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.

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

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.)

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

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)

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

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

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

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

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.

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)

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

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

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

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)

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.

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

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

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

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

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

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

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)

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)

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

Steam 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.)

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

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.)

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.

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

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.)

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

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

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.

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)

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.)

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)

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

International Nuclear Information System (INIS)

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

1977-10-01

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

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

International Nuclear Information System (INIS)

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

1977-10-01

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

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

International Nuclear Information System (INIS)

Esparza, V.; Sackett, K.E.

1977-10-01

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

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.

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.

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

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

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)

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

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

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

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

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

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

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

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)

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

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

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)

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

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)

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

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.

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)

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

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

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

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.

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

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

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

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)

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)

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

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

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

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)

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

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)

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

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)

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)

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)

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

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)

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)

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

Steam line rupture experiments with the PPOOLEX test facility

International Nuclear Information System (INIS)

Laine, J.; Puustinen, M.

2008-07-01

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

Steam line rupture experiments with the PPOOLEX test facility

Energy Technology Data Exchange (ETDEWEB)

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

2008-07-15

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

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.

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.

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.

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

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

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.)

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

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

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

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)

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.

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

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)

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.

Safety analysis of Ignalina NPP during shutdown conditions

International Nuclear Information System (INIS)

Kaliatka, A.; Uspuras, E.

2000-01-01

The accident analysis for the Ignalina NPP with RBMK-1500 reactors at normal operating conditions and at minimum controlled power level (during startup of the reactor) has been performed in the frame of the project I n-Depth Safety Assessment of the Ignalina NPP , which was completed in 1996. However, the plant conditions during the reactor shutdown differ from conditions during reactor operation at full power (equipment status in protection systems, set points for actuation of safety and protection systems, etc.). Results of RELAP5 simulation of two worst initiating events during reactor shutdown - Pressure Header rupture in case of steam reactor cooldown as well as Pressure Header rupture in case of water reactor cooldown are discussed in the paper. Results of analysis shown that reactor are reliably cooled in both cases. Further analysis for all range of initial events during reactor shutdown and at shutdown conditions is recommended. (author)

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

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)

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....

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

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.

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

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)

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

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)

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

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

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

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.

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)

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)

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)

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

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.

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)

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)

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

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)

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)

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

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

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)

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...

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

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.

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

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…

A CANDU designed for more tolerance to failures in large components

International Nuclear Information System (INIS)

Spinks, N.J.; Barclay, F.W.; Allen, P.J.; Yee, F.

1988-06-01

Current designs of CANDU reactors have several groups of fuel channels each served by an upstream coolant supply-train consisting of an outlet header, a steam generator, one or more pumps in parallel and an inlet header. Postulated failures in these large components put the heaviest demands on the safety systems. For example, the rupture of a header sets the requirements for the speed of shutdown and for the speed and capacity of emergency coolant injection, and it has a large impact on containment design. A CANDU design is being investigated to reduce the impact of failures in large components. Each group of fuel channels is supplied by more than one train so that if one train fails the rest continue to work. Reverse flow limiters reduce the loss-of-coolant from the unbroken trains to a broken supply train. The paper describes several design options for making the piping connections from multi supply-trains to fuel channels. It discusses progress in design and testing of flow limiters. A preliminary analysis is given of affected accidents

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.

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

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

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)

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.

Design of Boiler Welding for Improvement of Lifetime and Cost Control

OpenAIRE

Thong-On, Atcharawadi; Boonruang, Chatdanai

2016-01-01

Fe-2.25Cr-1Mo a widely used material for headers and steam tubes of boilers. Welding of steam tube to header is required for production of boiler. Heat affected zone of the weld can have poor mechanical properties and poor corrosion behavior leading to weld failure. The cost of material used for steam tube and header of boiler should be controlled. This study propose a new materials design for boiler welding to improve the lifetime and cost control, using tungsten inert gas (TIG) welding of F...

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...

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

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)

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.

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

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...

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.)

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.

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

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)

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...

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.

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)

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)

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

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

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.

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)

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

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)

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

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.

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

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

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)

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

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

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...

Dynamic simulation of steam generator failures

Energy Technology Data Exchange (ETDEWEB)

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

1988-07-01

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

Dynamic simulation of steam generator failures

International Nuclear Information System (INIS)

Meister, G.

1988-01-01

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

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.

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

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

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

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

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

Design of Boiler Welding for Improvement of Lifetime and Cost Control.

Science.gov (United States)

Thong-On, Atcharawadi; Boonruang, Chatdanai

2016-11-03

Fe-2.25Cr-1Mo a widely used material for headers and steam tubes of boilers. Welding of steam tube to header is required for production of boiler. Heat affected zone of the weld can have poor mechanical properties and poor corrosion behavior leading to weld failure. The cost of material used for steam tube and header of boiler should be controlled. This study propose a new materials design for boiler welding to improve the lifetime and cost control, using tungsten inert gas (TIG) welding of Fe-2.25Cr-1Mo tube to carbon steel pipe with chromium-containing filler. The cost of production could be reduced by the use of low cost material such as carbon steel pipe for boiler header. The effect of chromium content on corrosion behavior of the weld was greater than that of the microstructure. The lifetime of the welded boiler can be increased by improvement of mechanical properties and corrosion behavior of the heat affected zone.

Design of Boiler Welding for Improvement of Lifetime and Cost Control

Directory of Open Access Journals (Sweden)

Atcharawadi Thong-On

2016-11-01

Full Text Available Fe-2.25Cr-1Mo a widely used material for headers and steam tubes of boilers. Welding of steam tube to header is required for production of boiler. Heat affected zone of the weld can have poor mechanical properties and poor corrosion behavior leading to weld failure. The cost of material used for steam tube and header of boiler should be controlled. This study propose a new materials design for boiler welding to improve the lifetime and cost control, using tungsten inert gas (TIG welding of Fe-2.25Cr-1Mo tube to carbon steel pipe with chromium-containing filler. The cost of production could be reduced by the use of low cost material such as carbon steel pipe for boiler header. The effect of chromium content on corrosion behavior of the weld was greater than that of the microstructure. The lifetime of the welded boiler can be increased by improvement of mechanical properties and corrosion behavior of the heat affected zone.

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)

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

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.)

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

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.)

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

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

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)

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)

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

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.

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

A two-fluid two-phase model for thermal-hydraulic analysis of a U-tube steam generator

International Nuclear Information System (INIS)

Hung, Huanjen; Chieng, Chingchang; Pei, Baushei; Wang, Songfeng

1993-01-01

The Advanced Thermal-Hydraulic Analysis Code for Nuclear Steam Generators (ATHANS) was developed on the basis of the THERMIT-UTSG computer code for U-tube steam generators. The main features of the ATHANS model are as follows: (a) the equations are solved in cylindrical coordinates, (b) the number and the arrangement of the control volumes inside the steam generator can be chosen by the user, (c) the virtual mass effect is incorporated, and (d) the conjugate gradient squared method is employed to accelerate and improve the numerical convergence. The performance of the model is successfully validated by comparison with the test data from a Westinghouse model F steam generator at the Maanshan nuclear power plant. Better agreement with the test data can be obtained by a finer grid system using a cylindrical coordinate system and the virtual mass effect. With these advanced features, ATHANS provides the basic framework for further studies on the problems of steam generators, such as analyses of secondary-side corrosion and tube ruptures

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.

United States Advanced Ultra-Supercritical Component Test Facility for 760°C Steam Power Plants ComTest Project

Energy Technology Data Exchange (ETDEWEB)

Hack, Horst [Electric Power Research Institute (EPRI); Purgert, Robert Michael [Energy Industries of Ohio

2017-12-13

Following the successful completion of a 15-year effort to develop and test materials that would allow coal-fired power plants to be operated at advanced ultra-supercritical (A-USC) steam conditions, a United States-based consortium is presently engaged in a project to build an A-USC component test facility (ComTest). A-USC steam cycles have the potential to improve cycle efficiency, reduce fuel costs, and reduce greenhouse gas emissions. Current development and demonstration efforts are focused on enabling the construction of A-USC plants, operating with steam temperatures as high as 1400°F (760°C) and steam pressures up to 5000 psi (35 MPa), which can potentially increase cycle efficiencies to 47% HHV (higher heating value), or approximately 50% LHV (lower heating value), and reduce CO₂ emissions by roughly 25%, compared to today’s U.S. fleet. A-USC technology provides a lower-cost method to reduce CO₂ emissions, compared to CO₂ capture technologies, while retaining a viable coal option for owners of coal generation assets. Among the goals of the ComTest facility are to validate that components made from advanced nickel-based alloys can operate and perform under A-USC conditions, to accelerate the development of a U.S.-based supply chain for the full complement of A-USC components, and to decrease the uncertainty of cost estimates for future A-USC power plants. The configuration of the ComTest facility would include the key A-USC technology components that were identified for expanded operational testing, including a gas-fired superheater, high-temperature steam piping, steam turbine valve, and cycling header component. Membrane walls in the superheater have been designed to operate at the full temperatures expected in a commercial A-USC boiler, but at a lower (intermediate) operating pressure. This superheater has been designed to increase the temperature of the steam supplied by the host utility boiler up to 1400°F (760�

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.

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

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)

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)

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

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.

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

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.

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...

Cooling facility for reactor container

Energy Technology Data Exchange (ETDEWEB)

Fujimoto, Kiyoshi; Kataoka, Yoshiyuki; Kinoshita, Shoichiro

1996-05-31

A suction port of a condensator to a condensate pipe is connected to a main steam pipe, a discharge port of a incondensible gas exhaustion pipe is connected from an inlet header of the condensator to a main steam pipe by way of a valve, and an exhaustion port of the incondensible gas exhaustion pipe is connected from an exit header of the condensator to a pressure suppression pool by way of a valve. In addition, a condensate return pipe is connected from the exit header of the condensator to the pressure vessel by way of a value. When the reactor is isolated, steams are flown from the pressure vessel to a condensator by way of a main steam pipe. In this case, since incondensible gas is not present, the flow rate of inflown steams is great, the condensate heat conductivity is great and temperature difference between the inside and the outside of the pipes is great, the amount of heat released out of the container is increased. The value of the condensate return pipe is opened, condensates are injected to the pressure vessel. Upon occurrence of an accident, steams and incondensible gases are mixed and flown from the suction pipe of the condensator into the condensator, and noncondensed steams are discharged to a pressure suppression pool by the pressure difference between the inside of the condensate pipe and the inside of the pressure suppression chamber. (N.H.)

