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Sample records for candu pressure tubes

  1. Pressure tube life management in CANDU-6 nuclear plant

    International Nuclear Information System (INIS)

    Operating parameters of pressure tube in CANDU-6 reactor, the relation between pressure tube life and plant life improvement of pressure tube by AECL in past years were summarized, and the factors affecting pressure tube life, idea and main measures of pressure tube life management in QINSHAN CANDU-6 power plant introduced

  2. Ballooning of CANDU pressure tubes. Model assessment

    International Nuclear Information System (INIS)

    The transient creep equations used to analyze the possible ballooning and failure of Zr-2.5% Nb pressure tubes during a loss-of-coolant accident (LOCA) were developed and verified using as-received Zr-2.5% Nb pressure tube material. But in a CANDU reactor, the pressure tubes absorb deuterium and are exposed to a continuous neutron fluence. Consequently, a literature survey was done to determine how irradiation damage and deuterium might affect the creep rate and ductility of Zr-2.5% Nb pressure tubes in the temperature range from 600 to 800 degrees C. It was found that irradiation damage, dissolved deuterium and deuteride blisters could possibly affect the creep rate and ductility of ZR-2.5% Nb pressure tubes in this temperature range, but deuteride platelets are expected to have little effect. Further tests are required to determine the effect of irradiation damage and deuterium on the creep rate and ductility of pressure tubes

  3. Structural integrity evaluations of CANDU pressure tubes

    International Nuclear Information System (INIS)

    The core of a CANDU-6 pressurized heavy water reactor consists of some hundred horizontal pressure tubes that are manufactured from a Zr-2.5%Nb alloy and which contain the fuel bundles. These tubes are susceptible to a damaging phenomenon known as Delayed Hydride Cracking (DHC). The Zr-2.5%Nb alloy is susceptible to DHC phenomenon when there is diffusion of hydrogen atoms to a service-induced flaws, followed by the hydride platelets formation on the certain crystallographic planes in the matrix material. Finally, the development of hydride regions at the flaw-tip will happened. These hydride regions are able to fracture under stress-temperature conditions (DHC initiation) and the cracks can extend and grow by DHC mechanism. Some studies have been focused on the potential to initiate DHC at the blunt flaws in a CANDU reactor pressure tube and a methodology for structural integrity evaluation was developed. The methodology based on the Failure Assessment Diagrams (FAD's) consists in an integrated graphical plot, where the fracture failure and plastic collapse are simultaneously evaluated by means of two non-dimensional variables (Kr and Lr). These two variables represent the ratio of the applied value of either stress or stress intensity factor and the resistance parameter of corresponding magnitude (yield stress or fracture toughness, respectively). Once the plotting plane is determined by the variables Kr and Lr, the procedure defines a critical failure line that establishes the safe area. The paper will demonstrate the possibility to perform structural integrity evaluations by means of Failure Assessment Diagrams for flaws occurring in CANDU pressure tubes. (author)

  4. Ultrasonic crack-tip diffraction in CANDU reactor pressure tubes

    International Nuclear Information System (INIS)

    Currently there is no reliable method of measuring defect depths in CANDU reactor pressure tubes. The demonstrated success of crack-tip diffraction (or time-of-flight-testing) in round-robins on thick components has promoted an interest in this technique. In CANDU reactors, pressure tubes are effectively accessible only from the inside. Development work has concentrated on outside surface defects using 45 degree shear waves in contrast to the longitudinal waves usually used for testing thick components with this technique. Due to the small wall thickness of the pressure tubes (4.2 mm) and the typical sizes of defects of interest (0.15 mm or greater), frequencies of the order of 20 MHz are being used. A further complication comes from the orientation of the defects, which may be at any angle in pressure tubes. Initial studies have been performed on a series of outside surface notches and slots, plus a real fatigue crack. This crack was on the inside surface, so the technique required measuring this defect's depth from the outside. Initial results are encouraging. Even without signal processing, crack-tip diffracted signals were detectable from all but very large (2.5 mm) and very small (less than 0.076 mm) notches. Errors in estimates of defect depths were typically less than 0.1 mm for all the notches, and the results were consistent. Measurements on the fatigue crack showed similar random errors, though there appeared to be a deterministic error of about 0.1 mm as well

  5. Ballooning of CANDU pressure tubes - experiments with degraded tube material

    International Nuclear Information System (INIS)

    Three as-received Zr-2.5% Nb pressure tube specimens and three specimens with eight 0.5 mm deep defects machined on the inside surface were tested in the ballooning test rig at Stern Laboratories Inc. The temperature ramp rate was controlled between 28 K s-1 and 35 K s-1. Temperatures on the outside and inside surfaces of the specimens, and circumferential and longitudinal strains were recorded during the transients. Post-test longitudinal, circumferential and wall thickness strains were measured. All as-received specimens ruptured full-length near the top, i.e., the hottest point. All defected specimens failed at either or both upper defects, one rupture being full-length and the others limited to one to three times the length of the defect. (author). 4 refs., 2 tabs., 15 figs

  6. Mechanistic modeling of thermal-mechanical deformation of CANDU pressure tube under localized high temperature condition

    International Nuclear Information System (INIS)

    Thermal strain deformation is a pressure tube failure mechanism. The main objective of this paper is to develop mechanistic models to evaluate local thermal-mechanical deformation of a pressure tube in CANDU reactor and to investigate fuel channel integrity under localized contact between fuel elements and pressure tube. The consequence of concern is potential creep strain failure of a pressure tube and calandria tube. The initial focus will be on the case where a fuel rod contacts the pressure tube at full power with highly cooling condition

  7. Wet channel measurement of pressure tube to calandria tube spacing in CANDU reactors

    International Nuclear Information System (INIS)

    The pressure tube (PT) to calandria tube (CT) spacing in CANDU reactors is an important parameter that relates to the general condition of the fuel channels. The measurement system that was developed to measure this parameter during the wet channel inspections of Pickering Units 1 and 2 is described in this paper. A send-receive eddy current probe was designed which is primarily sensitive to variations in PT/CT spacing but is also affected by pressure tube wall thickness. A computer simulation showed that the phase angles of the response to these variables are similar for all usable frequencies, thus eliminating the possibility of multifrequency compensation. A marriage of technologies was proposed involving the ultrasonic measurement of wall thickness values which are then used to extract the spacing information from the eddy current signal. The accuracy of the system is approximately ±(30% +.1mm) which has been sufficient to determine if and where any of the pressure tubes have come in contact with their calandria tube. Field experience with the new system is discussed and areas for development are also outlined

  8. A probabilistic method for leak-before-break analysis of CANDU reactor pressure tubes

    Energy Technology Data Exchange (ETDEWEB)

    Puls, M.P.; Wilkins, B.J.S.; Rigby, G.L. [Whiteshell Labs., Pinawa (Canada)] [and others

    1997-04-01

    A probabilistic code for the prediction of the cumulative probability of pressure tube ruptures in CANDU type reactors is described. Ruptures are assumed to result from the axial growth by delayed hydride cracking. The BLOOM code models the major phenomena that affect crack length and critical crack length during the reactor sequence of events following the first indications of leakage. BLOOM can be used to develop unit-specific estimates of the actual probability of pressure rupture in operating CANDU reactors and supplement the existing leak before break analysis.

  9. A probabilistic method for leak-before-break analysis of CANDU reactor pressure tubes

    International Nuclear Information System (INIS)

    A probabilistic code for the prediction of the cumulative probability of pressure tube ruptures in CANDU type reactors is described. Ruptures are assumed to result from the axial growth by delayed hydride cracking. The BLOOM code models the major phenomena that affect crack length and critical crack length during the reactor sequence of events following the first indications of leakage. BLOOM can be used to develop unit-specific estimates of the actual probability of pressure rupture in operating CANDU reactors and supplement the existing leak before break analysis

  10. Pressure tube creep impact on the physics parameters for CANDU-6 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, W. Y.; Min, B. J. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Kam, S. C.; Kim, M. E. [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of)

    2004-07-01

    The lattice cell calculations are performed to assess the sensitivity of the reactor physics parameters to pressure tube creep resulting from radiation aging. The physics parameters of the lattice cell are calculated by using WIMSD-5B code, WIMS- AECL code, and MCNP code. The reference model(normal state) and two perturbed models accounting for the pressure tube creep are developed on the basis of CANDU-6 lattice cell. The 2.5% and 5% values of pressure tube diameter creep are considered. Also, The effects of the analyzed lattice parameters which are the coolant void reactivity, the fuel fission density and the atom density of Pu isotopes on the lattice.

  11. Critical heat flux in CANDU moderator following a pressure tube to calandria tube contact - part I

    International Nuclear Information System (INIS)

    Heavy water moderator surrounding each fuel channel is one of the important features in CANDU reactors that act as a heat sink for the fuel in the situations where other means of heat removal fail. In the critical break LOCA scenario, fuel cooling becomes severely degraded due to rapid flow reduction in the affected flow pass of the heat transport system. This can result in pressure tubes experiencing significant heat-up while coolant pressure is still high, thereby causing uniform thermal creep strain (ballooning) of the pressure tube (PT) into contact with its calandria tube (CT). The contact of the hot PT with the CT causes rapid redistribution of stored heat from the PT to CT and a large spike in heat flux from the CT to the moderator fluid. For lower subcooling conditions of the moderator, this heat flux spike can cause dryout of the CT. This can detrimentally affect channel integrity if the CT post-dryout temperature becomes sufficiently high to result in thermal creep strain deformation. The focus of this research is to develop a mechanistic model to predict Critical Heat Flux (CHF) on the CT surface following a contact with its pressure tube. A COMSOL multi-physics model using a two-dimensional transient fluid-thermal analysis of the CT surface undergoing heat up is used to predict flow and temperature profile on the CT surface. A mechanistic CHF model is to be proposed based on a concept of wall dry patch formation, prevention of rewetting and subsequent dry patch spreading. (author)

  12. Sizing cracks in thin-walled CANDU reactor pressure tubes using crack-tip diffraction

    International Nuclear Information System (INIS)

    The most practical nondestructive means of measuring the depth of cracks approximately 0.4 mm deep in CANDU reactor pressure tubes is the ultrasonic crack-tip diffraction method. Initially, optimum ultrasonic parameters for wave mode, transducer frequency, main-bang pulse characteristics, incident and diffracted angles were obtained on three fatigue cracks, based on the criteria of maximum signal amplitude and accuracy in determination of crack depth. In addition, three signal processing techniques, auto and cross-correlation, rectification and smoothing and the magnitude of the analytic signal, were used to obtain time measurements. The results of these measurements are presented. Except for the first fatigue crack, the depth calculations were accurate to within the specified range of ± 0.1 mm

