of the results

  19. Ruptured Massa Intermedia Secondary to Hydrocephalus.

    Science.gov (United States)

    El Damaty, Ahmed; Langner, Soenke; Schroeder, Henry W S

    2017-01-01

    We report a case of ruptured massa intermedia (MI) as a sequela of hydrocephalus. A single case report is presented describing the sequelae of tumor bed hematoma after a posterior fossa hemangioblastoma resection in which the patient bled 3 days after surgery, resulting in secondary hydrocephalus and subsequently dilatation of the third ventricle, which resulted in rupture of the MI. The patient was managed on emergency basis with an external ventricular drain then endoscopically with a third ventriculostomy and clot extraction. Absent MI is not uncommon in hydrocephalic patients, and it is assumed to be the result of rupture from acute dilatation of the third ventricle. Our case report proves this assumption and documents the presence and absence of the MI before and after developing hydrocephalus. Copyright © 2016 Elsevier Inc. All rights reserved.

  20. Creep rupture behavior of unidirectional advanced composites

    Science.gov (United States)

    Yeow, Y. T.

    1980-01-01

    A 'material modeling' methodology for predicting the creep rupture behavior of unidirectional advanced composites is proposed. In this approach the parameters (obtained from short-term tests) required to make the predictions are the three principal creep compliance master curves and their corresponding quasi-static strengths tested at room temperature (22 C). Using these parameters in conjunction with a failure criterion, creep rupture envelopes can be generated for any combination of in-plane loading conditions and ambient temperature. The analysis was validated experimentally for one composite system, the T300/934 graphite-epoxy system. This was done by performing short-term creep tests (to generate the principal creep compliance master curves with the time-temperature superposition principle) and relatively long-term creep rupture tensile tests of off-axis specimens at 180 C. Good to reasonable agreement between experimental and analytical results is observed.

  1. Linguine sign in musculoskeletal imaging: calf silicone implant rupture.

    Science.gov (United States)

    Duryea, Dennis; Petscavage-Thomas, Jonelle; Frauenhoffer, Elizabeth E; Walker, Eric A

    2015-08-01

    Imaging findings of breast silicone implant rupture are well described in the literature. On MRI, the linguine sign indicates intracapsular rupture, while the presence of silicone particles outside the fibrous capsule indicates extracapsular rupture. The linguine sign is described as the thin, wavy hypodense wall of the implant within the hyperintense silicone on T2-weighted images indicative of rupture of the implant within the naturally formed fibrous capsule. Hyperintense T2 signal outside of the fibrous capsule is indicative of an extracapsular rupture with silicone granuloma formation. We present a rare case of a patient with a silicone calf implant rupture and discuss the MRI findings associated with this condition.

  2. Linguine sign in musculoskeletal imaging: calf silicone implant rupture

    International Nuclear Information System (INIS)

    Duryea, Dennis; Petscavage-Thomas, Jonelle; Frauenhoffer, Elizabeth E.; Walker, Eric A.

    2015-01-01

    Imaging findings of breast silicone implant rupture are well described in the literature. On MRI, the linguine sign indicates intracapsular rupture, while the presence of silicone particles outside the fibrous capsule indicates extracapsular rupture. The linguine sign is described as the thin, wavy hypodense wall of the implant within the hyperintense silicone on T2-weighted images indicative of rupture of the implant within the naturally formed fibrous capsule. Hyperintense T2 signal outside of the fibrous capsule is indicative of an extracapsular rupture with silicone granuloma formation. We present a rare case of a patient with a silicone calf implant rupture and discuss the MRI findings associated with this condition. (orig.)

  3. Linguine sign in musculoskeletal imaging: calf silicone implant rupture

    Energy Technology Data Exchange (ETDEWEB)

    Duryea, Dennis; Petscavage-Thomas, Jonelle [Milton S. Hershey Medical Center, Department of Radiology, H066, 500 University Drive, P.O. Box 850, Hershey, PA (United States); Frauenhoffer, Elizabeth E. [Milton S. Hershey Medical Center, Department of Pathology, 500 University Drive, P.O. Box 850, Hershey, PA (United States); Walker, Eric A. [Milton S. Hershey Medical Center, Department of Radiology, H066, 500 University Drive, P.O. Box 850, Hershey, PA (United States); Uniformed Services University of the Health Sciences, Department of Radiology and Nuclear Medicine, Bethesda, MD, 20814 (United States)