Cooling facility for reactor container

International Nuclear Information System (INIS)

Fujimoto, Kiyoshi; Kataoka, Yoshiyuki; Kinoshita, Shoichiro.

1996-01-01

A suction port of a condensator to a condensate pipe is connected to a main steam pipe, a discharge port of a incondensible gas exhaustion pipe is connected from an inlet header of the condensator to a main steam pipe by way of a valve, and an exhaustion port of the incondensible gas exhaustion pipe is connected from an exit header of the condensator to a pressure suppression pool by way of a valve. In addition, a condensate return pipe is connected from the exit header of the condensator to the pressure vessel by way of a value. When the reactor is isolated, steams are flown from the pressure vessel to a condensator by way of a main steam pipe. In this case, since incondensible gas is not present, the flow rate of inflown steams is great, the condensate heat conductivity is great and temperature difference between the inside and the outside of the pipes is great, the amount of heat released out of the container is increased. The value of the condensate return pipe is opened, condensates are injected to the pressure vessel. Upon occurrence of an accident, steams and incondensible gases are mixed and flown from the suction pipe of the condensator into the condensator, and noncondensed steams are discharged to a pressure suppression pool by the pressure difference between the inside of the condensate pipe and the inside of the pressure suppression chamber. (N.H.)

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

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)

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

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

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

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-08-15

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

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.

Experimental investigation of non-condensable gases effect on operation of VVER steam generator in condensation mode

International Nuclear Information System (INIS)

Efanov, A. D.; Kalyakin, S. G.; Morozov, A. V.; Remizov, O. V.; Tsyganok, A. A.; Generalov, V. N.; Berkovich, V. M.; Taranov, G. S.

2008-01-01

To provide the safety in new Russian NPP designs, protection passive systems which don't depend upon human errors are widely used. In terms of safety, the design of NPP of new generation (NPP-2006) falls into the class of advanced NPPs. In the event of an beyond design basis accident with the rupture of the reactor primary circuit and accompanied by the loss of ac sources, the use of passive safety systems are provided for necessary core cooling. Among these is passive heat removal system (PHRS). In the case of leakage in the primary circuit this system ensures the transition of steam generators (SG) to operation in the mode of condensation of the primary circuit steam coming to SG piping from the reactor. As a result, the condensate from steam generators arrives at the core providing its additional cooling. The SG condensation capacity can be adversely affected by the presence of non-condensable gases in the primary circuit of the reactor. Their main sources are nitrogen arriving at the circuit, as hydro accumulators actuate, products of radiolysis of water and air drawn in from the containment through the pipeline rupture. The accumulation of non-condensable gases in SG piping can result in degradation of its condensation capacity to the extent that condensation completely terminates. In this case, the core cooling conditions may be impaired. To experimental investigation of the condensation mode of operation of WER steam generator, a large scale HA2M-SG test rig was constructed at the SSC RF IPPE. The test rig incorporates: buffer tank, equipped by steam supply system; SG model with volumetric-power scale is 1:46; PHRS heat exchanger imitator, cooling by process water. The rig main equipment connected by pipelines and equipped by valves. The elevations of the main equipment correspond to those of reactor project. The rig maximum operating parameters: steam pressure - 1.6 MPa, temperature - 200 Celsius degrees. Experiments at the HA2M-SG test rig have been

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

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

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.

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

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.

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.

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

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

Thermo-mechanical lifetime assessment of components for 700 °C steam turbine applications

International Nuclear Information System (INIS)

Ehrhardt, F.

2014-01-01

In order to increase thermal efficiency, steam turbine technology has been oriented to cover steam inlet temperatures above 700 °C and steam pressures exceeding 350 bar. These temperature levels require the use of nickel and cobalt based alloys. Nickel-based alloys were identified as being suitable for forgeable high-pressure steam turbine rotor materials, including welding procedures for joints between nickel-based alloys and alloyed ferritic steels. Expensive nickel-based alloys should be replaced with conventional heat-resistant steels in applications operating below ∼500-550°C. Since a welded rotor design is favoured, dissimilar metal weldments are required. The research work presented is aimed at the development of thermo-mechanical lifetime assessment methodologies for 700°C steam turbine components. The first main objective was the development of advanced creep-fatigue (CF) lifetime assessment methodologies for the evaluation of Alloy 617 steam turbine rotor features at maximum application temperatures. For the characterisation of the material behaviour under static loading conditions, creep rupture experiments for both medium temperatures and target application temperature have been conducted in order to investigate the influence of ageing treatment on Alloy 617. A creep deformation equation was developed on the basis of a modified Graham-Walles law. Continuous Low Cycle Fatigue (LCF) experiments have been performed. A plasticity model of Chaboche type has been developed. Cyclic/hold experiments have been conducted on Alloy 617. A modification on the creep law was introduced for the description of the material’s decreased creep resistance under combined CF loading. A very promising approach considering plastic and creep-dissipated energy was developed. The effectiveness of this energy exhaustion method was verified with the calculation of endurance curves for continuous cycling LCF and cyclic/hold conditions over a broad range of temperatures, strain

Thermo-mechanical lifetime assessment of components for 700 °C steam turbine applications

Energy Technology Data Exchange (ETDEWEB)

Ehrhardt, F.

2014-07-01

In order to increase thermal efficiency, steam turbine technology has been oriented to cover steam inlet temperatures above 700 °C and steam pressures exceeding 350 bar. These temperature levels require the use of nickel and cobalt based alloys. Nickel-based alloys were identified as being suitable for forgeable high-pressure steam turbine rotor materials, including welding procedures for joints between nickel-based alloys and alloyed ferritic steels. Expensive nickel-based alloys should be replaced with conventional heat-resistant steels in applications operating below ∼500-550°C. Since a welded rotor design is favoured, dissimilar metal weldments are required. The research work presented is aimed at the development of thermo-mechanical lifetime assessment methodologies for 700°C steam turbine components. The first main objective was the development of advanced creep-fatigue (CF) lifetime assessment methodologies for the evaluation of Alloy 617 steam turbine rotor features at maximum application temperatures. For the characterisation of the material behaviour under static loading conditions, creep rupture experiments for both medium temperatures and target application temperature have been conducted in order to investigate the influence of ageing treatment on Alloy 617. A creep deformation equation was developed on the basis of a modified Graham-Walles law. Continuous Low Cycle Fatigue (LCF) experiments have been performed. A plasticity model of Chaboche type has been developed. Cyclic/hold experiments have been conducted on Alloy 617. A modification on the creep law was introduced for the description of the material’s decreased creep resistance under combined CF loading. A very promising approach considering plastic and creep-dissipated energy was developed. The effectiveness of this energy exhaustion method was verified with the calculation of endurance curves for continuous cycling LCF and cyclic/hold conditions over a broad range of temperatures, strain

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.

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

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.)

Simulation of steam generator plugging tubes in a PWR to analyze the operating impact

Energy Technology Data Exchange (ETDEWEB)

Pla, Patricia, E-mail: patricia.pla-freixa@ec.europa.eu [Nuclear Reactor Safety Assessment Unit, Institute for Energy and Transport, Joint Research Centre (JRC) of the European Commission, Petten (Netherlands); Reventos, Francesc, E-mail: francesc.reventos@upc.edu [Technical University of Catalonia (UPC), Barcelona (Spain); Martin Ramos, Manuel, E-mail: manuel.martin-ramos@ec.europa.eu [Nuclear Safety and Security Coordination Unit, Policy Support Coordination, Joint Research Centre of the European Commission, Brussels (Belgium); Sol, Ismael, E-mail: isol@anacnv.com [Asociación Nuclear Ascó-Vandellós-II (ANAV), Tarragona (Spain); Strucic, Miodrag, E-mail: miodrag.strucic@ec.europa.eu [Nuclear Reactor Safety Assessment Unit, Institute for Energy and Transport, Joint Research Centre (JRC) of the European Commission, Petten (Netherlands)

2016-08-15

Highlights: • Plugging a fraction of the SG tubes does not affect power output of the plant. • There is a limit to SG plugging in the range of 10–15%. • The rupture of a SG tube in a 12% plugged SG has shown no significant differences in operator actions. • A SBLOCA in a 12% plugged SG has shown no significant differences in operator actions. - Abstract: A number of nuclear power plants (NPPs) with pressurized water reactors (PWR) in the world have replaced their steam generators (SG) due to degradation of the SG tubes caused by different problems. Several methods were attempted to correct the defects of the tubes, but eventually the only permanent solution was to plug them. The consequences of plugging the tubes are the decrease of heat transfer surface, the reduction of the flow area and subsequent reduction of the primary system mass flow and for a fraction of plugged tubes higher than a given value, the reduction of reactor output and economic losses. The objective of this paper is to analyze whether steam generator tube plugging has an impact in the effectiveness of accident management actions. An analysis with Relap5 Mod 3.3 patch03 for the Spanish reactor Ascó-2, a 3-loop 2940.6 MWth Westinghouse PWR, in which plugging of steam generator tubes are simulated, is presented in order to find the limit for the adequate operation of the plant. Several steady state calculations were performed with different fractions of plugged SG tubes, by modeling the reduction of the primary to secondary heat transfer surface and the reduction of the primary coolant mass flow area in the tubes as well. The results of the analysis yield that plugging 12% of the SG tubes is around the limit for optimal reactor operation. To complete the study two events, in which the steam generators are used to cooldown the plant, were simulated to find out if the plugging of SGs tubes could influence the efficiency of the operator actions described in the emergency operating

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

Aerosol trapping in steam generator (artist): an investigation of aerosol and iodine behaviour in the secondary side of a steam generator

International Nuclear Information System (INIS)

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

2000-01-01

Incidents such as a steam generator tube rupture (SGTR) with stuck-open relief valve are important accident sequences for analysis by virtue of the open path for release of radioactivity which ensues. The release may be mitigated 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 its full import is typically not taken into account in risk assessment studies. The ARTIST experimental programme at PSI 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. Scaling of the break flow represents a particular challenge since the aerosol retention processes operate at contrasting length scales. Preliminary calculations have identified a baseline set of conditions, and confirmed the feasibility of the rig design and scaling principles. Flexibility of the rig layout enables simulations to be performed for a range of SG designs, accident situations and accident management philosophies. (authors)

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

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)

Limit the effects of secondary circuit water or steam piping breaks in the reactor building

International Nuclear Information System (INIS)

Nachev, N.