  13. Mechanistic modeling of bearing pad to pressure tube contact under localized high temperature conditions in a CANDU fuel channel

    International Nuclear Information System (INIS)

    During a postulated critical break LOCA (loss of coolant accident) in a CANDU reactor the coolant flow rates in the fuel channels of the flow pass of the reactor core downstream of the pipe break can rapidly reduce to very low values and remain very low for a period of tens of seconds following the break. Under the sustained low flow conditions, the fuel sheath (cladding) temperature in the affected channels rapidly increases and the coolant in the channels becomes significantly voided. The pressure tubes in the affected pass heat up under a combination of convection heat transfer from the low flows of superheated seam and thermal radiation heat transfer from the hot fuel. Additionally, hot spots may develop on the inner surface of pressure tubes at locations where the fuel bearing pads are in direct contact with the pressure tube. Localized thermal creep strain deformation at the hot spots is a potential pressure tube failure mechanism which could challenge fuel channel integrity. This paper evaluates the local thermal-mechanical deformation of a pressure tube in a CANDU reactor under critical break LOCA conditions tube using a coupled thermal-mechanical finite element COMSOL multi-physics model and investigates the conditions resulting in fuel channel failure due to localized contact between bearing pad and pressure. The mechanistic models are validated against data obtained from COG funded experiments performed at WRL (Whiteshell Research Laboratory). Multiphysics calculations are performed in which the heat transfer, thermal-mechanical and creep strain equations are solved, simultaneously. Heat conduction from bearing pads to the inner surface of the pressure tube is modeled with appropriate convective and radiation heat transfer boundary conditions. Thermal creep strain deformation of the Zr-2.5%Nb pressure tube is modeled using correlations derived from separate uniaxial tests that are reported in the literature. Contact conductance models based on

  14. Oxidation and deuterium uptake of Zr-2.5Nb pressure tubes in CANDU-PHW reactors

    International Nuclear Information System (INIS)

    Oxidation and deuterium uptake in Zr-2.5Nb pressure tubes are being monitored by destructive examination of tubes removed from commercial Canadian deuterium uranium pressurized heavy-water (CANDU-PHW) stations and by analyses of microsamples, obtained in-situ, from the inside surface of tubes in the reactor. Unlike Zircaloy-2, there is no evidence for any acceleration in the oxidation rate for exposures up to about 4500 effective full power days. Changes towards a more equilibrium microstructure during irradiation may be partly responsible for maintaining the low oxidation rate, since thermal aging treatments, producing similar microstructural changes in initially cold worked tubes, were found to improve out-reactor corrosion resistance in 589 K water. With one exception, the deuterium uptake in Zr-2.5Nb tubes has been remarkably low and no greater than 3-mg/kg deuterium per year (0.39 mg/dm2 hydrogen per year) . The exception is the most recent surveillance tube removed from Pickering (NGS) Unit 3, which had a deuterium content near the outlet end about five times higher than that seen in the previous tube examined. Current investigations suggest that most of the uptake in that tube may have come from the gas annulus surrounding the tube where deuterium exists as an impurity, and oxidation has been insufficient to maintain a protective oxide film. Results from weight gain measurements, chemical analyses, metallography, scanning electron microscopy, and transmission electron microscopy of irradiated pressure tubes and of small coupons exposed out reactor are presented and discussed with respect to the observed corrosion and hydriding behavior of CANDU-PHW pressure tubes. (author)

  15. PARTICULARITIES REGARDING THE OPERATING PROCESS OF THE CUTTING AND EXTRACTION DEVICE IN THE CANDU HORIZONTAL FUEL CHANNELS PRESSURE TUBE DECOMMISSIONING PART II: CUTTING AND EXTRACTING PRESSURE TUBE PROCESS

    Directory of Open Access Journals (Sweden)

    Constantin POPESCU

    2016-05-01

    Full Text Available This paper presents some details of operation process for a Cutting and Extraction Device (CED in order to achieve the decommissioning of the horizontal fuel channels pressure tube in the CANDU 6 nuclear reactor. The most important characteristic of the Cutting and Extraction Device (CED is his capability of totally operator’s protection against the nuclear radiation during pressure tube decommissioning. The cutting and extracting pressure tube processes present few particularities due to special adopted technical solutions: a special module with three cutting rollers (system driven by an actuator, a guiding-extracting and connecting module (three fixing claws which are piloted by an actuator and block the device in the connecting position with extracting plugs. The Cutting and Extraction Device (CED is a train of modules equipped with special systems to be fully automated, connected with a Programmable Logic Controller (PLC and controlled by an operator panel type Human Machine Interface (HMI. All processes are monitored by video cameras. In case of error, the process is automatically stopped, the operator receiving an error message and the last sequence could be reinitialized or aborted due to safety reasons.

  16. Implementation of Wolsong Pump Model, Pressure Tube Deformation Model and Off-take Model into MARS Code for Regulatory Auditing of CANDU Reactors

    International Nuclear Information System (INIS)

    Korea currently has four operating units of the CANDU-6 type reactor in Wolsong. However, the safety assessment system for CANDU reactors has not been fully established due to lack of self-reliance technology. Although the CATHENA code had been introduced from AECL, it is undesirable to use vendor's code for regulatory auditing analysis. In Korea, the MARS code has been developed for decades and is being considered by KINS as a thermal hydraulic regulatory auditing tool for nuclear power plants. Before this decision, KINS (Korea Institute of Nuclear Safety) had developed RELAP5/MOD3/CANDU code for CANDU safety analyses by modifying the model of existing PWR auditing tool, RELAP5/MOD3. The main purpose of this study is to transplant the CANDU models of RELAP5/MOD3/CANDU code to MARS code including quality assurance of the developed models. This first part of the research series presents the implementation and verification of the Wolsong pump model, the pressure tube deformation model, and the off-take model for arbitrary-angled branch pipes

  17. Implementation of Wolsong Pump Model, Pressure Tube Deformation Model and Off-take Model into MARS Code for Regulatory Auditing of CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, C.; Rhee, B. W.; Chung, B. D. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Y. J.; Kim, M. W. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2008-05-15

    Korea currently has four operating units of the CANDU-6 type reactor in Wolsong. However, the safety assessment system for CANDU reactors has not been fully established due to lack of self-reliance technology. Although the CATHENA code had been introduced from AECL, it is undesirable to use vendor's code for regulatory auditing analysis. In Korea, the MARS code has been developed for decades and is being considered by KINS as a thermal hydraulic regulatory auditing tool for nuclear power plants. Before this decision, KINS (Korea Institute of Nuclear Safety) had developed RELAP5/MOD3/CANDU code for CANDU safety analyses by modifying the model of existing PWR auditing tool, RELAP5/MOD3. The main purpose of this study is to transplant the CANDU models of RELAP5/MOD3/CANDU code to MARS code including quality assurance of the developed models. This first part of the research series presents the implementation and verification of the Wolsong pump model, the pressure tube deformation model, and the off-take model for arbitrary-angled branch pipes.

  18. Development of a CANDU fuel channel model to assess the effect of a pressure tube creep on the safety related parameters

    International Nuclear Information System (INIS)

    Recently the effect of pressure tube creep on the reactor safety in CANDUs emerges as an important issue of safety analysis due to a need for an extended operation. The accident analysis for the aged plants needs to incorporate major degradations of the plant performance in the safety analysis. In this paper, a CATHENA fuel channel model for studying the effects of the vertical offset of the fuel bundles in a crept pressure tube on the fuel and pressure tube cooling is developed. The current practice of the CANDU safety analysis assumes that the fuel bundles stay in a manner concentric to the pressure tube centerline even in the crept pressure tubes, whereas in reality the bundles sit at the bottom of the pressure tube. With this point in mind, 37-pin models with and without vertical offset of the bundle in the crept fuel channel are developed and tested for Reactor Outlet Header (ROH) 100% break LOCA accident, and results compared. As a result, it was found that the difference between the uncrept fuel channel model and the two crept fuel channel models, a concentric one and another vertically offset one, is quite significant, whereas the difference between the two crept fuel channel models is insignificant. Therefore it is concluded that the use of the concentric crept fuel channel model for the aged CANDU-6 safety analysis is justifiable for the first 200 sec into an accident. (author)

  19. Characterization of excel alloy pressure tube material for CANDU SCW reactors

    International Nuclear Information System (INIS)

    The phase transformation temperatures, aging response, and creep rupture strength of Zr alloy Excel (Zr- 3.5%Sn- 0.8%Nb- 0.8%Mo) pressure tube material were investigated. The α → α+β and α+β → β transus temperatures were found to be in the range of 600-690 °C and 962-975 °C respectively. Precipitation hardening was observed in the microstructures water-quenched from high in the α+β or β regions followed by aging at 400-500 °C for 1 hr. The results of creep-rupture experiments at 400 °C suggest that a fully martensitic and aged microstructure has better creep properties at high stress levels (>700 MPa) and a microstructure obtained by air-cooling from high in the α+β region shows good creep properties at lower stresses (<560 MPa). (author)

  20. Silicon carbide TRIPLEX materials for CANDU fuel cladding and pressure tubes

    International Nuclear Information System (INIS)

    Ceramic Tubular Products has developed a superior silicon carbide (SiC) material TRIPLEX, which can be used for both fuel cladding and other zirconium alloy materials in light water reactor (LWR) and heavy water reactor (CANDU) systems. The fuel cladding can replace Zircaloy cladding and other zirconium based alloy materials in the reactor systems. It has the potential to provide higher fuel performance levels in currently operating natural UO2 (NEU) fuel design and in advanced fuel designs (UO2(SEU), MOX thoria) at higher burnups and power levels. In all the cases for fuel designs TRIPLEX has increased resistance to severe accident conditions. The interaction of SiC with steam and water does not produce an exothermic reaction to produce hydrogen as occurs with zirconium based alloys. In addition the absence of creep down eliminates clad ballooning during high temperature accidents which occurs with Zircaloy blocking water channels required to cool the fuel. (author)

  1. Selection of instruments used for vibration measurement of fuel bundles in a pressure tube under CANDU reactor operating conditions

    International Nuclear Information System (INIS)

    Vibration characteristics of CANDU fuel bundle and fuel elements is a key parameter considered in the design of a fuel bundle. Out-reactor frequency and temperature sweep tests, under reactor operating conditions, are performed to verify vibration characteristics of CANDU fuel bundles. Several options have been considered in the selection of vibration instrumentation to perform out-reactor frequency and temperature sweep tests. This paper compares the benefits and disadvantages of various vibration instruments and summarizes the rationale behind the selection of instruments used for vibration measurements over a range of temperature and pressure pulsation frequencies. The conclusions are presented from the bench tests performed, which confirm the use of the selected instruments. (author)