    2015-08-15

    Imaging findings of breast silicone implant rupture are well described in the literature. On MRI, the linguine sign indicates intracapsular rupture, while the presence of silicone particles outside the fibrous capsule indicates extracapsular rupture. The linguine sign is described as the thin, wavy hypodense wall of the implant within the hyperintense silicone on T2-weighted images indicative of rupture of the implant within the naturally formed fibrous capsule. Hyperintense T2 signal outside of the fibrous capsule is indicative of an extracapsular rupture with silicone granuloma formation. We present a rare case of a patient with a silicone calf implant rupture and discuss the MRI findings associated with this condition. (orig.)

  4. Lungmen ABWR containment analyses during short-term main steam line break LOCA using GOTHIC

    International Nuclear Information System (INIS)

    Chen, Yen-Shu; Yuann, Yng-Ruey; Dai, Liang-Che

    2012-01-01

    Highlights: ► The Lungmen ABWR containment responses due to the main steam line break are analyzed. ► In the Lungmen FSAR, the peak drywell temperature is greater than the designed value. ► GOTHIC is used to calculate the containment responses in this study. ► With more realistic conditions, the drywell temperature can be reasonably suppressed. - Abstract: Lungmen Nuclear Power Plant in Taiwan is a GE-designed twin-unit Advanced Boiling Water Reactor (ABWR) plant with rated thermal power of 3926 MWt. Both units are currently under construction. In the Lungmen Final Safety Analysis Report (FSAR) section 6.2, the calculated peak drywell temperature during the short-term Main Steam Line Break (MSLB) event is 176.3 °C, which is greater than the designed temperature of 171.1 °C. It resulted in a controversial issue in the FSAR review process conducted by the Atomic Energy Council in Taiwan. The purpose of this study is to independently investigate the Lungmen ABWR containment pressure and temperature responses to the MSLB using the GOTHIC program. Blowdown conditions are either calculated by using a simplified reactor vessel volume in GOTHIC model, or provided by the RELAP5 transient analysis. The blowdown flow rate from the steam header side is calculated with a more reasonable pressure loss coefficient of the open main steam isolation valves, and the peak drywell temperature is then reduced. By using the RELAP5 blowdown data, the peak drywell temperature can be further reduced because of the initial liquid entrainment in the blowdown flow. The drywell space is either treated as a single volume, or separated into a upper drywell and a lower drywell to reflect the real configuration of the Lungmen containment. It is also found that a single drywell volume may not present the overheating of the upper drywell. With more realistic approaches and assumptions, the drywell temperature can be reasonably below the design value and the Lungmen containment integrity

  5. A Rare Case of Simultaneous Acute Bilateral Quadriceps Tendon Rupture and Unilateral Achilles Tendon Rupture

    Directory of Open Access Journals (Sweden)

    Wei Yee Leong

    2013-07-01

    Full Text Available Introduction: There have been multiple reported cases of bilateral quadriceps tendon ruptures (QTR in the literature. These injuries frequently associated with delayed diagnosis, which results in delayed surgical treatment. In very unusual cases, bilateral QTRs can be associated with other simultaneous tendon ruptures. Case Report: We present a rare case of bilateral QTR with a simultaneous Achilles Tendon Rupture involving a 31 years old Caucasian man who is a semi-professional body builder taking anabolic steroids. To date bilateral QTR with additional TA rupture has only been reported once in the literature and to our knowledge this is the first reported case of bilateral QTR and simultaneous TA rupture in a young, fit and healthy individual. Conclusion: The diagnosis of bilateral QTR alone can sometimes be challenging and the possibility of even further tendon injuries should be carefully assessed. A delay in diagnosis could result in delay in treatment and potentially worse outcome for the patient. Keywords: Quadriceps tendon rupture; Achilles tendon rupture; Bilateral.