2001-01-01

The existing design of the WWER-1000 Model 320 does not include provisions against the local mechanical effects of pipe ruptures of the secondary system piping. This situation may lead to accidental effects beyond the design basis of the plant in case of a postulated secondary pipe rupture event. The aim of the present safety enhancement measure is to overcome this safety deficit, that means to carry out some analyses and to suggest protection measures, by which the specified design basis of the plant concerning secondary circuit design basis accidents will be assured. The systems to be considered include the main steam lines (MSL) and the main feedwater lines (MFWL) in the safety related system areas. These areas are the system portions, which are located in the reactor building (containment and room A820 outside the containment). The pipe rupture effects to be considered include the local effects, that means pipe whip impact and jet forces on the adjacent equipment and structures, as well as reaction forces due to blowdown thrust forces and pressure waves in the broken piping system. (author)

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

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

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

Steam turbine cycle

International Nuclear Information System (INIS)

Okuzumi, Naoaki.

1994-01-01

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

Steam 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.)

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.

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

International Nuclear Information System (INIS)

Hori, M.

1980-01-01

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

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...

Fourth and last part in the modelling implementation project of two assimetric cooling systems for ALMOD 3 computer codes

International Nuclear Information System (INIS)

Dominguez, L.; Camargo, C.T.M.

1985-01-01

A two loop simulation capability was developed to the ALMOD 3W2 code through modelling the steam header line connecting the secondary side of steam generators. A brief description of the model is presented and two test cases are shown. Basic code changes are addressed. (Author) [pt

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...

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

Rupture behaviour of nuclear fuel cladding during loss-of-coolant accident

Energy Technology Data Exchange (ETDEWEB)

Suman, Siddharth [Department of Mechanical Engineering, Indian Institute of Technology Patna, Patna 801 103 (India); Khan, Mohd Kaleem, E-mail: mkkhan@iitp.ac.in [Department of Mechanical Engineering, Indian Institute of Technology Patna, Patna 801 103 (India); Pathak, Manabendra [Department of Mechanical Engineering, Indian Institute of Technology Patna, Patna 801 103 (India); Singh, R.N.; Chakravartty, J.K. [Mechanical Metallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)

2016-10-15

Highlights: • Modelling of nuclear fuel cladding during loss-of-coolant accident transient. • Phase transformation, corrosion, and creep combined to evaluate burst criterion. • Effect of oxygen concentration on burst stress and burst strain. • Effect of heating rate, internal pressure fluctuation, shear modulus incorporated. - Abstract: A burst criterion model accounting the simultaneous phenomena of corrosion, solute-strengthening effect of oxygen, oxygen concentration based non-isothermal phase transformation, and thermal creep has been developed to predict the rupture behaviour of zircaloy-4 nuclear fuel cladding during the loss-of-coolant accident transients. The present burst criterion model has been validated using experimental data obtained from single-rod transient burst tests performed in steam environment. The predictions are in good agreement with the experimental results. A detailed computational analysis has been performed to assess the role of different parameters in the rupture of zircaloy cladding during loss-of-coolant accidents. This model reveals that at low temperatures, lower heating rates produce higher burst strains as oxidation effect is nominal. For high temperatures, the lower heating rates produce less burst strains, whereas higher heating rates yield greater burst strains.

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

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)

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 ...

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.)

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

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.

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.

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

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

Nuclear power plant

International Nuclear Information System (INIS)

Nishio, Masahide

1986-01-01

Purpose: To provide a constitution capable of previously and reliably preventing radioactivity from releasing into the atmosphere upon occurrence of main steam pipe rupture accidents in a main steam tunnel chamber. Constitution: The outer circumference at the penetration portion of a nuclear reactor container is tightly closed and the main steam tunnel chamber has a tightly closed vessel structure, which is cooled by a local cooler during normal operation. The main steam tunnel chamber is in communication with a pressure control chamber by way of a release line and a releaf valve is disposed at the midway of the release line. Upon occurrence of rupture accident to the main steam pipes in the main steam tunnel chamber, while steams are issued from the ruptured portion, they are discharged through the release line to the suppression chamber and condensated. As a result, excess pressure in the main steam tunnel can be prevented and when the rupture accident is detected, the main steam isolation valve is closed rapidly to interrupt the steam feeding, whereby the steam released from the ruptured pipeways is stopped to avoid the radioactivity release to the atmosphere. (Kamimura, M.)

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.

Creep Rupture Analysis and Life Estimation of 1.25Cr-0.5Mo, 2.25Cr-1Mo and Modified 9Cr-1Mo Steel: A Comparative Study

Science.gov (United States)

Roy, Prabir Kumar

2018-04-01

This paper highlights a comparative assessment of creep life of 1.25Cr-0.5Mo, 2.25Cr-1Mo and modified 9Cr-1Mo steels based on accelerated creep rupture tests. Creep rupture test data have been analysed and creep life of the above mentioned materials have been assessed using Larson Miller parameter at the stress levels of 60 and 42 MPa for different temperatures. Limiting steam temperatures for minimum design life of 105 h at 42 and 60 MPa for the above mentioned steels have also been calculated. Microstructural studies for the three above mentioned steels are also done.

Steam generator tube support plate degradation in French plants: maintenance strategy

International Nuclear Information System (INIS)

Gauchet, J.-P.; Gillet, N.; Stindel, M.

1998-01-01

This paper reports on the degradations of Steam Generator (SG) Tube Support Plates (TSPs) observed in French plants and the maintenance strategy adopted to continue operating the plant without any decrease of the required safety level. Only drilled carbon steel TSPs of early SGs are affected. Except the particular damage of the TSP8 of FESSENHEIM 2 caused by chemical cleaning procedures implemented in 1992, two main problems were observed almost exclusively on the upper TSP: Ligaments ruptured near the aseismic block located at 215 degrees. This degradation is perfectly detectable by bobbin coil inspection. It occurs very early in the life of the SG as can be seen from the records of previous inspections and no evolution of the signals was observed. This damage can be detected for 51M model SGs on several sites; Wastage of the ligaments resulting in enlargement of flow holes with in some cases complete consumption of a ligament. This damage was only observed for SGs of at GRAVELINES. This damage evolved cycle after cycle. Detailed studies were performed to analyze tubing behavior when a tube is not supported by the upper TSP because of missing ligaments. These studies evaluated the risk of vibratory instability, the behavior of both the TSP and the tubing in case of a seismic event or a LOCA and finally the behavior of the TSP in case of a Steam Line Break. Concerning vibratory instability it was possible to define zones where stability could not be demonstrated. Dampine, cables and sentinel plugs were then used when necessary to eliminate the risk of Steam Generator Tube Rupture (SGTR). For accidental conditions, it could be shown that no unacceptable damage occurs and that the core cooling function of the SG is always maintained if some tubes are plugged. From this analysis, It was possible to define the inspection programs for the different plants taking into account the specific situation of each plant regarding the damages detected. These programs include

Life prediction of steam generator tubing due to stress corrosion crack using Monte Carlo Simulation

International Nuclear Information System (INIS)

Hu Jun; Liu Fei; Cheng Guangxu; Zhang Zaoxiao

2011-01-01

Highlights: → A life prediction model for SG tubing was proposed. → The initial crack length for SCC was determined. → Two failure modes called rupture mode and leak mode were considered. → A probabilistic life prediction code based on Monte Carlo method was developed. - Abstract: The failure of steam generator tubing is one of the main accidents that seriously affects the availability and safety of a nuclear power plant. In order to estimate the probability of the failure, a probabilistic model was established to predict the whole life-span and residual life of steam generator (SG) tubing. The failure investigated was stress corrosion cracking (SCC) after the generation of one through-wall axial crack. Two failure modes called rupture mode and leak mode based on probabilistic fracture mechanics were considered in this proposed model. It took into account the variance in tube geometry and material properties, and the variance in residual stresses and operating conditions, all of which govern the propagations of cracks. The proposed model was numerically calculated by using Monte Carlo Simulation (MCS). The plugging criteria were first verified and then the whole life-span and residual life of the SG tubing were obtained. Finally, important sensitivity analysis was also carried out to identify the most important parameters affecting the life of SG tubing. The results will be useful in developing optimum strategies for life-cycle management of the feedwater system in nuclear power plants.

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.

Vibrations as a contributor to the cracking of PGV-1000 steam generator cold collector

International Nuclear Information System (INIS)

Verezemskij, V.G.

1997-01-01

The results of multiple investigations of cold collector ruptures at NPPs with WWER-1000 reactor as a complex and multi-parameter process are generalized. It is shown that the temperature of cold collector operation (280-290 deg C) at which environment corrosion effects are maximum has an important role for revealing the factors causing the damage. When the reactor plant operates under stationary and transient conditions the primary coolant circuit equipment, main circulation pipelines and main steam lines become involved into vibrations with different intensities as a result of pressure pulsations at reverse and multiple frequencies of the main circulation pumps connected with formation of standing pressure waves in the primary circuit and steam lines. The analysis made gives an opportunity to conclude that dynamic loads (vibrations) play the leading part in reaching the limits of cold collector metal cyclic strength and its cracking. It means that the measures for cold collector service life increasing should be directed on vibration amplitude lowering and cyclic stress decreasing

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.