  2. Ultrasonic measurement method of calandria tube sagging in CANDU reactor

    International Nuclear Information System (INIS)

    Calandria tube wrapping each pressure tube is one of the key structural components of CANDU reactor (calandria) which is consisted of many pressure tubes containing nuclear fuel assemblies. As the calandria tube (made of zirconium alloy) is sagging due to its thermal and irradiation creep during the plant operation, it possibly contacts with liquid injection nozzle crossing beneath the calandria tube, which subsequently results in difficulties on the safe operation. It is therefore necessary to check the gap for the confirmation of no contacts between the two tubes, calandria tube and liquid inject ion tube, with a proper measure during the life of plant. In this study, an ultrasonic measurement method was selected among several methods investigated. The ultrasonic device being developed for the measurement of the gap was introduced and its preliminary performance test results were presented here

  3. PARTICULARITIES REGARDING THE OPERATING PROCESS OF THE CUTTING AND EXTRACTION DEVICE IN THE CANDU HORIZONTAL FUEL CHANNELS PRESSURE TUBE DECOMMISSIONING PART I: MOVEMENT AND FIXING DEVICE INSIDE THE PRESSURE TUBE

    Directory of Open Access Journals (Sweden)

    Constantin POPESCU

    2016-05-01

    Full Text Available This paper presents some details of operation process for a Cutting and Extraction Device (CED in order to achieve the decommissioning of the horizontal fuel channels pressure tube in the CANDU 6 nuclear reactor. The most important characteristic of the Cutting and Extraction Device (CED is his capability of totally operator’s protection against the nuclear radiation during pressure tube decommissioning. The movement and fixing processes present few particularities due to special adopted technical solutions: train guiding-fixing modules equipped with elastic guiding rollers and fixing claws, traction modules with elastic rollers and variable pitch, also with propriety to adapt the system according to various dimensions of the tube. The Cutting and Extraction Device (CED is a train of modules equipped with special systems to be fully automated, connected with a Programmable Logic Controller (PLC and controlled by an operator panel type Human Machine Interface (HMI. All processes are monitored by video cameras. In case of error, the process is automatically stopped, the operator receiving an error message and the last sequence could be reinitialized or aborted due to safety reasons

  4. Ultrasonic measurement of gap between calandria tube and liquid injection nozzle in CANDU reactor

    International Nuclear Information System (INIS)

    Calandria tube wrapping each pressure tube is one of the key structural components of CANDU reactor(Calandria) which is consisted of many pressure tubes containing nuclear fuel assemblies. As the Calandria tube(made of zirconium alloy) is sagging due to its thermal and irradiation creep during the plant operation, ti possibly contacts with liquid injection nozzle crossing beneath the Calandria tube, which subsequently results in difficulties on the safe operation. It is therefore necessary to check the gap for the confirmation of no contacts between the two tubes, Calandria tube and liquid injection tube, with a proper measure during the life of plant. In this study, an ultrasonic measurement method was selected among several methods investigated. The ultrasonic device being developed for the measurement of the gap was introduced and its preliminary performance test results were presented here. The gap between LIN and CT at site was measured using by this ultrasonic device at site

  5. Designing and calculating the pressure loses for different geometries of CANDU type fuel clusters

    International Nuclear Information System (INIS)

    It is well known that circulation of the coolant through the pressure tube of a CANDU type reactor must ensure, through its flow rate values, the optimal conditions of heat transfer from the fuel clusters towards the heavy water. The flow rate through fuel channels differs from one another (up to 24 kg/s) depending on the fuel element sheath temperature, the latter depending in turn one the channels/clusters positions in the calandria vessel. In these conditions, one of the main problem of design in the CANDU type reactor plants is related to the hydraulic resistance represented by the fuel clusters loading the pressure tube or, in other words, the problem of pressure losses (pressure drops) over the length of the fuel cluster column. More precisely, this hydraulic resistance should not exceed a given value imposed by the performance calculations for the pumps used. A sustained activity of analysing comparatively the different geometry types of the fuel clusters was developed at INR Pitesti, a special attention being paid to their behavior as hydraulic resistances. The paper presents a set of computation programs devoted on one hand to the design of fuel clusters of different types and to an estimating computation of the pressure losses resulting from loading these clusters into a specific fuel channel of the CANDU type reactor, on the other hand. During the presentation of the work, different computing codes will be run for demonstration

  6. Bearing pad to pressure tube contact simulation

    Energy Technology Data Exchange (ETDEWEB)

    Talebi, F.; Behdadi, A.; Luxat, J.C., E-mail: farshat@mcmaster.ca, E-mail: behdada@mcmaster.ca, E-mail: luxatj@mcmaster.ca [McMaster Univ., Dept. of Engineering Physics, Hamilton, Ontario (Canada)

    2011-07-01

    Thermal creep strain deformation is a very important pressure tube failure mechanism. During a postulated LOCA (loss of coolant accident) with failure of emergency core injection sys- tem (ECIS), the fuel cladding temperature rapidly increases and the pressure tube becomes completely dry in a few seconds after flow stagnation occurs. Subsequently, the pressure tube circumference is heated by thermal radiation except at the spots where the bearing pads are in direct contact with the pressure tube. Therefore, the localized hot spots are developed on the pressure tube's inner surface under the bearing pads. The main objective of this paper is to evaluate the local thermal-mechanical deformation of a pressure tube in a CANDU reactor and to investigate the fuel channel integrity under localized contact between bearing pad and pressure tube. Furthermore, the mechanistic models are validated against the experimental works per- formed at WRL (Whiteshell research laboratory). Calculations are performed using the finite element method in which the heat, thermal mechanical and creep strain equations are solved, simultaneously. According to the experimental set up, the heat conduction from bearing pads to the inner surface of the pressure tube with appropriate convective and radiation boundary conditions has been simulated. Furthermore, the thermal creep strain deformation has been obtained for when the pressure tube is still under operational condition. It is observed that the pressure tube thermal strain will occur if sufficient high temperature is reached however, depending on the severity of flow degradation in the fuel channel, these localized hot spots could represent a potential creep strain failure of the pressure tube. Whether the pressure tube would fail at these hot spots before contacting the calandria tube depends on the localized temperature and experienced pressure transients. Sensitivity analysis is performed in order to evaluate the contact conductance

  7. Innovation in pressure tube life assessment

    International Nuclear Information System (INIS)

    The hydrogen equivalent concentration and the rate of hydrogen ingress (in particular, deuterium) in pressure tubes are important parameters that must be assessed to determine the fitness-for-service of CANDU reactors. This paper presents the latest refinement in a process referred to as 'Pressure Tube Sampling', which is the only fully qualified and proven method that allows accurate determination of both the hydrogen equivalent concentration and the rate deuterium ingress without performing an expensive fuel channel removal. Pressure Tube Sampling has evolved over the past fifteen years during which over 2,300 samples have been obtained from CANDU reactors around the globe. In-reactor sampling is the standard method for determining the hydrogen equivalent concentrations and deuterium ingress rates in CANDU reactors. Over the past fifteen years, continual improvements in the Pressure Tube Sampling process have resulted in: the capability to obtain circumferential and axial samples, reduced 'on-face' time, reduced cost, reduced dose to workers, and improved analysis accuracy. Most recently, the new Multi-Head Sampling Tool (MHST) has been developed that continues this trend by using one tool to sample at all four axial pressure tube locations in a single visit to the fuel channel, thereby further improving efficiency. In 2001 October, the MHST was successfully deployed at Wolsong 1 by AECL for Korea Hydro and Nuclear Power. The tool was delivered using their Advanced Delivery Machine (ADM) and a total of sixteen samples were obtained from four channels. A significant saving in time was achieved with a rate of one channel (four samples) being sampled every 2 1/2 hours. For a typical 10-channel campaign, this could equate to a 2 to 3 days time/saving, which is significant in terms of outage schedule, cost, and worker dose. This paper provides a description of some of the latest innovations, with specific details on site application, performance, and end results

  8. CANDU bundle junction. Misalignment probability and pressure-drop correlation

    International Nuclear Information System (INIS)

    The pressure drop over the bundle junction is an important component of the pressure drop in a CANDU (Canada Deuterium Uranium) fuel channel. This component can represent from ∼ 15% for aligned bundles to ∼ 26% for rotationally misaligned bundles, and is dependent on the degree of misalignment. The geometry of the junction increases the mixing between subchannels, and hence improves the thermal performance of the bundle immediately downstream. It is therefore important to model the junction's performance adequately. This paper summarizes a study sponsored by COG (CANDU Owners Group) and an NSERC (National Science and Engineering Research Council) Industrial Research Grant, undertaken, at CRL (Chalk River Laboratories) to identify and develop a bundle-junction model for potential implementation in the ASSERT (Advanced Solution of Subchannel Equations in Reactor Thermalhydraulics) subchannel code. The work reported in this paper consists of two components of this project: an examination of the statistics of bundle misalignment, demonstrating that there are no preferred positions for the bundles and therefore all misalignment angles are equally possible; and, an empirical model for the single-phase pressure drop across the junction as a function of the misalignment angle. The second section of this paper includes a brief literature review covering the experimental, analytical and numerical studies concerning the single-phase pressure drop across bundle junctions. 32 refs., 9 figs

  9. Aging of elastomers in CANDU pressure boundary service

    International Nuclear Information System (INIS)

    This report describes the properties and aging of elastomers, and examines the performance of major elastomeric components in CANDU pressure boundary service. The components examined are vacuum building roof seals, pressure relief duct seals, airlock door seals, fuelling machine hoses, and cable penetrations. For each of these components, the design requirements, technical specifications and component testing procedures are compared with applicable standards. Information on actual and recommended monitoring and maintenance methods is presented. Operational and environmental stressors are identified. Component failure modes, causes and frequencies are described, as well as the remedial action taken. Many different elastomers are used in CANDU plants, for many different applications. Standards and manufacturers' recommendations are not consistent and may vary from one component to another. Accordingly, the monitoring, maintenance and replacement practices tend to vary from one application to another, and may also be different at different stations. Recommendations are given in this report for improved monitoring and maintenance, in an attempt to provide more consistency in approach. A summary of some experiences with elastomers from non-Canadian sources is contained in the last section. 125 refs

  10. Pressure tube type reactor

    International Nuclear Information System (INIS)

    Heretofore, a pressure tube type reactor has a problem in that the evaluation for the reactor core performance is complicate and no sufficient consideration is made for the economical property, to increase the size of a calandria tank and make the cost expensive. Then, in the present invention, the inner diameter of a pressure tube is set to greater than 50% of the lattice gap in a square lattice like arrangement, and the difference between the inner and the outer diameters of the calandria tube is set smaller than 20% of the lattice gap. Further, the inner diameter of the pressure tube is set to greater than 40% and the difference between the inner and the outer diameters of the calandria tube is set smaller than 30% of the lattice gap in a triangle lattice arrangement. Then, heavy water-to-fuel volume ratio can be determined appropriately and the value for the coolant void coefficient is made more negative side, to improve the self controllability inherent to the reactor. In particular, when 72 to 90 fuel rods are arranged per one pressure tube, the power density per one fuel rod is can be increased by about twice. Accordingly, the number of the pressure tubes can be reduced about to one-half, thereby enabling to remarkably decrease the diameter of the reactor core and to reduce the size of the calandria, which is economical. (N.H.)