  6. Relap5/Mod2.5 analyses of SG primary collector head rupture in WWER-440 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Szczurek, J. [Inst. of Atomic Energy, Swierk (Poland)

    1995-12-31

    The paper presents the results of the analyses of steam generator (SG) manifold cover rupture performed with RELAP5/MOD2.5 (version provided by RMA, Albuquerque, for PC PPS). The calculations presented are based on RELAP5 input deck for WWER-440/213 Bobunice NPP, developed within the framework of IAEA TC Project RER/9/004. The presented analyses are directed toward determining the maximum amount of reactor coolant discharged into the secondary coolant system and the maximum amount of contaminated coolant release to the atmosphere. In all cases considered in the analysis, maximum ECCS injection capacity is assumed. The paper includes only the cases without any operator actions within the time period covered by the analyses. In particular, the primary loop isolation valves are not used for isolating the broken steam generator. Two scenarios are analysed: with and without the SG safety valve stuck open. 3 refs.

  7. Relap5/Mod2.5 analyses of SG primary collector head rupture in WWER-440 reactor

    International Nuclear Information System (INIS)

    Szczurek, J.

    1995-01-01

    The paper presents the results of the analyses of steam generator (SG) manifold cover rupture performed with RELAP5/MOD2.5 (version provided by RMA, Albuquerque, for PC PPS). The calculations presented are based on RELAP5 input deck for WWER-440/213 Bobunice NPP, developed within the framework of IAEA TC Project RER/9/004. The presented analyses are directed toward determining the maximum amount of reactor coolant discharged into the secondary coolant system and the maximum amount of contaminated coolant release to the atmosphere. In all cases considered in the analysis, maximum ECCS injection capacity is assumed. The paper includes only the cases without any operator actions within the time period covered by the analyses. In particular, the primary loop isolation valves are not used for isolating the broken steam generator. Two scenarios are analysed: with and without the SG safety valve stuck open

  8. Relap5/Mod2.5 analyses of SG primary collector head rupture in WWER-440 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Szczurek, J [Inst. of Atomic Energy, Swierk (Poland)

    1996-12-31

    The paper presents the results of the analyses of steam generator (SG) manifold cover rupture performed with RELAP5/MOD2.5 (version provided by RMA, Albuquerque, for PC PPS). The calculations presented are based on RELAP5 input deck for WWER-440/213 Bobunice NPP, developed within the framework of IAEA TC Project RER/9/004. The presented analyses are directed toward determining the maximum amount of reactor coolant discharged into the secondary coolant system and the maximum amount of contaminated coolant release to the atmosphere. In all cases considered in the analysis, maximum ECCS injection capacity is assumed. The paper includes only the cases without any operator actions within the time period covered by the analyses. In particular, the primary loop isolation valves are not used for isolating the broken steam generator. Two scenarios are analysed: with and without the SG safety valve stuck open. 3 refs.

  9. Thermal-hydraulics in recirculating steam generators

    International Nuclear Information System (INIS)

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

    1981-04-01

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

  10. 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

  11. Analysis of performance for centrifugal steam compressor

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Seung Hwan; Ryu, Chang Kook; Ko, Han Seo [Sungkyunkwan University, Suwon (Korea, Republic of)

    2016-12-15

    In this study, mean streamline and Computational fluid dynamics (CFD) analyses were performed to investigate the performance of a small centrifugal steam compressor using a latent heat recovery technology. The results from both analysis methods showed good agreement. The compression ratio and efficiency of steam were found to be related with those of air by comparing the compression performances of both gases. Thus, the compression performance of steam could be predicted by the compression performance of air using the developed dimensionless parameters.

  12. Analysis of performance for centrifugal steam compressor

    International Nuclear Information System (INIS)

    Kang, Seung Hwan; Ryu, Chang Kook; Ko, Han Seo

    2016-01-01

    In this study, mean streamline and Computational fluid dynamics (CFD) analyses were performed to investigate the performance of a small centrifugal steam compressor using a latent heat recovery technology. The results from both analysis methods showed good agreement. The compression ratio and efficiency of steam were found to be related with those of air by comparing the compression performances of both gases. Thus, the compression performance of steam could be predicted by the compression performance of air using the developed dimensionless parameters

  13. 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

  14. Testicular Rupture: A Tough Nut to Crack

    Directory of Open Access Journals (Sweden)