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.

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

An Improved Steam Injection Model with the Consideration of Steam Override

Directory of Open Access Journals (Sweden)

He Congge

2017-01-01

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

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

Science.gov (United States)

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

2017-02-01

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

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)

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

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

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...

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

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

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.

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)

X-rays diffraction characterization of corrosion products transported by secondary side of a CANDU NPP

International Nuclear Information System (INIS)

Dinu, A.; Tunaru, M.; Velciu, L.

2016-01-01

To verify the chemistry of secondary side of CANDU steam generators, Millipore filters are used to sampling from condensing extraction pump, from feed water header and blow down of steam generator. These filters retain the corrosion products as very fine particles and are used as samples in chemistry water control. X-Ray diffraction technique is the able to distinguish the different crystallographic compounds present in oxide films deposited on the Millipore filters and gives information referring to the nature of corrosion products transported in secondary side. The XRD analysis has identified the following substance in deposited layer: magnetite (Fe_3O_4), hematite (Fe_2O_3), and iron oxide hydroxide (FeOOH). By optical microscopy it was observed a brown-reddish background specific to hematite and iron oxide hydroxide, especially for filters extracted from condensing extraction pump. The black colour of crud present on filters extracted from feed water header and blow down of steam generator shows the presence of magnetite. (authors)

Stress analysis for CANDU reactor structure assembly following a postulated p/t, c/t rupture after flow blockage

Energy Technology Data Exchange (ETDEWEB)

Soliman, S A; Lee, T; Ibrahim, A M; Hodgson, S [Atomic Energy of Canada Ltd., Saskatoon, SK (Canada)

1996-12-31

This paper describes the collapse load calculations for the reactor structure assembly under the postulated fuel channel flow blockage Level D (faulted) loading condition. Under the flow blockage condition, the primary coolant flow path is obstructed between the inlet and outlet feeder connections to the headers. This, in turn, is postulated to cause the pressure tube and calandria tube to rupture and release hot molten fuel into the moderator, producing a hydrodynamic transient within the calandria shell. The most severe hydrodynamic loads occur within a fraction of a second (0.14 second). The peak pressure for the limiting case scenario for Level D condition is 120 psig, due to a single channel failure event. Under this accident condition, it is shown that the reactor structure assembly can withstand the pressure transient and the structural integrity of the core is assured. A finite element model is generated and used to calculate the minimum collapse load. The ANSYS code is used with element type Stif-43 for elastic/plastic, large deformation and small strain analysis. (author). 1 ref., 3 tabs., 9 figs.

Stress analysis for CANDU reactor structure assembly following a postulated p/t, c/t rupture after flow blockage

International Nuclear Information System (INIS)

Soliman, S.A.; Lee, T.; Ibrahim, A.M.; Hodgson, S.

1995-01-01

This paper describes the collapse load calculations for the reactor structure assembly under the postulated fuel channel flow blockage Level D (faulted) loading condition. Under the flow blockage condition, the primary coolant flow path is obstructed between the inlet and outlet feeder connections to the headers. This, in turn, is postulated to cause the pressure tube and calandria tube to rupture and release hot molten fuel into the moderator, producing a hydrodynamic transient within the calandria shell. The most severe hydrodynamic loads occur within a fraction of a second (0.14 second). The peak pressure for the limiting case scenario for Level D condition is 120 psig, due to a single channel failure event. Under this accident condition, it is shown that the reactor structure assembly can withstand the pressure transient and the structural integrity of the core is assured. A finite element model is generated and used to calculate the minimum collapse load. The ANSYS code is used with element type Stif-43 for elastic/plastic, large deformation and small strain analysis. (author). 1 ref., 3 tabs., 9 figs

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.

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

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

International Nuclear Information System (INIS)

Novotny, J.; Novotny, J. Jr.

2003-01-01

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

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

Procedure for generating steam and steam generator for operating said procedure

International Nuclear Information System (INIS)

Chlique, Bernard.

1975-01-01

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

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

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

Steam generator tube integrity program: Annual report, August 1995--September 1996. Volume 2

International Nuclear Information System (INIS)

Diercks, D.R.; Bakhtiari, S.; Kasza, K.E.; Kupperman, D.S.; Majumdar, S.; Park, J.Y.; Shack, W.J.

1998-02-01

This report summarizes work performed by Argonne National Laboratory on the Steam Generator Tube Integrity Program from the inception of the program in August 1995 through September 1996. The program is divided into five tasks: (1) assessment of inspection reliability, (2) research on ISI (inservice-inspection) technology, (3) research on degradation modes and integrity, (4) tube removals from steam generators, and (5) program management. Under Task 1, progress is reported on the preparation of facilities and evaluation of nondestructive evaluation techniques for inspecting a mock-up steam generator for round-robin testing, the development of better ways to correlate failure pressure and leak rate with eddy current (EC) signals, the inspection of sleeved tubes, workshop and training activities, and the evaluation of emerging NDE technology. Results are reported in Task 2 on closed-form solutions and finite-element electromagnetic modeling of EC probe responses for various probe designs and flaw characteristics. In Task 3, facilities are being designed and built for the production of cracked tubes under aggressive and near-prototypical conditions and for the testing of flawed and unflawed tubes under normal operating, accident, and severe-accident conditions. Crack behavior and stability are also being modeled to provide guidance for test facility design, develop an improved understanding of the expected rupture behavior of tubes with circumferential cracks, and predict the behavior of flawed and unflawed tubes under severe accident conditions. Task 4 is concerned with the acquisition of tubes and tube sections from retired steam generators for use in the other research tasks. Progress on the acquisition of tubes from the Salem and McGuire 1 nuclear plants is reported

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

International Nuclear Information System (INIS)

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

1996-01-01

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

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

International Nuclear Information System (INIS)

Bessouat, Roger; Marjollet, Jacques.

1976-01-01

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

RELAP5 Prediction of Transient Tests in the RD-14 Test Facility

International Nuclear Information System (INIS)

Lee, Sukho; Kim, Manwoong; Kim, Hho-Jung; Lee, John C.

2005-01-01

Although the RELAP5 computer code has been developed for best-estimate transient simulation of a pressurized water reactor and its associated systems, it could not assess the thermal-hydraulic behavior of a Canada deuterium uranium (CANDU) reactor adequately. However, some studies have been initiated to explore the applicability for simulating a large-break loss-of-coolant accident in CANDU reactors. In the present study, the small-reactor inlet header break test and the steam generator secondary-side depressurization test conducted in the RD-14 test facility were simulated with the RELAP5/MOD3.2.2 code to examine its extended capability for all the postulated transients and accidents in CANDU reactors. The results were compared with experimental data and those of the CATHENA code performed by Atomic Energy of Canada Limited.In the RELAP5 analyses, the heated sections in the facility were simulated as a multichannel with five pipe models, which have identical flow areas and hydraulic elevations, as well as a single-pipe model.The results of the small-reactor inlet header break and the steam generator secondary-side depressurization simulations predicted experimental data reasonably well. However, some discrepancies in the depressurization of the primary heat transport system after the header break and consequent time delay of the major phenomena were observed in the simulation of the small-reactor inlet header break test

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.

Analysis of an accident with the main circulation tube rupture at the WWER-1000

International Nuclear Information System (INIS)

Boyadzhiev, A.I.; Stefanova, S.J.

1984-01-01

In connection with the forthcoming construction of a npp with the wwer-1000 reactor the loss of coolant accident associated with the main circulation tube rupture at the inlet near the reactor is analyzed. The relap4/mod6 program is used for the analysis. The data obtained show that the coolant outflow stage continues for about 25s. On the average the pressure in the circuits varies from 16 to 10 mpa per 0.1s and then it continues to decrease slowly. The pressure in the steam generator at the secondary circuits end increases approximately up to 6.9 MPa as a result of steam generator blocking and remaining coolant heating and then somewhat decreases owing to the primary circuit cooling. By the end of the fuel and can temperatures are equalized and the heat transfer coefficient is stabilized at the level of 100 w/1 (m 2 xK). It is concluded that during a loss of coolant accident at the wwer-1000 reactor in procesess of coolant blowdown in the medium power fuel elemets neither the fuel, melting temperature (3000 k), nor the critical temperature (1000 k) of plastic deformation zirconiu can initiation are attained

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.

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

International Nuclear Information System (INIS)

Ding Xunshen; Cui Baoyuan; Xue Yunkui; Liu Shixun

1989-01-01

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

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

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

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

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

International Nuclear Information System (INIS)

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

1999-12-01

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

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.)

Acoustic signal emission monitoring as a novel method to predict steam pops during radiofrequency ablation: preliminary observations.