  11. Thermalhydraulics of advanced 37-element fuel bundle in crept pressure tubes

    Directory of Open Access Journals (Sweden)

    Park Joo Hwan

    2016-01-01

    Full Text Available A CANDU-6 reactor, which has 380 fuel channels of a pressure tube type, is suffering from aging or creep of the pressure tubes. Most of the aging effects for the CANDU primary heat transport system were originated from the horizontal crept pressure tubes. As the operating years of a CANDU reactor proceed, a pressure tube experiences high neutron irradiation damage under high temperature and pressure. The crept pressure tube can deteriorate the Critical Heat Flux (CHF of a fuel channel and finally worsen the reactor operating performance and thermal margin. Recently, the modification of the central subchannel area with increasing inner pitch length of a standard 37-element fuel bundle was proposed and studied in terms of the dryout power enhancement for the uncrept pressure tube since a standard 37-element fuel bundle has a relatively small flow area and high flow resistance at the central region. This study introduced a subchannel analysis for the crept pressure tubes loaded with the inner pitch length modification of a standard 37-element fuel bundle. In addition, the subchannel characteristics were investigated according to the flow area change of the center subchannels for the crept pressure tubes. Also, it was discussed how much the crept pressure tubes affected the thermalhydraulic characteristics of the fuel channel as well as the dryout power for the modification of a standard 37-element fuel bundle.

  12. Pressure tubes cracking due to DHC mechanism

    International Nuclear Information System (INIS)

    Zr-2.5wt%Nb alloy, used in fabrication CANDU and RMBK pressure tubes, fulfils the requirements of a material to be used under specific thermal, mechanical, irradiation and corrosion environment conditions in a nuclear reactor. Despite these advantages, the structural integrity of this assembly can be affected under certain conditions (stress, temperature, hydrogen concentration above the terminal limit of solubility), the crack initiation and propagation process being the mechanism responsible of this behaviour. During their operation the pressure tubes are susceptible to a stable cracking process referred to as Delayed Hydride Cracking (DHC). This phenomenon is one of the most important factors responsible for the degradation of these reactor components. The hydrogen concentration and the stress distribution are the parameters affecting this mechanism, leading to an embrittlement effect of the material, to a loss in the ductility and in the fracture toughness. Therefore, the structural integrity and the in-service lifetime are affected. The pressure tubes fabricated from zirconium alloy occlude exothermically hydrogen during manufacture and during operational service. When the concentration in solution exceeds the TLS (Terminal Limit of Solubility), the excess hydrogen precipitates as platelets of hydride. These hydrides are generally brittle in nature and therefore they cause a structural embrittlement, a loss of ductility and fracture toughness. Under certain thermo-mechanical conditions, the hydrides oriented after the fabrication process in the circumferential direction, tend to reorient. This phenomenon is responsible for a time-dependent failure through the hydride zone and it will arrest in the ductile zirconium matrix. This paper presents the experimental results obtained as a part of the IAEA Co-ordinated Research Project on 'Hydrogen and Hydride Induced Degradation of the Mechanical and Physical Properties of Zirconium-based Alloys', included like a

  13. Experimental modeling of flow-induced vibration of multi-span U-tubes in a CANDU steam generator

    International Nuclear Information System (INIS)

    Flow-induced vibration of the tubes in a nuclear steam generator is a concern for designers who are trying to increase the life span of these units. The dominant excitation mechanisms are fluidelastic instability and random turbulence excitation. The outermost U-bend region of the tubes is of greatest concern because the flow is almost perpendicular to the tube axis and the unsupported span is relatively long. The support system in this region must be well designed in order to minimize fretting wear of the tubes at the support locations. Much of the previous testing was conducted on straight single-span or cantilevered tubes in cross-flow. However, the dynamic response of steam generator multi-span U-tubes with clearance supports is expected to be different. Accurate modeling of the tube dynamics is important to properly simulate the dynamic interaction of the tube and supports. This paper describes a test program that was developed to measure the dynamic response of a bundle of steam generator U-tubes with Anti-Vibration Bar (AVB) supports, subjected to Freon two-phase cross-flow. The tube bundle has similar geometrical conditions to those expected for future CANDU steam generators. Future steam generators will be larger than previous CANDU steam generators, nearly twice the heat transfer area, with significant changes in process conditions in the U-bend region, such as increased steam quality and a broader range of flow velocities. This test program was initiated at AECL to demonstrate that the tube support design for future CANDU steam generators will meet the stringent requirements associated with a 60 year design life. The main objective of the tests is to address the issue of in-plane and out-of-plane fluidelastic instability and random turbulent excitation of a U-tube bundle with Anti-Vibration Bar (AVB) supports. Details of the test rig, measurement techniques and preliminary instrumentation results are described in the paper. (author)

  14. Economics of CANDU-PHW

    International Nuclear Information System (INIS)

    The CANDU-Pressurized Heavy Water (CANDU-PHW) type of nuclear-electric generating station has been developed jointly by Atomic Energy of Canada Limited and Ontario Hydro. This paper discusses the cost of producing electricity from CANDU, presents actual cost experience of CANDU and coal in Ontario, presents projected CANDU and coal costs in Ontario and compares CANDU and Light Water Reactor cost estimates in Ontario

  15. Pressure Tube and Pressure Vessel Reactors; certain comparisons

    International Nuclear Information System (INIS)

    In a comparison between pressure tube and pressure vessel type reactors for pressurized D2O coolant and natural uranium, one can say that reactors of these two types having the same net electrical output, overall thermal efficiency, reflected core volume and fuel lattice have roughly the same capital cost. In these circumstances, the fuel burn-up obtainable has a significant influence on the relative economics. Comparisons of burn-up values made on this basis are presented in this report and the influence on the results of certain design assumptions are discussed. One of the comparisons included is based on the dimensions and ratings proposed for CANDU. Moderator temperature coefficients are compared and differences in kinetic behaviour which generally result in different design philosophies for the two types are mentioned, A comparison of different methods of obtaining flux flattening is presented. The influence of slight enrichment and other coolants, (boiling D2O and gases) on the comparison between pressure tube and pressure vessel designs is discussed and illustrated with comparative designs for 400 MW electrical output. This paper was presented at the EAES Enlarged Symposium on Heterogeneous Heavy Water Power Reactors, Mallorca, October 10 - 14, 1960

  16. CANDU operating experience

    International Nuclear Information System (INIS)

    The CANDU Pressurized Heavy Water (CANDU-PHW) type of nuclear electric generating station has been developed jointly by Atomic Energy of Canada Limited and Ontario Hydro. This paper highlights Ontario Hydro's operating experience using the CANDU-PHW system, with a focus on the operating performance and costs, reliability of system components and nuclear safety considerations both to the workers and the public

  17. Two-phase natural circulation experiments in a pressurized water loop with CANDU geometry

    International Nuclear Information System (INIS)

    To provide information on two-phase natural circulation in a CANDU-type coolant circuit a series of tests has been performed in the RD-12 loop at the Whiteshell Nuclear Research Establishment. RD-12 is a 10-MPa pressurized-water loop containing two active boilers, two pumps, and two, or four, heated horizontal channels arranged in a symmetrical figure-of-eight configuration characteristic of the CANDU reactor primary heat-transport system. In the tests, single-phase natural circulation was established in the loop and void was introduced by controlled draining, with the surge tank (pressurizer) valved out of the system. The paper reviews the experimental results obtained and describes the evolution of natural circulation flow in particular cases as voidage is progressively increased. The stability behaviour is discussed briefly with reference to a simple stability model

  18. Determination of dislocation density in Zr-2.5Nb pressure tubes by x-ray

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Suk; Isaenkova, Perlovich; Cheong, Y. M.; Kim, S. S.; Yim, K. S.; Kwon, Sang Chul

    2000-11-01

    For X-ray determination of the dislocation density in CANDU Zr-2.5%Nb pressure tubes, a program was developed, using the Fourier analysis of X-ray line profiles and calculation of dislocation density by values of the coherent block size and the lattice distortion. The coincidence of obtained values of c- and a-dislocations with those, determined by the X-ray method for the same tube in AECL, was assumed to be the main criterion of validity of the developed program. The final variant of the program allowed to attain a rather close coincidence of calculated dislocation densities with results of AECL. The dislocation density was determined in all the zirconium grains with different orientations based on the texture of the stree-relieved CANDU tube. The complete distribution of c-dislocation density in -Zr grains depecding on their crystallographic orientations was constructed. The distribution of a-dislocation density within the texture maximum at L-direction, containing prismatic axes of all grains, was constructed as well. The analysis of obtained distributions testifies that -Zr grains of the stree-relieved CANDU tube significantly differ in their dislocation densities. Plotted diagrams of correlation between the dislocation density and the pole density allow to estimate the actual connection between texture and dislocation distribution in the studied tube. The distributions of volume fractions of all the zirconium grains depending on their dislocation density were calculated both for c- and a-dislocations. The distributions characterizes quantitatively the inhomogeneity of substructure conditions in the stress-relieved CANDU tube. the optimal procedure for determination of Nb content in {beta}-phases of CANDU Zr-2.5%Nb pressure tubes was also established.