    Tyler L. Holliday

    2017-07-01

    Full Text Available Blunt scrotal injury represents a diagnostic dilemma for emergency physicians (EP. Consequently, point-of-care ultrasound (POCUS has emerged as a tool for early investigation of the acute scrotum in the emergency department. We describe a case where an EP used scrotal POCUS to immediately visualize the loss of testicular contour and underlying heterogeneous parenchyma to rapidly make the diagnosis of testicular rupture in a young male presenting with scrotal trauma. The use of POCUS in this case expedited therapy, likely improving the patient’s outcome. To our knowledge, this is the first detailed description of testicular rupture diagnosed with POCUS by an EP

  15. Spontaneous Achilles tendon rupture in alkaptonuria

    Directory of Open Access Journals (Sweden)

    Omar A. Alajoulin

    2015-12-01

    Full Text Available Alkaptonuria (AKU is a rare inborn metabolic disease characterized by accumulation of homogentisic acid (HGA. Excretion of HGA in urine causes darkening of urine and its deposition in connective tissues causes dark pigmentation (ochronosis, early degeneration of articular cartilage, weakening of the tendons, and subsequent rupture. In this case report, we present a rare case of a patient presented with unilateral spontaneous rupture of Achilles tendon due to AKU. The patient developed most of the orthopedic manifestations of the disease earlier than typical presentations. Alkaptonuria patients should avoid strenuous exercises and foot straining especially in patients developing early orthopedic manifestations.

  16. Spontaneous Achilles tendon rupture in alkaptonuria.

    Science.gov (United States)

    Alajoulin, Omar A; Alsbou, Mohammed S; Ja'afreh, Somayya O; Kalbouneh, Heba M

    2015-12-01

    Alkaptonuria (AKU) is a rare inborn metabolic disease characterized by accumulation of homogentisic acid (HGA). Excretion of HGA in urine causes darkening of urine and its deposition in connective tissues causes dark pigmentation (ochronosis), early degeneration of articular cartilage, weakening of the tendons, and subsequent rupture. In this case report, we present a rare case of a patient presented with unilateral spontaneous rupture of Achilles tendon due to AKU. The patient developed most of the orthopedic manifestations of the disease earlier than typical presentations. Alkaptonuria patients should avoid strenuous exercises and foot straining especially in patients developing early orthopedic manifestations.

  17. Physiotherapy in patients with rupture LCA

    OpenAIRE

    Čížková, Karolína

    2012-01-01

    Author: Karolína Čížková Institution: Rehab clinic LF UK in Hradec Králové Topic of bachelor's thesis: Physiotherapy in patiens with rupture of ACL Supervisor: Mgr. Zuzana Hamarová Number of pages: 108 Number of annex: 11 Year of vindication 2012 Key words: ACL, knee, rupture of ACL, rehabilitation In general part of this bachelor's thesis is described anatomy, kinesiology and biomechanics of the knee joint, mostly anterior cruciate ligament. It contents causes of injury of anterior cruciate ...

  18. Proximal Hamstring Tendinosis and Partial Ruptures.

    Science.gov (United States)

    Startzman, Ashley N; Fowler, Oliver; Carreira, Dominic

    2017-07-01

    Proximal hamstring tendinosis and partial hamstring origin ruptures are painful conditions of the proximal thigh and hip that may occur in the acute, chronic, or acute on chronic setting. Few publications exist related to their diagnosis and management. This systematic review discusses the incidence, treatment, and prognosis of proximal hamstring tendinosis and partial hamstring ruptures. Conservative treatment measures include nonsteroidal anti-inflammatory drugs, physical therapy, rest, and ice. If these measures fail, platelet-rich plasma or shockwave therapy may be considered. When refractory to conservative management, these injuries may be treated with surgical debridement and hamstring reattachment. [Orthopedics. 2017; 40(4):e574-e582.]. Copyright 2017, SLACK Incorporated.

  19. [Development of Achilles tendon rupture in skiing].

    Science.gov (United States)

    Suckert, K; Benedetto, K P; Vogel, A

    1983-06-01

    This is an analysis of decline of rupture of the Achilles tendon in skiing while there is a steady increase of skiing injuries. Three groups, equipped with three different types of ski boots were observed once on a plane slope on the other hand on a bump track. The simultaneous size of angle of knee and ankle was measured by telemetry. The high plastic ski boot, which obstructs the ankle forward and lateral is apart from the rise of heel in the boot, the safety binding and the new skiing style the main reason for decline of rupture of the Achilles tendon in skiing.

  20. 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