Science.gov (United States)

Chik, William W B; Kosobrodov, Roman; Bhaskaran, Abhishek; Barry, Michael Anthony Tony; Nguyen, Doan Trang; Pouliopoulos, Jim; Byth, Karen; Sivagangabalan, Gopal; Thomas, Stuart P; Ross, David L; McEwan, Alistair; Kovoor, Pramesh; Thiagalingam, Aravinda

2015-04-01

Steam pop is an explosive rupture of cardiac tissue caused by tissue overheating above 100 °C, resulting in steam formation, predisposing to serious complications associated with radiofrequency (RF) ablations. However, there are currently no reliable techniques to predict the occurrence of steam pops. We propose the utility of acoustic signals emitted during RF ablation as a novel method to predict steam pop formation and potentially prevent serious complications. Radiofrequency generator parameters (power, impedance, and temperature) were temporally recorded during ablations performed in an in vitro bovine myocardial model. The acoustic system consisted of HTI-96-min hydrophone, microphone preamplifier, and sound card connected to a laptop computer. The hydrophone has the frequency range of 2 Hz to 30 kHz and nominal sensitivity in the range -240 to -165 dB. The sound was sampled at 96 kHz with 24-bit resolution. Output signal from the hydrophone was fed into the camera audio input to synchronize the video stream. An automated system was developed for the detection and analysis of acoustic events. Nine steam pops were observed. Three distinct sounds were identified as warning signals, each indicating rapid steam formation and its release from tissue. These sounds had a broad frequency range up to 6 kHz with several spectral peaks around 2-3 kHz. Subjectively, these warning signals were perceived as separate loud clicks, a quick succession of clicks, or continuous squeaking noise. Characteristic acoustic signals were identified preceding 80% of pops occurrence. Six cardiologists were able to identify 65% of acoustic signals accurately preceding the pop. An automated system identified the characteristic warning signals in 85% of cases. The mean time from the first acoustic signal to pop occurrence was 46 ± 20 seconds. The automated system had 72.7% sensitivity and 88.9% specificity for predicting pops. Easily identifiable characteristic acoustic emissions

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

Analysis of MSGTR events for APR1400 by means of best estimate thermal-hydraulic system code

International Nuclear Information System (INIS)

Jeong, Ji Hwan; Kim, Sang Jae; Chang, Keun Sun; Lee, Jae Hun

2001-01-01

A multiple steam generator tube rupture (MSGTR) event has never occurred in the history of commercial nuclear reactor operation while single steam generator tube rupture (SGTR) event is reported to occur every two years. As there is no history of MSGTR event, the understandings of transients and consequences of this event are not so much. In this study, a postulated MSGTR event in advanced power reactor 1400 (APR1400) is analyzed using thermal-hydraulic system code. The APR 1400 is a two-loop, 1000 MWe, PWR supposed to be built in 2009. MARS1.4 is used in this study. The present study aims to understand the effects of rupture location in heat transfer tubes and selection of affected steam generator following a MSGTR event. The effects of five tube rupture locations are compared with each other. The comparison shows that the response of APR1400 is to allow shortest time for operator action following a tubes rupture in the vicinity of hot-leg side tube sheet and to allow longest time following a tube ruptures at the tube top. The MSSV lift time for rupture at tube-top is evaluated as 24.5% larger than that for rupture at hot-leg side tube sheet. Also, the MSSV lift time for four cases are compared in order to examine how long operator response time is allowed depending on which steam generator is affected. The comparison shows that the cases for both of two steam generators are affected allow longer time for operator action compared with the cases that a single steam generator is affected. Further more, the tube ruptures in the steam generator where a pressurizer is linked leads to the shortest operator response time

Rupture of the Pitáycachi Fault in the 1887 Mw 7.5 Sonora, Mexico earthquake (southern Basin-and-Range Province): Rupture kinematics and epicenter inferred from rupture branching patterns

Science.gov (United States)

Suter, Max

2015-01-01

During the 3 May 1887 Mw 7.5 Sonora earthquake (surface rupture end-to-end length: 101.8 km), an array of three north-south striking Basin-and-Range Province faults (from north to south Pitáycachi, Teras, and Otates) slipped sequentially along the western margin of the Sierra Madre Occidental Plateau. This detailed field survey of the 1887 earthquake rupture zone along the Pitáycachi fault includes mapping the rupture scarp and measurements of surface deformation. The surface rupture has an endpoint-to-endpoint length of ≥41.0 km, dips 70°W, and is characterized by normal left-lateral extension. The maximum surface offset is 487 cm and the mean offset 260 cm. The rupture trace shows a complex pattern of second-order segmentation. However, this segmentation is not expressed in the 1887 along-rupture surface offset profile, which indicates that the secondary segments are linked at depth into a single coherent fault surface. The Pitáycachi surface rupture shows a well-developed bipolar branching pattern suggesting that the rupture originated in its central part, where the polarity of the rupture bifurcations changes. Most likely the rupture first propagated bilaterally along the Pitáycachi fault. The southern rupture front likely jumped across a step over to the Teras fault and from there across a major relay zone to the Otates fault. Branching probably resulted from the lateral propagation of the rupture after breaching the seismogenic part of the crust, given that the much shorter ruptures of the Otates and Teras segments did not develop branches.

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).

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.

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

Steam generator with perfected dryers

International Nuclear Information System (INIS)

Fenet, J.C.

1987-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2009-07-01

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

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

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.

Analytical studies on optimization of containment design pressure

International Nuclear Information System (INIS)

Haware, S.K.; Ghosh, A.K.; Kushwaha, H.S.

2005-01-01

The containment of the proposed Advanced Heavy Water Reactor (AHWR) is divided into two main volumes viz. V1 and V2 interconnected by vent system via suppression pool. The arrangement is such that the volume V2 surrounds the volume V1 (see Fig.1). Blow Out Panels (BOPs), installed on volume V1 are designed to rupture at a differential pressure of 50 kPa. The containment was analysed using the in-house developed code CONTRAN, for three different scenario considered viz. (i) Loss of Coolant Accident (LOCA) involving double ended break in the downcomer pipe, (ii) LOCA involving double ended break in the reactor inlet header and (iii) Main Steam Line Break (MSLB) Accident. It was revealed that the accident involving the double-ended break of reactor inlet header results in the maximum value of the containment peak pressure. Results of the analyses indicated that the size of the BOP has bearing on the containment peak pressure. Therefore, five cases were analysed, varying the size of BOP from 0 to 10 m 2 , in order to quantify the influence of the size of BOP on the containment peak pressure. The blowdown mass and energy discharge data calculated using the code RELAP5/MOD3.2 was used in the analysis. It was observed that the vents are cleared in around 0.41 seconds into the accident. The containment peak pressures obtained in various cases are presented in Fig.2. The containment peak pressure varies with the size of BOP and passes through minima for a BOP size of around 5 m 2 . There are two flow processes, competing with each other viz. the steam-air mixture passage through the vent system via suppression pool and direct passage of steam air mixture through BOP bypassing the suppression pool. Though the energy suppression efficiency of the suppression pool decreases with increasing size of BOP, the pressure suppression efficiency was found to be maximum at around 5 m 2 size of BOP. The containment peak pressure passing through minima indicates that there is a scope for

Steam Digest: Volume IV

Energy Technology Data Exchange (ETDEWEB)

2004-07-01

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

Steam Digest Volume IV

Energy Technology Data Exchange (ETDEWEB)

None

2004-07-01

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

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

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)

Materials for Advanced Ultrasupercritical Steam Turbines Task 4: Cast Superalloy Development

Energy Technology Data Exchange (ETDEWEB)

Thangirala, Mani

2015-09-30

The Steam Turbine critical stationary structural components are high integrity Large Shell and Valve Casing heavy section Castings, containing high temperature steam under high pressures. Hence to support the development of advanced materials technology for use in an AUSC steam turbine capable of operating with steam conditions of 760°C (1400°F) and 35 Mpa (5000 psia), Casting alloy selection and evaluation of mechanical, metallurgical properties and castability with robust manufacturing methods are mandated. Alloy down select from Phase 1 based on producability criteria and creep rupture properties tested by NETL-Albany and ORNL directed the consortium to investigate cast properties of Haynes 282 and Haynes 263. The goals of Task 4 in Phase 2 are to understand a broader range of mechanical properties, the impact of manufacturing variables on those properties. Scale up the size of heats to production levels to facilitate the understanding of the impact of heat and component weight, on metallurgical and mechanical behavior. GE Power & Water Materials and Processes Engineering for the Phase 2, Task 4.0 Castings work, systematically designed and executed casting material property evaluation, multiple test programs. Starting from 15 lbs. cylinder castings to world’s first 17,000 lbs. poured weight, heavy section large steam turbine partial valve Haynes 282 super alloy casting. This has demonstrated scalability of the material for steam Turbine applications. Activities under Task 4.0, Investigated and characterized various mechanical properties of Cast Haynes 282 and Cast Nimonic 263. The development stages involved were: 1) Small Cast Evaluation: 4 inch diam. Haynes 282 and Nimonic 263 Cylinders. This provided effects of liquidus super heat range and first baseline mechanical data on cast versions of conventional vacuum re-melted and forged Ni based super alloys. 2) Step block castings of 300 lbs. and 600 lbs. Haynes 282 from 2 foundry heats were evaluated which

Review of damages of nuclear power plants steam generator's tubes and way of detecting by using eddy current method

International Nuclear Information System (INIS)

Stanic, D.

1996-01-01

Steam generator tubing integrity is very important factor for reliable and safe operation of NPP. Several different types of tube degradation mechanisms were experienced in SG operation. To avoid possible tube rupture and primary-to-secondary leak, the EC examination of tubing should be performed. Different eddy current techniques may be used for detecting defects and theirs characterization. A comparison of data analysis results with pulled tube destructive metallography results can provide valuable insights in determining the capability of existing technology and provide guidance for procedure or technology improvements. (author)

Selling steam

International Nuclear Information System (INIS)

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

1991-01-01

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

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.

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

International Nuclear Information System (INIS)

Teichel, H.

1978-01-01

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

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.

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

International Nuclear Information System (INIS)

1988-09-01

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

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

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.

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.)

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.

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

1992-08-01

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

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...

Sodium and steam leak simulation studies for fluidized bed steam generators

International Nuclear Information System (INIS)

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

1976-01-01

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

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

Science.gov (United States)

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

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

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

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)

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)

Horizontal steam generator thermal-hydraulics

Energy Technology Data Exchange (ETDEWEB)

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

1995-09-01

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

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

Science.gov (United States)

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

2014-09-01

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

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

International Nuclear Information System (INIS)

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

1977-01-01

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

Steam generators - problems and prognosis

International Nuclear Information System (INIS)

Tapping, R.L.

1997-05-01

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

Steam 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)

Inservice Inspection in the Fugen

OpenAIRE

1983-01-01

Pressure tube type reactors have specific components orstructures,compared with light water reactors.They are(1) steam drums(2) reactor inlet headers(3) reactor inlet and outlet pipies(4) pressure tubes.Much attention is paid upon Inservice Inspection (ISI)of the above components.

Steam power plant

International Nuclear Information System (INIS)

Campbell, J.W.E.

1981-01-01

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

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

International Nuclear Information System (INIS)

Kawamura, H.; Meister, G.