  19. Development of delayed hydride cracking resistant-pressure tube

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Suk; Kwon, Sang Chul; Kim, S. S.; Yim, K. S

    2000-10-01

    For the first time, we demonstrate that the pattern of nucleation and growth of a DHC crack is governed by the precipitation of hydrides so that the DHC velocity and K{sub IH} are determined by an angle of the cracking plane and the hydride habit plane 10.7. Since texture controls the distribution of the 10.7 habit plane in Zr-2.5Nb pressure tube, we draw a conclusion that a textural change in Zr-2.5Nb tube from a strong tangential texture to the radial texture shall increase the threshold stress intensity factor, K{sub IH}, and decrease the delayed hydride cracking velocity. This conclusion is also verified by a complimentary experiment showing a linear dependence of DHCV and K{sub IH} with an increase in the basal component in the cracking plane. On the basis of the study on the DHC mechanism and the effect of manufacturing processes on the properties of Zr-2.5Nb tube, we have established a manufacturing procedure to make pressure tubes with improved DHC resistance. The main features of the established manufacturing process consist in the two step-cold pilgering process and the intermediate heat treatment in the {alpha} + {beta} phase for Zr-2.5Nb alloy and in the {alpha} phase for Zr-1Nb-1.2Sn-0.4Fe alloy. The manufacturing of DHC resistant-pressure tubes of Zr-2.5Nb and Zr-1N-1.2Sn-0.4Fe was made in the ChMP zirconium plant in Russia under a joint research with Drs. Nikulina and Markelov in VNIINM (Russia). Zr-2.5Nb pressure tube made with the established manufacturing process has met all the specification requirements put by KAERI. Chracterization tests have been jointly conducted by VNIINM and KAERI. As expected, the Zr-2.5Nb tube made with the established procedure has improved DHC resistance compared to that of CANDU Zr-2.5Nb pressure tube used currently. The measured DHC velocity of the Zr-2.5Nb tube meets the target value (DHCV <5x10{sup -8} m/s) and its other properties also were equivalent to those of the CANDU Zr-2.5Nb tube used currently. The Zr-1Nb-1

  20. Analysis of the pressure tube failure at Pickering NGS A unit 2

    International Nuclear Information System (INIS)

    The failure of a Zircaloy-2 pressure tube in Pickering Unit 2 in August 1983 has been found to have been caused by an accelerated pickup rate of deuterium, contact between the pressure tube and its surrounding calandria tube, and rapid growth of zirconium hydride blisters. The pressure tubes in all later CANDU reactors are made from zirconium- niobium alloy. Examination of several Zircaloy-2 and zirconium niobium pressure tubes from different reactors has clearly shown that the deuterium pickup rate of the two materials is significantly different. The zirconium niobium pressure tubes have absorbed very little deuterium and they do not appear to be susceptible to the type of failure experienced at Pickering Unit 2

  1. The influence of heating rate on the pressure tube microstructure

    International Nuclear Information System (INIS)

    The aim of this paper is the study of the influence of heating rate on the microstructure morphology developed in pressure tube (Zr-2, 5% Nb alloy) at temperatures in the thermal transient conditions similar to LOCA (Loss of Coolant Agent) accident. The samples were tested using some thermal transient scenarios with different rates of heating. The thermal transients are performed at temperatures between 300 deg. C and 900 deg. C with the following heating rates: 5 deg. C/s, 10 deg. C/s and 20 deg. C/s. The average grain size and microhardness determination were performed in cross and longitudinal sections on the blank and tested samples. The analysis of microphotographs and data is presented as figures and diagrams. Some results of these experiments can be used in the study of degradation of the mechanical and micro-structural properties of the pressure tube materials in CANDU reactors. (authors)

  2. Licensing assessment of the Candu Pressurized Heavy Water Reactor. Preliminary safety information document. Volume II. [USA

    Energy Technology Data Exchange (ETDEWEB)

    1977-06-01

    ERDA has requested United Engineers and Constructors (UE and C) to evaluate the design of the Canadian natural uranium fueled, heavy water moderated (CANDU) nuclear reactor power plant to assess its conformance with the licensing criteria and guidelines of the U.S. Nuclear Regulatory Commission (USNRC) for light water reactors. This assessment was used to identify cost significant items of nonconformance and to provide a basis for developing a detailed cost estimate for a 1140 MWe, 3-loop Pressurized Heavy Water Reactor (PHWR) located at the Middletown, USA Site.

  3. Licensing assessment of the Candu Pressurized Heavy Water Reactor. Preliminary safety information document. Volume II

    International Nuclear Information System (INIS)

    ERDA has requested United Engineers and Constructors (UE and C) to evaluate the design of the Canadian natural uranium fueled, heavy water moderated (CANDU) nuclear reactor power plant to assess its conformance with the licensing criteria and guidelines of the U.S. Nuclear Regulatory Commission (USNRC) for light water reactors. This assessment was used to identify cost significant items of nonconformance and to provide a basis for developing a detailed cost estimate for a 1140 MWe, 3-loop Pressurized Heavy Water Reactor (PHWR) located at the Middletown, USA Site

  4. Economics of CANDU

    International Nuclear Information System (INIS)

    The cost of producing electricity from CANDU reactors is discussed. The total unit energy cost of base-load electricity from CANDU reactors is compared with that of coal-fired plants in Ontario. In 1980 nuclear power was 8.41 m$/kW.h less costly for plants of similar size and vintage. Comparison of CANDU with pressurized water reactors indicated that the latter would be about 26 percent more costly in Ontario

  5. Ontario Hydro CANDU operating experience

    International Nuclear Information System (INIS)

    The CANDU Pressurized Heavy Water (CANDU-PHW) type of nuclear-electric generating station has been developed jointly by Atomic Energy of Canada Limited and Ontario Hydro. This report highlights Ontario Hydro's operating experience using the CANDU-PHW system, with a focus on the operating performance and costs, reliability of system components and nuclear safety considerations for the workers and the public

  6. Ontario Hydro CANDU operating experience

    International Nuclear Information System (INIS)

    The CANDU Pressurized Heavy Water (CANDU-PHW) type of nuclear-electric generating station has been developed jointly by Atomic Energy of Canada Limited and Ontario Hydro. This report highlights Ontario Hydro's operating experience using the CANDU-PHW system, with a focus on worker and public safety, operating performance and costs, and reliability of system components

  7. CANDU steam generator life management

    International Nuclear Information System (INIS)

    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)

  8. Development of the safety regulatory guides on the refurbishment for the CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, M. W.; Chin, T. E.; Rho, H. Y.; Park, H. B.; Yeom, H. G.; Hwang, G. M.; Hwang, B. G.; Seo, Y. H.; Lee, J. W. [Korea Power Engineering Co. Inc., Yongin (Korea, Republic of)

    2007-02-15

    In this study, requirements and standards concerned with safety performance for CANDU type reactors and review guidelines for facilities and performance concerned with refurbishment of major facilities such as pressure tubes, calandria tubes, and feeder popes were developed. To develop review guidelines for facilities and performance review concerned with refurbishment of CANDU reactors, review activities related with refurbishment and performance were categorized into designing and planning of equipments, removal and refurbishment of equipment, and confirmation of installation and inspection. As a result, following detailed review guidelines concerned with refurbishment of pressure tubes, calandria tubes, and feeder pipes in directly or indirectly referring to FSAR, design manual, startup-test manual were developed.

  9. Inert medium (helium) irradiation testing of pressure tube samples

    International Nuclear Information System (INIS)

    Irradiation tests currently performed in C-5 capsule aim at obtaining data and information concerning behavior to irradiation of pressure tubes of CANDU type fuel channel, to evidence the factors limiting operation life span. A calculation code for analysis and prediction of pressure tube behavior should be based upon periodical inspection results, post irradiation examination of the removed from reactor pressure tubes as well as on the experimental results obtained with materials subjected to irradiation conditions identical with the operational ones. Mechanical behavior analysis should focus both complex thermal-mechanical type stresses and mechanical properties alteration under irradiation. The experimental results should be applied: - to evaluate the irradiation effects upon mechanical properties of Zr-2.5% Nb exposed to fluences up to 1021 n·cm-2; - to gather data concerning the real stress / real deformation characteristic from which characteristic quantities can be deduced as, for instance, elasticity modulus, plasticity modulus, exponent of stress term in the Tsu-Berteles relation, to be used within the CANTUP simulation code describing pressure tube behavior, currently developed at INR Pitesti; - to develop prediction methods of pressure tube behavior and merging with in-service inspection procedure in order to forecast the life span and the proper timing for replacement before major failures occur. The samples irradiated in C-5 capsule were extracted from the ends of Zr-2.5% Nb pressure tubes resulting from Cernavoda NPP Unit 1. The samples for tensile tests were extracted on longitudinal and transversal directions of the pressure tube. The tests were carried out under following conditions: - test environment temperature, 260 - 280 deg.C; - testing medium, helium at 1 - 6 b pressure; - neutron flux (En > 1 MeV), 1 - 2 · 1013 ncm-2s-1; - neutron fluence (En > 1 MeV), 4 · 1020 ncm-2. The following characteristics were obtained from tensile test: - real

  10. Distortion Of Pressure Signals In Pneumatic Tubes

    Science.gov (United States)

    Whitmore, Stephen A.; Gilyard, Glenn B.; Curry, Robert; Lindsey, William

    1993-01-01

    NASA technical memorandum describes experimental investigation of distorting effects of propagation of pressure signals along narrow pneumatic tubes from pressure-sensing orifices on surfaces of models or aircraft to pressure sensors distant from orifices. Pressure signals distorted principally by frictional damping along walls of tubes and by reflections at orifice and sensor ends.

  11. Development of Regulatory Requirements and Inspection Guides for CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, M. W.; Kim, K.; Ryu, Y. H. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Ro, H. Y.; Jin, T. E. [Korea Power Engineering Co., Yongin (Korea, Republic of)

    2009-05-15

    The first domestic CANDU power reactor, Wolsong unit 1, has been operated for about twenty years since commercial operation in 1983, and has been raised common aging issues of CANDU reactors in pressure tubes, calandria tubes, feeder pipes, etc. To solve these aging issues, utility is promoting the refurbishment activities for these major components. Therefore, confirmation and improvement for insufficient requirements considering the CNSC regulatory documents, regulatory principles between regulatory body and utilities related with refurbishment activities are required. These review contents are described herein, and representative review results are presented.