1982-10-01

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

Steam turbines for the future

International Nuclear Information System (INIS)

Trassl, W.

1988-01-01

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

The Invisibility of Steam

Science.gov (United States)

Greenslade, Thomas B., Jr.

2014-01-01

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

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

International Nuclear Information System (INIS)

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

1997-05-01

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

Evaluation of maintenance strategies for steam generator tubes in pressurized waster reactors. 2. Cost and profitability analyses

International Nuclear Information System (INIS)

Isobe, Y.; Sagisaka, M.; Yoshimura, S.; Yagawa, G.

2000-01-01

As an application of probabilistic fracture mechanics (PFM), risk-benefit analysis was carried out to evaluate maintenance activities of steam generator (SG) tubes used in pressurized water reactors (PWRs). The analysis was conducted for SG tubes made of Inconel 600, and Inconel 690 as well assuming its crack initiation and crack propagation law based on Inconel 600 data. The following results were drawn from the analysis. Improvement of inspection accuracy reduces the maintenance costs significantly and is preferable from the viewpoint of profitability due to reduction of SG tube leakage and rupture. There is a certain region of SCC properties of SG tubes where sampling inspection is effective. (author)

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.

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)

Rupture disc opening property for using pipe rupture test in JAERI

International Nuclear Information System (INIS)

Kato, Rokuro

1983-03-01

In the Mechanical Strength and Structure Lab of JAERI there are being performed pipe break tests which are a postulated instantaneous guillotine break of the primary coolant piping in nuclear power plants. The test being performed are pipe whip tests and jet discharging tests. The bursting of the rupture disc is initiated by an electrical arc and is concluded by the internal pressure. Because the time characteristics during the opening of the rupture disc affects the dynamic thrust force of the pipe, it is necessary to measure these time characteristics. However, it is difficult to measure the conditions during this continuous opening because at the same time of the opening the high temperature and high pressure water is flashing. Therefore, the rupture disc opening was postulated on the measuring of the effective opening characteristics with electric contraction terminals which were attached to the inner surface of the test pipe downstream of the rupture disc and were extended toward the pipe centerline in a ring whose area is about 60 % of the area of the pipe flow sectional area. The measurement voltage was recorded when the data recorder was started in sequence with the electrical arc release from a trigger signal. As a result, it is evident that under high temperature and high pressure water the effective opening time is delayed by a few milliseconds. (author)

Steam-water separator

International Nuclear Information System (INIS)

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

1978-01-01

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

EPRI steam generator programs

International Nuclear Information System (INIS)

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

1977-01-01

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

Experimental and numerical investigation of the flow measurement method utilized in the steam generator of HTR-PM

Energy Technology Data Exchange (ETDEWEB)

Wang, Shiming; Ren, Cheng; Sun, Yangfei [Institute of Nuclear and New Energy Technology of Tsinghua University, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Beijing 100084 (China); Tu, Jiyuan [Institute of Nuclear and New Energy Technology of Tsinghua University, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Beijing 100084 (China); School of Aerospace, Mechanical & Manufacturing Engineering, RMIT University, Melbourne, VIC 3083 (Australia); Yang, Xingtuan, E-mail: yangxt107@sina.com [Institute of Nuclear and New Energy Technology of Tsinghua University, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Beijing 100084 (China)

2016-08-15

Highlights: • The flow confluence process in the steam generator is very important for HTR-PM. • The complicated flow in the unique pipeline configuration is studied by both of experimental and numerical method. • The pressure uniformity at the bottom of the model was tested to evaluate the accuracy of the experimental results. • Flow separation and the secondary flow is described for explaining the nonuniformity of the flow distribution. - Abstract: The helium flow measurement method is very important for the design of HTR-PM. Water experiments and numerical simulation with a 1/5 scaled model are conducted to investigate the flow measurement method utilized in the steam generator of HTR-PM. Pressure information at specific location of the 90° elbows with the diameter of 46.75 mm and radius ratio of 1.5 is measured to evaluate the flow rate in the riser-pipes. Pressure uniformity at the bottom of the experimental apparatus is tested to evaluate the influence of the equipment error on the final experimental results. Numerical results obtained by using the realizable k–ε model are compared with the experimental data. The results reveal that flow oscillation does not occur in the confluence system. For every single riser-pipe, the flow is stable despite the nonuniformity of the flow distribution. The average flow rates of the two pipe series show good repeatability regardless of the increases and decreases of the average velocity. In the header box, the flows out of the riser-pipes encounter with each other and finally distort the pressure distribution and the nonuniformity of the flow distribution becomes more significant along with the increasing Reynolds number.

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.

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)

Rupture of Model 48Y UF6 cylinder and release of uranium hexafluoride, Sequoyah Fuels Facility, Gore, Oklahoma, January 4, 1986. Volume 1

International Nuclear Information System (INIS)

1986-02-01

At 11:30 a.m. on January 4, 1986, a Model 48Y UF 6 cylinder filled with uranium hexafluoride (UF 6 ) ruptured while it was being heated in a steam chest at the Sequoyah Fuels Conversion Facility near Gore, Oklahoma. One worker died because he inhaled hydrogen fluoride fumes, a reaction product of UF 6 and airborne moisture. Several other workers were injured by the fumes, but none seriously. Much of the facility complex and some offsite areas to the south were contaminated with hydrogen fluoride and a second reaction product, uranyl fluoride. The interval of release was approximately 40 minutes. The cylinder, which had been overfilled, ruptured while it was being heated because of the expansion of UF 6 as it changed from the solid to the liquid phase. The maximum safe capacity for the cylinder is 27,560 pounds of product. Evidence indicates that it was filled with an amount exceeding this limit. 18 figs

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.

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.)

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

CSIR Research Space (South Africa)

Price, T

2010-08-01

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

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.)

Long-term results after repair of ruptured and non-ruptured abdominal aortic aneurysm

Directory of Open Access Journals (Sweden)

Kuzmanović Ilija B.

2004-01-01

Full Text Available INTRODUCTION Abdominal aortic aneurysm can be repaired by elective procedure while asymptomatic, or immediately when it is complicated - mostly due to rupture. Treating abdominal aneurysm electively, before it becomes urgent, has medical and economical reason. Today, the first month mortality after elective operations of the abdominal aorta aneurysm is less than 3%; on the other hand, significant mortality (25%-70% has been recorded in patients operated immediately because of rupture of the abdominal aneurysm. In addition, the costs of elective surgical treatment are significantly lower. OBJECTIVE The objective of this study is to compare long-term survival of patients that underwent elective or immediate repair of abdominal aortic aneurysm (due to rupture, and to find out the factors influencing the long-term survival of these patients. MATERIAL AND METHODS Through retrospective review of prospectively collected data of the Institute for Cardiovascular Diseases of Clinical Center of Serbia, Belgrade, 56 patients that had elective surgery and 35 patients that underwent urgent operation due to rupture of abdominal aneurysm were followed up. Only the patients that survived 30 postoperative days were included in this review, and were followed up (ranging from 2 to 126 months. Electively operated patients were followed during 58.82 months on the average (range 7 to 122, and urgently operated were followed over 52.26 months (range 2 to 126. There was no significant difference of the length of postoperative follow-up between these two groups. RESULTS During this period, out of electively operated and immediately operated patients, 27 and 22 cases died, respectively. There was no significant difference (p>0,05a of long-term survival between these two groups. Obesity and early postoperative complications significantly decreased long-term survival of both electively and immediately operated patients. Graft infection, ventral hernia, aneurysm of

Steam purity in PWRs

International Nuclear Information System (INIS)

Hopkinson, J.

1982-01-01

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

Wet-steam erosion of steam turbine disks and shafts

International Nuclear Information System (INIS)

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

2011-01-01

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

Simulation of heat and mass transfer processes in the experimental section of the air-condensing unit of Scientific Production Company "Turbocon"

Science.gov (United States)

Artemov, V. I.; Minko, K. B.; Yan'kov, G. G.; Kiryukhin, A. V.

2016-05-01

A mathematical model was developed to be used for numerical analysis of heat and mass transfer processes in the experimental section of the air condenser (ESAC) created in the Scientific Production Company (SPC) "Turbocon" and mounted on the territory of the All-Russia Thermal Engineering Institute. The simulations were performed using the author's CFD code ANES. The verification of the models was carried out involving the experimental data obtained in the tests of ESAC. The operational capability of the proposed models to calculate the processes in steam-air mixture and cooling air and algorithms to take into account the maldistribution in the various rows of tube bundle was shown. Data on the influence of temperature and flow rate of the cooling air on the pressure in the upper header of ESAC, effective heat transfer coefficient, steam flow distribution by tube rows, and the dimensions of the ineffectively operating zones of tube bundle for two schemes of steam-air mixture flow (one-pass and two-pass ones) were presented. It was shown that the pressure behind the turbine (in the upper header) increases significantly at increase of the steam flow rate and reduction of the flow rate of cooling air and its temperature rise, and the maximum value of heat transfer coefficient is fully determined by the flow rate of cooling air. Furthermore, the steam flow rate corresponding to the maximum value of heat transfer coefficient substantially depends on the ambient temperature. The analysis of the effectiveness of the considered schemes of internal coolant flow was carried out, which showed that the two-pass scheme is more effective because it provides lower pressure in the upper header, despite the fact that its hydraulic resistance at fixed flow rate of steam-air mixture is considerably higher than at using the one-pass schema. This result is a consequence of the fact that, in the two-pass scheme, the condensation process involves the larger internal surface of tubes

On the separation and association of vapour in large power stations; Ueber die Trennung und Vereinigung von Dampf in Grosskraftwerken

Energy Technology Data Exchange (ETDEWEB)