  12. CANDU safety under severe accidents

    International Nuclear Information System (INIS)

    The characteristics of the CANDU reactor relevant to severe accidents are set first by the inherent properties of the design, and second by the Canadian safety/licensing approach. The pressure-tube concept allows the separate, low-pressure, heavy-water moderator to act as a backup heat sink even if there is no water in the fuel channels. Should this also fail, the calandria shell itself can contain the debris, with heat being transferred to the water-filled shield tank around the core. Should the severe core damage sequence progress further, the shield tank and the concrete reactor vault significantly delay the challenge to containment. Furthermore, should core melt lead to containment overpressure, the containment behaviour is such that leaks through the concrete containment wall reduce the possibility of catastrophic structural failure. The Canadian licensing philosophy requires that each accident, together with failure of each safety system in turn, be assessed (and specified dose limits met) as part of the design and licensing basis. In response, designers have provided CANDUs with two independent dedicated shutdown systems, and the likelihood of Anticipated Transients Without Scram is negligible. Probabilistic safety assessment studies have been performed on operating CANDU plants, and on the 4 x 880 MW(e) Darlington station now under construction; furthermore a scoping risk assessment has been done for a CANDU 600 plant. They indicate that the summed severe core damage frequency is of the order of 5 x 10-6/year. 95 refs, 3 tabs

  13. Modelling nuclear fuel vibrations in horizontal CANDU reactors

    International Nuclear Information System (INIS)

    Flow-induced fuel vibrations in the pressure tubes of CANDU reactors are of vital interest to designers because fretting damage may result. Computer simulation is being used to study how bundles vibrate and to identify bundle design features which will reduce vibration and hence fretting. (author)

  14. Economic Analysis on Direct Use of Spent Pressurized Water Reactor Fuel in CANDU Reactors - I: DUPIC Fuel Fabrication Cost

    International Nuclear Information System (INIS)

    A preliminary conceptual design of a Direct Use of spent Pressurized water reactor (PWR) fuel In Canada deuterium uranium (CANDU) reactors (DUPIC) fuel fabrication plant was studied, which annually converts spent PWR fuel of 400 tonnes heavy element (HE) into CANDU fuel. The capital and operating costs were estimated from the viewpoint of conceptual design. Assuming that the annual discount rate is 5% during the construction (5 yr) and operation period (40 yr) and contingency is 25% of the capital cost, the levelized unit cost (LUC) of DUPIC fuel fabrication was estimated to be 616 $/kg HE, which is mostly governed by annual operation and maintenance costs that correspond to 63% of LUC. Among the operation and maintenance cost components being considered, the waste disposal cost has the dominant effect on LUC (∼49%). From sensitivity analyses of production capacity, discount rate, and contingency, it was found that the production capacity of the plant is the major parameter that affects the LUC

  15. An emergency water injection system (EWIS) for future CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Marques, Andre L.F. [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil). E-mail: momarques@uol.com.br; Todreas, Neil E.; Driscoll, Michael J. [Massachusetts Inst.of Tech., Cambridge, MA (United States). Nuclear Engineering Dept.

    2000-07-01

    This paper deals with the investigation of the feasibility and effectiveness of water injection into the annulus between the calandria tubes and the pressure tubes of CANDU reactors. The purpose is to provide an efficient decay heat removal process that avoids permanent deformation of pressure tubes severe accident conditions, such as loss of coolant accident (LOCA). The water injection may present the benefit of cost reduction and better actuation of other related safety systems. The experimental work was conducted at the Massachusetts Institute of Technology (MIT), in a setup that simulated, as close as possible, a CANDU bundle annular configuration, with heat fluxes on the order of 90 kW/m{sup 2}: the inner cylinder simulates the pressure tube and the outer tube represents the calandria tube. The experimental matrix had three dimensions: power level, annulus water level and boundary conditions. The results achieved overall heat transfer coefficients (U), which are comparable to those required (for nominal accident progression) to avoid pressure tube permanent deformation, considering current CANDU reactor data. Nonetheless, future work should be carried out to investigate the fluid dynamics such as blowdown behavior, in the peak bundle, and the system lay-out inside the containment to provide fast water injection. (author)

  16. Research and development for CANDU fuel channels and fuel

    International Nuclear Information System (INIS)

    The CANDU nuclear reactor is distinctly different from BWR and PWR reactors in that it uses many small pressure tubes rather than one large pressure vessel to contain the fuel and coolant. To exploit the advantages of the natural uranium fuel, the pressure tubes, like other core components, are manufactured from zirconium alloys which have low neutron capture cross sections. Also, because natural uranium fuel only achieves a modest burnup, a simple and inexpensive fuel design has been developed. The present paper reviews the features and the research that have led to the very satisfactory performance of the pressure tubes and the fuel in CANDU reactors. Reference is made to current research and development that may lead to further economies in the design and operation of future power reactors. (author)

  17. Development of Zirconium alloys (for pressure tubes)

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Suk; Kwon, Sang Chul; Choo, Ki Nam; Jung, Chung Hwan; Yim, Kyong Soo; Kim, Sung Soo; Baek, Jong Hyuk; Jeong, Yong Hwan; Kim, Kyong Ho; Cho, Hae Dong [Korea Atomic Energy Research Inst., Daeduk (Korea, Republic of); Hwang, S. K.; Kim, M. H. [Inha Univ., Incheon (Korea, Republic of); Kwon, S. I [Korea Univ., Seoul (Korea, Republic of); Kim, I. S. [Korea Advanced Inst. of Science and Technology, Taejon (Korea, Republic of)

    1997-09-01

    The objective of this research is to set up the basic technologies for the evaluation of pressure tube integrity and to develop improved zirconium alloys to prevent pressure tube failures due to DHC and hydride blister caused by excessive creep-down of pressure tubes. The experimental procedure and facilities for characterization of pressure tubes were developed. The basic research related to a better understanding of the in-reactor performances of pressure tubes leads to noticeable findings for the first time : the microstructural effect on corrosion and hydrogen pick-up behavior of Zr-2.5Nb pressure tubes, texture effect on strength and DHC resistance and enhanced recrystallization by Fe in zirconium alloys and etc. Analytical methodology for the assessment of pressure tubes with surface flaws was set up. A joint research is being under way with AECL to determine the fracture toughness of O-8 at the EOL (End of Life) that had been quadruple melted and was taken out of the Wolsung Unit-1 after 10 year operation. In addition, pressure tube with texture controlled is being made along with VNINM in Russia as a joint project between KAERI and Russia. Finally, we succeeded in developing 4 different kinds of zirconium alloys with better corrosion resistance, low hydrogen pickup fraction and higher creep strength. (author). 121 refs., 65 tabs., 260 figs

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

    International Nuclear Information System (INIS)

    This is a PowerPoint presentation on behalf of COMPCONTROL ING, a Romanian private company established in 1997 the main services of which are enlisted. It is stressed that the most suitable type of inspection in terms of safety and reliability for the steam generator tubes is eddy current (EC) method. The advantages of EC testing include the following: - Extremely fast; - Accurate in detection and sizing of discontinuities; - Very good method for baseline screening; - Very high detection sensitivity to physical-chemical variations of the test specimen; - Easy setup and application for automated inspection; - Portable equipment designing; - Use of multiple channels and multi-frequencies for a better screening of signals and efficiency; - High capability to store the data for future review and comparison (using data history to evaluate the rate of degradation and life assessment studies). Between 2003 and 2005 ECT was applied to Cernavoda NPP U1 SGs as follows: - in 2003, SG-4; - in 2004, SG-2; - in 2005, SG-1; - in 2005, SG-3; - in 2005, SG-4. The purpose of inspection with eddy currents of SGs tubes was: - Detection, sizing and evaluation of possible degradations of the tubes and at the interface tube/support structures (tubesheet, tube support plates and baffles); - Completion of the baseline data for future review and comparison. The software used for acquisition and analysis of eddy current data and for inspection management were: - ZETEC Eddynet-R Zetec Acquisition Control-ZAC; - ZETEC Eddynet-R Data Analysis (bobbin and MRPC); - ZETEC Eddynet-R Data Management. The equipment ZR-1 is described and its advantages as well. Advantages of the automated scanning system are highlighted as follows: - Repeatability; - High resolution mapping; - Accurate indexing; - Minimize changes in lift-off resulting from probe wobble, eccentricity of the tube and surface irregularities; - 3-part design makes each component lighter and more compact for easier, faster installation

  19. Pressure tube rupture in a closed tank

    International Nuclear Information System (INIS)

    A study has been prepared on the feasibility of conducting pressure tube/calandria tube rupture tests in a closed tank, simulating a scaled-down calandria vessel. The study includes: i) a review of previous work, ii) an analytical investigation of the scaling problem of the calandria vessel and relevant in-core structures, iii) selection of a method for initiating pressure tube/calandria tube rupture, iv) a set of specifications for the test assembly, v) general arrangement drawings, vi) a proposal for a test matrix, vii) a survey and evaluation of existing facilities which could provide the required high pressure, temperature and fluid inventory, and viii) a cost estimate for the detailed design and construction, instrumentation, data acquisition and reduction, testing and reporting. The study concludes that it is both technically and practically feasible to conduct pressure tube rupture tests in a closed tank

  20. Probabilities of failure of a seam-welded calandria tube after a spontaneous pressure tube rupture

    International Nuclear Information System (INIS)

    This paper describes a methodology for calculating probabilities of seam-welded calandria tube (CT) failure after a sudden pressure tube (PT) rupture for operating conditions of CANDU reactors. Such a calculation is required in certain analyses or design option assessments such as those related to the issue of PT failure with consequential loss of moderator coincident with loss of emergency coolant injection (ECI). This accident scenario is the subject of Generic Action Item (GAl) 95G02 that was raised by the Canadian Nuclear Safety Commission (CNSC) in 1995. The CT failure probabilities were required as part of the resolution process for GAl-95G02. Two modes of CT failure considered are the prompt failure and the delayed creep failure. During the first half-second of CT pressurization by a spontaneous PT rupture, the plausible failure mechanism is the CT circumferential strain caused by a water-hammer type overpressure transient. The probability of prompt CT failure is calculated using a distribution of the measured failure strains for irradiated CTs and the calculated maximum CT strains resulting from water-hammer overpressure transients. If the CT survives the initial transient loading, the CT becomes the temporary pressure boundary to the primary heat transport system. Under certain pressure and temperature conditions the CT can experience slow-strain-rate plastic deformation and eventually fail by plastic strain. A time-to-rupture model is used to calculate the probability of this delayed creep failure within 5 to 15 minutes after a PT rupture. Then, the CT failure probability is calculated by combining prompt failure with delayed creep failure. (author)

  1. Evolution of CANDU vacuum building and pressure relief structures from Pickering NGS A to Darlington NGS A

    International Nuclear Information System (INIS)

    The vacuum building (VB) and pressure relief structures (PRS) are the unique features of multiple unit CANDU containments. In case of loss-of-coolant accident, the released radionuclides are drawn through the PRS into the subatmospheric VB, doused and contained without being released to the environment. This paper describes the differences in design, configuration and layout of the VB and PRS from Pickering NGS A to Darlington NGS A due to new developments in design concepts and to requirements which have proceeded from the experience gained in both the design and operation of the nuclear stations. (orig.)