Matschke, Alan

2012-07-01

The main objectives for the improvement of utility boilers are to increase their overall efficiency. Therefore, steam parameters need to be increased leading to temperatures which may exceed the limiting temperatures of the tube materials. Considering these aspects, the single tube temperatures of a powerplant have to be determined to prevent material failure. In order to account for the mutual impact of combustion and water-steam-cycle, a coupled simulation is performed to calculate the single tube temperatures. Apart from the interaction of combustion and water-steam-cycle, the design of the heat exchanger headers is relevant. On the basis of a detailed investigation of the pressure distribution in steam distributors and headers, a new quality regarding the prediction and calculation of single tube temperatures will be obtained. By means of a similarity analysis, the governing geometry relations associated with the pressure change in steam distributors and headers are identified. The parametrisation of this approach is done by comparing the measured pressure values of an air-flow divider-collector-combination. The portability of the air-flow results to water, respectively steam is approved. Based on the experimental data, computational fluid dynamic simulations of all governing geometrical relations are performed. The validity of the similarity based approach is shown for nearly all geometrical relations corresponding to real utility boiler application. Deviations between experiment and calculations based on the similarity approach are only evident for very short distances of separating tubes originating from secondary vortices. However, the impact of these vortices to the global mass distribution is minimal. The comparison between measured and calculated single tube temperatures from a lignite-fired utility boiler shows the good match between the simulated and measured temperatures. The single tube temperatures of a heat exchanger across the boiler width are

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

Science.gov (United States)

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

2017-12-01

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

Liver Hydatid Cyst with Transdiaphragmatic Rupture and Lung Hydatid Cyst Ruptured into Bronchi and Pleural Space

International Nuclear Information System (INIS)

Arıbaş, Bilgin Kadri; Dingil, Gürbüz; Köroğlu, Mert; Üngül, Ümit; Zaralı, Aliye Ceylan

2011-01-01

The aim of this case study is to present effectiveness of percutaneous drainage as a treatment option of ruptured lung and liver hydatid cysts. A 65-year-old male patient was admitted with complicated liver and lung hydatid cysts. A liver hydatid cyst had ruptured transdiaphragmatically, and a lung hydatid cyst had ruptured both into bronchi and pleural space. The patient could not undergo surgery because of decreased respiratory function. Both cysts were drained percutaneously using oral albendazole. Povidone–iodine was used to treat the liver cyst after closure of the diaphragmatic rupture. The drainage was considered successful, and the patient had no recurrence of signs and symptoms. Clinical, laboratory, and radiologic recovery was observed during 2.5 months of catheterization. The patient was asymptomatic after catheter drainage. No recurrence was detected during 86 months of follow-up. For inoperable patients with ruptured liver and lung hydatid cysts, percutaneous drainage with oral albendazole is an alternative treatment option to surgery. The percutaneous approach can be life-saving in such cases.

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

International Nuclear Information System (INIS)

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

2015-01-01

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

[Simultaneous Traumatic Rupture of Patellar Ligament and Contralateral Rupture of Quadriceps Femoris Muscle].

Science.gov (United States)

Hladký, V; Havlas, V

2017-01-01

Our paper presents a unique case of a 64-year-old patient after a fall, treated with oral antidiabetic drugs for type II diabetes mellitus. Following a series of examinations, a bilateral injury was diagnosed - patellar ligament tear on the right side and rupture of quadriceps femoris muscle on the left side. It is a rare injury, complicated by simultaneous involvement of both knee joints. The used therapy consisted of a bilateral surgery followed by gradual verticalisation, first with the support of a walking frame and later with the use of forearm crutches. During the final examination, the patient demonstrated full flexion at both knees, while an extension deficit of approx. 5 degrees was still present on the left side. The right knee X-ray showed a proper position of the patella after the removal of temporary tension band wire. Although the clinical results of operative treatment of both the patellar ligament rupture and rupture of quadriceps femoris muscle are in most cases good, early operative treatment, proper technique and post-operative rehabilitation are a prerequisite for success. Key words: knee injuries, patellar ligament, quadriceps muscle, rupture.

Hybrid preheat/recirculating steam generator

International Nuclear Information System (INIS)

Lilly, G.P.

1985-01-01

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

STEAM 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…

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...

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.

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)

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

Creep-rupture behavior of candidate Stirling engine iron supperalloys in high-pressure hydrogen. Volume 2: Hydrogen creep-rupture behavior

Science.gov (United States)

Bhattacharyya, S.; Peterman, W.; Hales, C.

1984-01-01

The creep rupture behavior of nine iron base and one cobalt base candidate Stirling engine alloys is evaluated. Rupture life, minimum creep rate, and time to 1% strain data are analyzed. The 3500 h rupture life stress and stress to obtain 1% strain in 3500 h are also estimated.

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.

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

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

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

International Nuclear Information System (INIS)

Lavalerie, Claude; Borrel, Christian.

1982-01-01

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

Comparison between smaller ruptured intracranial aneurysm and larger un-ruptured intracranial aneurysm: gene expression profile analysis.

Science.gov (United States)

Li, Hao; Li, Haowen; Yue, Haiyan; Wang, Wen; Yu, Lanbing; ShuoWang; Cao, Yong; Zhao, Jizong

2017-07-01

As it grows in size, an intracranial aneurysm (IA) is prone to rupture. In this study, we compared two extreme groups of IAs, ruptured IAs (RIAs) smaller than 10 mm and un-ruptured IAs (UIAs) larger than 10 mm, to investigate the genes involved in the facilitation and prevention of IA rupture. The aneurismal walls of 6 smaller saccular RIAs (size smaller than 10 mm), 6 larger saccular UIAs (size larger than 10 mm) and 12 paired control arteries were obtained during surgery. The transcription profiles of these samples were studied by microarray analysis. RT-qPCR was used to confirm the expression of the genes of interest. In addition, functional group analysis of the differentially expressed genes was performed. Between smaller RIAs and larger UIAs, 101 genes and 179 genes were significantly over-expressed, respectively. In addition, functional group analysis demonstrated that the up-regulated genes in smaller RIAs mainly participated in the cellular response to metal ions and inorganic substances, while most of the up-regulated genes in larger UIAs were involved in inflammation and extracellular matrix (ECM) organization. Moreover, compared with control arteries, inflammation was up-regulated and muscle-related biological processes were down-regulated in both smaller RIAs and larger UIAs. The genes involved in the cellular response to metal ions and inorganic substances may facilitate the rupture of IAs. In addition, the healing process, involving inflammation and ECM organization, may protect IAs from rupture.

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

International examples of steam generator replacement

International Nuclear Information System (INIS)

Wiechmann, K.

1993-01-01

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

Comparing slow and fast rupture in laboratory experiments

Science.gov (United States)

Aben, F. M.; Brantut, N.; David, E.; Mitchell, T. M.

2017-12-01

During the brittle failure of rock, elastically stored energy is converted into a localized fracture plane and surrounding fracture damage, seismic radiation, and thermal energy. However, the partitioning of energy might vary with the rate of elastic energy release during failure. Here, we present the results of controlled (slow) and dynamic (fast) rupture experiments on dry Lanhélin granite and Westerly granite samples, performed under triaxial stress conditions at confining pressures of 50 and 100 MPa. During the tests, we measured sample shortening, axial load and local strains (with 2 pairs of strain gauges glued directly onto the sample). In addition, acoustic emissions (AEs) and changes in seismic velocities were monitored. The AE rate was used as an indicator to manually control the axial load on the sample to stabilize rupture in the quasi-static failure experiments. For the dynamic rupture experiments a constant strain rate of 10-5 s-1 was applied until sample failure. A third experiment, labeled semi-controlled rupture, involved controlled rupture up to a point where the rupture became unstable and the remaining elastic energy was released dynamically. All experiments were concluded after a macroscopic fracture had developed across the whole sample and frictional sliding commenced. Post-mortem samples were epoxied, cut and polished to reveal the macroscopic fracture and the surrounding damage zone. The samples failed with average rupture velocities varying from 5x10-6 m/s up to >> 0.1 m/s. The analyses of AE locations on the slow ruptures reveal that within Westerly granite samples - with a smaller grain size - fracture planes are disbanded in favor of other planes when a geometrical irregularity is encountered. For the coarser grained Lanhélin granite a single fracture plane is always formed, although irregularities are recognized as well. The semi-controlled experiments show that for both rock types the rupture can become unstable in response to these

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 ...

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

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

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

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

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

Steam jet ejectors are examined automatically

International Nuclear Information System (INIS)

Lardiere, C.

2013-01-01

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

Steam generator tube integrity program

International Nuclear Information System (INIS)

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

1996-01-01

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

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.

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.

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.

Physical therapy in the conservative treatment for anterior cruciate ligament rupture followed by contralateral rupture: case report

OpenAIRE

Almeida, Gabriel Peixoto Leão; Arruda, Gilvan de Oliveira; Marques, Amélia Pasqual

2014-01-01

Although the surgical reconstruction be the obvious indication for the anterior cruciate ligament (ACL) lesion, there is no consensus on whether the results of surgery are superior to those obtained with nonsurgical management. The objective of this report was to describe a case of nonsurgical treatment for ACL rupture followed by a contralateral rupture. A 28-year-old female practitioner of muay-thai and handball suffered a non-contact ACL rupture in the left knee, and three months after the...

Steam generator water lancing

International Nuclear Information System (INIS)

Kamler, F.; Schneider, W.

1992-01-01

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

Steam turbines for nuclear power plants

International Nuclear Information System (INIS)

Kosyak, Yu.F.

1978-01-01

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

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

International Nuclear Information System (INIS)

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

1978-01-01

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

Use of ICD-10 codes to monitor uterine rupture

DEFF Research Database (Denmark)

Thisted, Dorthe L A; Mortensen, Laust Hvas; Hvidman, Lone

2014-01-01

OBJECTIVES: Uterine rupture is a rare but severe complication in pregnancies after a previous cesarean section. In Denmark, the monitoring of uterine rupture is based on reporting of relevant diagnostic codes to the Danish Medical Birth Registry (MBR). The aim of our study was to examine the vali......OBJECTIVES: Uterine rupture is a rare but severe complication in pregnancies after a previous cesarean section. In Denmark, the monitoring of uterine rupture is based on reporting of relevant diagnostic codes to the Danish Medical Birth Registry (MBR). The aim of our study was to examine...... uterine ruptures, the sensitivity and specificity of the codes for uterine rupture were 83.8% and 99.1%, respectively. CONCLUSION: During the study period the monitoring of uterine rupture in the MBR was inadequate....

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 ...

Safety Picks up "STEAM"

Science.gov (United States)

Roy, Ken

2016-01-01

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

Steam generator tube extraction

International Nuclear Information System (INIS)

Delorme, H.

1985-05-01

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

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

Science.gov (United States)

Zhang, Mingxing; Chen, Haiyan

2018-04-20

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

Particle deposition modeling in the secondary side of a steam generator bundle model

Energy Technology Data Exchange (ETDEWEB)

Mukin, Roman, E-mail: roman.mukin@psi.ch; Dehbi, Abdel, E-mail: abdel.dehbi@psi.ch

2016-04-01

A steam generator (SG) tube rupture (SGTR) model is studied in this paper. This model based on a experimental facility called Aerosol Trapping In a Steam Generator (ARTIST), which is a model of a scaled steam generator tube bundle consisting of 270 tubes and a guillotine tube to address aerosol deposition phenomena on two different scales: near the tube break, where the gas velocities and turbulence are very intensive, and far away from the break, where the flow velocities are three orders of magnitude lower. Owing to complexity of the flow, 3D simulations with highly resolved computational mesh near the break were done. First, the flow inside an isolated tube with a guillotine tube break has been studied in the framework of Reynolds Averaged Navier Stokes (RANS) approach. The next part is devoted to the simulation of an inclined gas jet entering the SG tube bundle via the guillotine tube breach with more advanced CFD tools. In particular, Detached Eddy Simulation (DES) and RANS are applied to tackle the wide range of flow scales. Flow field velocity comparison showed that DES results are reproducing wavy structure of the flow field in far field from the break observed in experiment. Particle transport and deposition is modelled by Lagrangian continuous random walk (CRW) model, which has been developed and validated previously. It is found that the DES combined with the CRW to supply fluctuating velocity components predicts deposition rates that are generally within the scatter of the measured data. Monodisperse, spherical SiO{sub 2} particles with AMMD = 1.4 μm were used as aerosol particles in simulations. To be economically feasible, the computations were made with the open source CFD code OpenFOAM. Comparison of the calculated flow with the experimental axial velocity distribution data at different vertical levels has been performed.

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.

49 CFR 229.105 - Steam generator number.

Science.gov (United States)

2010-10-01

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

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.

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.

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

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)

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)

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

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…

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.)

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.)

Penelitian Kerusakan pada sebuah Pipa Ketel Uap

OpenAIRE

Adnyana, D.N

2007-01-01

Failure Investigation on Ruptured Boiler Tube, This paper presents a failureinvestigation on a ruptured steam boiler tube. This boilertubewas madeoflow carbonsteel, having 76.1mm outside diameter and 5 mmthickness. Theboiler tube was installed in horizontal position on thefrontrear walls ofa steam boiler that was operated at steam pressure and temperature of60 bar and 275"C.respectively. The failure investigation was carried out by performing a number of examinationsincluding: macroscopy, met...

Steam generator reliability improvement project

International Nuclear Information System (INIS)

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

1987-01-01

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

Steam generator reliability improvement project

International Nuclear Information System (INIS)

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

1987-01-01

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

A drier unit for steam separators

International Nuclear Information System (INIS)

Peyrelongue, J.-P.

1973-01-01

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

Dynamic rupture simulation of the 2017 Mw 7.8 Kaikoura (New Zealand) earthquake: Is spontaneous multi-fault rupture expected?

Science.gov (United States)

Ando, R.; Kaneko, Y.

2017-12-01

The coseismic rupture of the 2016 Kaikoura earthquake propagated over the distance of 150 km along the NE-SW striking fault system in the northern South Island of New Zealand. The analysis of In-SAR, GPS and field observations (Hamling et al., 2017) revealed that the most of the rupture occurred along the previously mapped active faults, involving more than seven major fault segments. These fault segments, mostly dipping to northwest, are distributed in a quite complex manner, manifested by fault branching and step-over structures. Back-projection rupture imaging shows that the rupture appears to jump between three sub-parallel fault segments in sequence from the south to north (Kaiser et al., 2017). The rupture seems to be terminated on the Needles fault in Cook Strait. One of the main questions is whether this multi-fault rupture can be naturally explained with the physical basis. In order to understand the conditions responsible for the complex rupture process, we conduct fully dynamic rupture simulations that account for 3-D non-planar fault geometry embedded in an elastic half-space. The fault geometry is constrained by previous In-SAR observations and geological inferences. The regional stress field is constrained by the result of stress tensor inversion based on focal mechanisms (Balfour et al., 2005). The fault is governed by a relatively simple, slip-weakening friction law. For simplicity, the frictional parameters are uniformly distributed as there is no direct estimate of them except for a shallow portion of the Kekerengu fault (Kaneko et al., 2017). Our simulations show that the rupture can indeed propagate through the complex fault system once it is nucleated at the southernmost segment. The simulated slip distribution is quite heterogeneous, reflecting the nature of non-planar fault geometry, fault branching and step-over structures. We find that optimally oriented faults exhibit larger slip, which is consistent with the slip model of Hamling et al

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...

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

Multi-Fault Rupture Scenarios in the Brawley Seismic Zone

Science.gov (United States)

Kyriakopoulos, C.; Oglesby, D. D.; Rockwell, T. K.; Meltzner, A. J.; Barall, M.

2017-12-01

Dynamic rupture complexity is strongly affected by both the geometric configuration of a network of faults and pre-stress conditions. Between those two, the geometric configuration is more likely to be anticipated prior to an event. An important factor in the unpredictability of the final rupture pattern of a group of faults is the time-dependent interaction between them. Dynamic rupture models provide a means to investigate this otherwise inscrutable processes. The Brawley Seismic Zone in Southern California is an area in which this approach might be important for inferring potential earthquake sizes and rupture patterns. Dynamic modeling can illuminate how the main faults in this area, the Southern San Andreas (SSAF) and Imperial faults, might interact with the intersecting cross faults, and how the cross faults may modulate rupture on the main faults. We perform 3D finite element modeling of potential earthquakes in this zone assuming an extended array of faults (Figure). Our results include a wide range of ruptures and fault behaviors depending on assumptions about nucleation location, geometric setup, pre-stress conditions, and locking depth. For example, in the majority of our models the cross faults do not strongly participate in the rupture process, giving the impression that they are not typically an aid or an obstacle to the rupture propagation. However, in some cases, particularly when rupture proceeds slowly on the main faults, the cross faults indeed can participate with significant slip, and can even cause rupture termination on one of the main faults. Furthermore, in a complex network of faults we should not preclude the possibility of a large event nucleating on a smaller fault (e.g. a cross fault) and eventually promoting rupture on the main structure. Recent examples include the 2010 Mw 7.1 Darfield (New Zealand) and Mw 7.2 El Mayor-Cucapah (Mexico) earthquakes, where rupture started on a smaller adjacent segment and later cascaded into a larger

Steam explosion studies review

International Nuclear Information System (INIS)

Hwang, Moon Kyu; Kim, Hee Dong

1999-03-01

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

Water regime of steam power plants

International Nuclear Information System (INIS)

Oesz, Janos

2011-01-01

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

[Effects of posterior tibial slope on non-contact anterior cruciate ligament rupture and stability of anterior cruciate ligament rupture knee].

Science.gov (United States)

Yue, De-bo; E, Sen; Wang, Bai-liang; Wang, Wei-guo; Guo, Wan-shou; Zhang, Qi-dong

2013-05-07

To retrospectively explore the correlation between anterior cruciate ligament (ACL)-ruptured knees, stability of ACL-rupture knee and posterior tibial slope (PTS). From January 2008 to October 2012, 150 knees with ACL rupture underwent arthroscopic surgery for ACL reconstruction. A control group was established for subjects undergoing arthroscopic surgery without ACL rupture during the same period. PTS was measured on a digitalized lateral radiograph. Lachman and mechanized pivot shift tests were performed for assessing the stability of knee. There was significant difference (P = 0.007) in PTS angle between the patients with ACL rupture (9.5 ± 2.2 degrees) and the control group (6.6 ± 1.8 degrees). Only among females, increased slope of tibial plateau had effect on the Lachman test. There was a higher positive rate of pivot shift test in patients of increased posterior slope in the ACL rupture group. Increased posterior tibial slope (>6.6) appears to contribute to non-contact ACL injuries in females. And the changes of tibial slope have no effect upon the Lachman test. However, large changes in tibial slope affect pivot shift.

Moisture separator for steam generator level measurement system

International Nuclear Information System (INIS)

Cantineau, B.J.

1987-01-01

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

Future development of large steam turbines

International Nuclear Information System (INIS)

Chevance, A.

1975-01-01

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

Vapor generator steam drum spray heat

International Nuclear Information System (INIS)

Fasnacht, F.A. Jr.

1978-01-01

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

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

International Nuclear Information System (INIS)

Ki Bang, Sung; Young Jin, Chang

2001-01-01

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

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.

High-temperature oxidation of Zircaloy in hydrogen-steam mixtures

International Nuclear Information System (INIS)

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

1982-09-01

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

Thermal hydraulic studies in steam generator test facility

International Nuclear Information System (INIS)

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

2005-01-01

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

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

Improving Steam System Performance: A Sourcebook for Industry

Energy Technology Data Exchange (ETDEWEB)

2002-06-01

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

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

Liquid metal steam generator

International Nuclear Information System (INIS)

Wolowodiuk, W.

1975-01-01

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

Steam purity in PWRs

International Nuclear Information System (INIS)

Hopkinson, J.; Passell, T.

1982-01-01

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

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

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...

CAREM-25 Steam Generator Stability Analysis

International Nuclear Information System (INIS)

Rabiti, A.; Delmastro, D.

2003-01-01

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

46 CFR 61.15-5 - Steam piping.

Science.gov (United States)

2010-10-01

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