  2. Advanced NDE (ANDE) and its application for pressure tube inspections in OPG reactors

    International Nuclear Information System (INIS)

    Periodic and in-service inspections of CANDU fuel channels are essential for the proper assessment of the structural integrity of these vital components. The arrival of new delivery devices for fuel channel inspections (Universal Delivery Machine) has driven new methods for gathering and analyzing NDE data. The Advanced Non-Destructive Examination (ANDE) system has been designed and field implemented as a high speed data acquisition system to meet the requirements of the CSA N285.4 code. It was built from the solid foundation of CIGAR experience and uses cutting edge hardware and software to attain high speed data collection enabling relatively quick inspection of a large number of fuel channels. The capabilities of the ANDE inspection system include: Surface and volumetric inspection of pressure tube by ultrasonics; Flaw characterization by ultrasonics; Pressure tube diameter measurements; Pressure tube thickness measurements; Garter Spring location by Eddy Current; Garter Spring location by ultrasonics; Pressure tube sag measurement. In addition to the above, selected flaws/areas of a pressure tube can be replicated using a two plate ANDE replica tool. At the heart of the inspection system is a set of twelve ultrasonic probes positioned in such a way that the inspected areas are examined from various angles and directions and by various ultrasonic wave modes (shear and longitudinal). High frequency ultrasound used for the examinations allows for reliable detection of small flaws. Separate sensors have been installed on the inspection head for Garter Spring location and sag measurements. (author)

  3. Heavy water cycle in the CANDU reactor

    International Nuclear Information System (INIS)

    Hydrogen atom has two isotopes: deuterium 1H2 and tritium 1H3. The deuterium oxide D2O is called heavy water due to its density of 1105.2 Kg/m3. Another important physical property of the heavy water is the low neutron capture section, suitable to moderate the neutrons into natural uranium fission reactor as CANDU. Due to the fact that into this reactor the fuel is cooled into the pressure tubes surrounded by a moderator, the usage of D2O as primary heat transport (PHT) agent is mandatory. Therefore a large amount of heavy water (approx. 500 tons) is used in a CANDU reactor. Being a costly resource - it represents 20% of the initial plant capital cost, D2O management is required to preserve it. (author)

  4. Eddy current monitoring of fatigue crack growth in Zr-2.5% Nb pressure tube

    Science.gov (United States)

    Krause, T. W.; Martin, A. E.; Sheppard, R. R.; Schankula, J. J.

    2000-05-01

    Zr-2.5% wt. Nb pressure tubes (PTs) form the core of the heat transport system in CANDU nuclear reactors. These 6 m long, 100 mm diameter tubes are operated at elevated temperatures (nominally 300 °C) and at pressures that produce hoop stresses that are 25% of the ultimate tensile strength of the PT (120 Mpa). Therefore, detection and characterization of flaws in these components becomes crucial for their continued pressure retaining integrity. If a flaw is detected, however, the cost of PT replacement is expensive. Periodic in-service inspection of a flaw that demonstrates no change in flaw characteristics can be used to allow a pressure tube to remain in-service. This requires confidence in the accuracy and reliability of methods used to inter flaw characteristics. Such confidence can only be developed by comparing nondestructive predictions with results from destructive examinations. In this work, eddy current testing was used to monitor the progressive stages of a fatigue crack, grown through pressure cycling from a notch on the inner surface of a PT. Results from a differential lift-off compensated eddy current probe were used to produce sizing estimates of the crack grown between 35% (base of notch) and 74% of the PT wall. A comparison with a destructive examination of the crack demonstrated sensitivity too changes in crack depth accurate to 5% of the tube wall thickness. Such results, combined with similar information obtained from ultrasonics will increase confidence in interpretation of PT inspection data.

  5. Inspection of Candu Nuclear Reactor Fuel Channels

    International Nuclear Information System (INIS)

    The Channel Inspection and Gauging Apparatus of Reactors (CIGAR) is a fully atomated, remotely operated inspection system designed to perform multi-channel, multi-task inspection of CANDU reactor fuel channels. Ultrasonic techniques are used for flaw detection, (with a sensitivity capable of detecting a 0.075 mm deep notch with a signal to noise ratio of 10 dB) and pressure tube wall thickness and diameter measurements. Eddy currrent systems are used to detect the presence of spacers between the coaxial pressure tube and calandria tube, as well as to measure their relative spacing. A servo-accelerometer is used to estimate the sag of the fuel channels. This advanced inspection system was commissioned and declared in service in September 1985. The paper describes the inspection systems themselves and discussed the results achieved to-date. (author)

  6. CANDU safety analysis system establishment

    Energy Technology Data Exchange (ETDEWEB)

    Min, Byung Joo; Rhee, B. W.; Park, J. H.; Kim, H. T.; Choi, H. B.; Shim, J. I.; Yoon, C.; Yang, M. K

    2002-03-01

    To develop CANDU safety analysis system, methodology, and assessment technology, GAIs from CNSC and GSIs drived by IAEA are summarized. Furthermore, the following safety items are investigated in the present study. - It is intended to secure credibility of the void reactivity in the stage of nuclear design and analysis. The measurement data concerned with the void reactivity were reviewed and used to assess the physics code such as POWDERPUFS-V/RFSP, and the lattice code such as WIMS-AECL and MCNP-4B. - Reviewing the Final Safety Analysis Report for Wolsong-2/3/4 Units, the safety analysis methodology, classification for accident scenarios, safety analysis codes, their interface, etc. were examined. - The development of 3D CFD transient analysis model has been performed to predict local subcooling of the moderator in the vicinity of Calandria tubes in a CANDU-6 reactor in the case of Large LOCA transient. - The trip coverage analysis methodology based on CATHENA code is developed. The simulation of real plant transient showed good agreement. The trip coverage map was generated successfully for two typical depressurization and pressurization event. - The multi-dimensional analysis methodology for hydrogen distribution and hydrogen burning phenomena in PHWR containment is developed using GOTHIC code. The multi-dimensional analysis predicts the local hydrogen behaviour compared to the lumped parameter model.

  7. Characterization of elastic properties of Zr-2.5% Nb pressure tube by measurements of sound velocity

    International Nuclear Information System (INIS)

    The cold-worked Zr-2.5% Nb alloy is used as material for the pressure tubes of CANDU (CANadian Deuterium Uranium) nuclear reactors. During the service life in reactor, diffusion of hydrogen and/or deuterium in the pressure tubes wall will occur. Below a certain temperature, a stable hydride of zirconium will form, as a brittle phase which could lead to catastrophic failure. In the present paper, the influence of hydrogen on the acoustic-elastic properties of Zr-2.5% Nb alloy will be investigated using non-destructive method based on measurements of ultrasonic velocity. In order to obtain the most usual elastic coefficients on a given direction (axial and circumferential) of the tube, both longitudinal VL and transversal VT phase velocities have been experimentally determined. (authors)

  8. Basic research and industrialization of CANDU advanced fuel

    International Nuclear Information System (INIS)

    Wolsong Unit 1 as the first heavy water reactor in Korea has been in service for 17 years since 1983. It would be about the time to prepare a plan for the solution of problems due to aging of the reactor. The aging of CANDU reactor could lead especially to the steam generator cruding and pressure tube sagging and creep and then decreases the operation margin to make some problems on reactor operations and safety. The counterplan could be made in two ways. One is to repair or modify reactor itself. The other is to develop new advanced fuel to increase of CANDU operation margin effectively, so as to compensate the reduced operation margin. Therefore, the first objectives in the present R and D is to develop the CANFLEX-NU (CANDU Flexible fuelling-Natural Uranium) fuel as a CANDU advanced fuel. The second objectives is to develop CANDU advanced fuel bundle to utilize advanced fuel cycles such as recovered uranium, slightly enriched uranium, etc. and so to raise adaptability for change in situation of uranium market. Also, it is to develop CANDU advanced fuel technology which improve uranium utilization to cope with a world-wide imbalance between uranium supply and demand, without significant modification of nuclear reactor design and refuelling strategies. As the implementations to achieve the above R and D goal, the work contents and scope of technology development of CANDU advanced fuel using natural uranium (CANFLEX-NU) are the fuel element/bundle designs, the nuclear design and fuel management analysis, the thermalhydraulic analysis, the safety analysis, fuel fabrication technologies, the out-pile thermalhydraulic test and in-pile irradiation tests performed. At the next, the work scopes and contents of feasibility study of CANDU advanced fuel using recycled uranium (CANFLEX-RU) are the fuel element/bundle designs, the reactor physics analysis, the thermalhydraulic analysis, the basic safety analysis of a CANDU-6 reactor with CANFLEX-RU fuel, the fabrication and

  9. Basic research and industrialization of CANDU advanced fuel

    Energy Technology Data Exchange (ETDEWEB)

    Chun, Suk Ho; Park, Joo Hwan; Jun, Ji Su [and others

    2000-04-01

    Wolsong Unit 1 as the first heavy water reactor in Korea has been in service for 17 years since 1983. It would be about the time to prepare a plan for the solution of problems due to aging of the reactor. The aging of CANDU reactor could lead especially to the steam generator cruding and pressure tube sagging and creep and then decreases the operation margin to make some problems on reactor operations and safety. The counterplan could be made in two ways. One is to repair or modify reactor itself. The other is to develop new advanced fuel to increase of CANDU operation margin effectively, so as to compensate the reduced operation margin. Therefore, the first objectives in the present R and D is to develop the CANFLEX-NU (CANDU Flexible fuelling-Natural Uranium) fuel as a CANDU advanced fuel. The second objectives is to develop CANDU advanced fuel bundle to utilize advanced fuel cycles such as recovered uranium, slightly enriched uranium, etc. and so to raise adaptability for change in situation of uranium market. Also, it is to develop CANDU advanced fuel technology which improve uranium utilization to cope with a world-wide imbalance between uranium supply and demand, without significant modification of nuclear reactor design and refuelling strategies. As the implementations to achieve the above R and D goal, the work contents and scope of technology development of CANDU advanced fuel using natural uranium (CANFLEX-NU) are the fuel element/bundle designs, the nuclear design and fuel management analysis, the thermalhydraulic analysis, the safety analysis, fuel fabrication technologies, the out-pile thermalhydraulic test and in-pile irradiation tests performed. At the next, the work scopes and contents of feasibility study of CANDU advanced fuel using recycled uranium (CANFLEX-RU) are the fuel element/bundle designs, the reactor physics analysis, the thermalhydraulic analysis, the basic safety analysis of a CANDU-6 reactor with CANFLEX-RU fuel, the fabrication and

  10. Uses of Plutonium Fuel in Pressure-Tube-Type, Heavy-Water-Moderated Thermal Reactors

    International Nuclear Information System (INIS)

    In 1962, a feasibility study was begun in the JAERI on the uses of various nuclear fuels for pressure-tube-type, heavy-water-moderated thermal reactors. This study began with analysis of the use of uranium in heavy-water-moderated thermal reactors such as the CANDU-PHW, CANDU-BLW, SGHW, EL-4, and Ref. 15, D and E lattices, which is designed in the JAERI, from the standpoint of the core design. Then, the ways of using plutonium fuel in the same types were investigated using WATCHTOWER, FLARE and VENUS codes, including: (1) direct substitution of the plutonium from light-water reactors or Magnox reactors, (2) recycle use of the plutonium from heavy-water-moderated reactors, (3) plutonium self-sustaining cycle, and (4) plutonium phoenix fuel. The following conclusions are reported: (1) In the direct substitution of plutonium, somewhat depleted plutonium is more suitable for core design than the plutonium from Magnox reactors or light-water reactors, because the increase in the initial reactivity due to large plutonium absorption cross-section must be prevented. (2) In the plutonium self-sustaining cycle, the fuel burn-up of about 15 000 ∼20000 MWd/t would be expected from natural uranium, and the positive void reactivity which always occurs in the uraniumloaded SGHW or CANDU-BLW lattices is greatly reduced, the latter property giving some margin to bum-out heat flux. (3) It may be concluded from the fuel cycle analysis that the plutonium self-sustaining cycle is equivalent to using slightly enriched uranium (about 1.0 at.%). It may be concluded that the use of plutonium in heavy-water-moderated reactors is technologically feasible and economically advantageous. (author)

  11. Fuel condition in Canadian CANDU 6 reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hu, R.H.; Macici, N [Hydro-Quebec, Montreal, Quebec (Canada); Gibb, R. [New Brunswick Power, Lepreau, NB (Canada); Purdy, P.L.; Manzer, A.M. [Atomic Energy of Canada Limited, Mississauga, Ontario (Canada); Kohn, E. [Ontario Hydro, Toronto, Ontario (Canada)

    1997-07-01

    The cornerstone of the CANDU concept is its natural uranium fuel, and the success of its reactor operation hinges on the fuel condition in the reactor. Neutron economy, on power refuelling, and simple fuel design are among the unique characteristics of CANDU fuel. In Canadian CANDU 6 reactors (Gentilly 2 and Point Lepreau), the 37-element fuel has provided an enviable record of safe, economic and reliable plant operation for 29 reactor years to date. The fuelling cost is among the lowest in the world - a corollary of high neutron economy, simple fuel design, and judicial fuelling scheme. The reliability of fuel is high: only 21 of the 60000 bundles discharged from Gentilly 2 were confirmed defective and the five-year period from March 1992 to February 1997 saw no defect at all at Gentilly-2. Also, thanks to the inherent on-power refuelling capability and an effective defect detection and removal system, the primary coolant loops are kept extremely clean (very low activity level) - benefiting both maintenance and safety. Moreover, the inventories of fission products in the core and in the channel are maintained within the safety analysis envelope, due to on-power fuelling and sophisticated fuel management. In this paper, CANDU 6 fuel performance is reviewed against the feedback from post-irradiation examinations, and the findings from our ongoing R and D program. The results suggest that the fuel behavior m reactor are basically as originally anticipated, despite an evolutionary 3% increase in bundle uranium mass in the 1980's. For operating conditions within the CANDU 6 37-element experience, the average strains are typically 0.09%; and fission gas release, 2.7%. The UO{sub 2} fuel remains stoichiometric after irradiation. In-core measurements of pressure tube fitting are generally low. All these observations are consistent with the excellent fuel performance statistics coming out of the two Canadian CANDU 6 reactors. Additionally, this paper will briefly

  12. Fuel condition in Canadian CANDU 6 reactors

    International Nuclear Information System (INIS)

    The cornerstone of the CANDU concept is its natural uranium fuel, and the success of its reactor operation hinges on the fuel condition in the reactor. Neutron economy, on power refuelling, and simple fuel design are among the unique characteristics of CANDU fuel. In Canadian CANDU 6 reactors (Gentilly 2 and Point Lepreau), the 37-element fuel has provided an enviable record of safe, economic and reliable plant operation for 29 reactor years to date. The fuelling cost is among the lowest in the world - a corollary of high neutron economy, simple fuel design, and judicial fuelling scheme. The reliability of fuel is high: only 21 of the 60000 bundles discharged from Gentilly 2 were confirmed defective and the five-year period from March 1992 to February 1997 saw no defect at all at Gentilly-2. Also, thanks to the inherent on-power refuelling capability and an effective defect detection and removal system, the primary coolant loops are kept extremely clean (very low activity level) - benefiting both maintenance and safety. Moreover, the inventories of fission products in the core and in the channel are maintained within the safety analysis envelope, due to on-power fuelling and sophisticated fuel management. In this paper, CANDU 6 fuel performance is reviewed against the feedback from post-irradiation examinations, and the findings from our ongoing R and D program. The results suggest that the fuel behavior m reactor are basically as originally anticipated, despite an evolutionary 3% increase in bundle uranium mass in the 1980's. For operating conditions within the CANDU 6 37-element experience, the average strains are typically 0.09%; and fission gas release, 2.7%. The UO2 fuel remains stoichiometric after irradiation. In-core measurements of pressure tube fitting are generally low. All these observations are consistent with the excellent fuel performance statistics coming out of the two Canadian CANDU 6 reactors. Additionally, this paper will briefly discuss our

  13. Absorber materials in CANDU PHWRs

    International Nuclear Information System (INIS)

    In a CANDU reactor the fuel channels are arranged on a square lattice in a calandria filled with heavy water moderator. This arrangement allows five types of tubular neutron absorber devices to be located in the relatively benign environment of low pressure, low temperature heavy water between neighbouring rows or columns of fuel channels. This paper will describe the roles of the devices and outline the design requirements of the absorber component from a reactor physics viewpoint. Nuclear heating and activation problems associated with the different absorbers will be briefly discussed. The design and manufacture of the devices will be also discussed. The control rod absorbers and shut off materials are cadmium and stainless steel. In the tubular arrangement, the cadmium is sandwiched between stainless steel tubes. This type of device has functioned well, but there is now concern over the availability and expense of cadmium which is used in two types of CANDU control devices. There are also concerns about the toxicity of cadmium during the fabrication of the absorbers. These concerns are prompting AECL to study alternatives. To minimize design changes, pure boron-10 alloyed in stainless steel is a favoured option. Work is underway to confirm the suitability of the boron-loaded steel and identify other encapsulated absorber materials for practical application. Because the reactivity devices or their guide tubes span the calandria vessel, the long slender components must be sufficiently rigid to resist operational vibration and also be seismically stable. Some of these components are made of Zircaloy to minimize neutron absorption. Slow irradiation growth and creep can reduce the spring tension, and periodic adjustments to the springs are required. Experience with the control absorber devices has generally been good. In one instance liquid zone controllers had a problem of vibration induced fretting but a redesigned back-fit resolved the problem. (author). 3 refs, 8

  14. Leak detection capability in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Azer, N.; Barber, D.H.; Boucher, P.J. [and others

    1997-04-01

    This paper addresses the moisture leak detection capability of Ontario Hydro CANDU reactors which has been demonstrated by performing tests on the reactor. The tests confirmed the response of the annulus gas system (AGS) to the presence of moisture injected to simulate a pressure tube leak and also confirmed the dew point response assumed in leak before break assessments. The tests were performed on Bruce A Unit 4 by injecting known and controlled rates of heavy water vapor. To avoid condensation during test conditions, the amount of moisture which could be injected was small (2-3.5 g/hr). The test response demonstrated that the AGS is capable of detecting and annunciating small leaks. Thus confidence is provided that it would alarm for a growing pressure tube leak where the leak rate is expected to increase to kg/hr rapidly. The measured dew point response was close to that predicted by analysis.

  15. Leak detection capability in CANDU reactors

    International Nuclear Information System (INIS)

    This paper addresses the moisture leak detection capability of Ontario Hydro CANDU reactors which has been demonstrated by performing tests on the reactor. The tests confirmed the response of the annulus gas system (AGS) to the presence of moisture injected to simulate a pressure tube leak and also confirmed the dew point response assumed in leak before break assessments. The tests were performed on Bruce A Unit 4 by injecting known and controlled rates of heavy water vapor. To avoid condensation during test conditions, the amount of moisture which could be injected was small (2-3.5 g/hr). The test response demonstrated that the AGS is capable of detecting and annunciating small leaks. Thus confidence is provided that it would alarm for a growing pressure tube leak where the leak rate is expected to increase to kg/hr rapidly. The measured dew point response was close to that predicted by analysis

  16. Candu technology: the next generation now

    International Nuclear Information System (INIS)

    We describe the development philosophy, direction and concepts that are being utilized by AECL to refine the CANDU reactor to meet the needs of current and future competitive energy markets. The technology development path for CANDU reactors is based on the optimization of the pressure tube concept. Because of the inherent modularity and flexibility of this basis for the core design, it is possible to provide a seamless and continuous evolution of the reactor design and performance. There is no need for a drastic shift in concept, in technology or in fuel. By continual refinement of the flow and materials conditions in the channels, the basic reactor can be thermally and operationally efficient, highly competitive and economic, and highly flexible in application. Thus, the design can build on the successful construction and operating experience of the existing plants, and no step changes in development direction are needed. This approach minimizes investor, operator and development risk but still provides technological, safety and performance advances. In today's world energy markets, major drivers for the technology development are: (a) reduced capital cost; (b) improved operation; (c) enhanced safety; and (d) fuel cycle flexibility. The drivers provide specific numerical targets. Meeting these drivers ensures that the concept meets and exceeds the customer economic, performance, safety and resource use goals and requirements, including the suitable national and international standards. This logical development of the CANDU concept leads naturally to the 'Next Generation' of CANDU reactors. The major features under development include an optimized lattice for SEU (slightly enriched uranium) fuel, light water cooling coupled with heavy water moderation, advanced fuel channels and CANFLEX fuel, optimization of plant performance, enhanced thermal and BOP (balance of plant) efficiency, and the adoption of layout and construction technology adapted from successful on

  17. Influence of aqueous environment pH on the corrosion behaviour of the CANDU steam generator tubing material

    International Nuclear Information System (INIS)

    The generalized corrosion is an undesirable process because it is accompanied by deposition of the corrosion products which affect the steam generator performances. It is very important to understand the generalized corrosion mechanism in order to evaluate the amounts of corrosion products which exist in the steam generator after a determined period of operation. The purpose of the experimental research consists in the assessment of corrosion behavior of the tube material (Incoloy-800) at normal secondary circuit parameters (temperature - 260 deg. C, pressure - 5.1 MPa). The testing environment was the demineralized water without impurities, at different pH values regulated with morpholine and cycloheyilamine (all volatile treatment). The results are presented as micrographs and graphics representing loss of metal by corrosion, corrosion rate, the total corrosion products, the adherent corrosion product, the released corrosion products and the release of the metal. (authors)

  18. The development of a remote gauging and inspection capability for fuel channels in Candu reactors

    International Nuclear Information System (INIS)

    Equipment under development for the inspection and gauging of pressure tubes in CANDU (Canadian Deuterium Uranium) type reactors is described. A brief overview of the mechanical scanning system is presented followed by a detailed description of the measurement and data processing systems for the gauging of diameter and wall thickness, volumetric inspection of the tube wall and gauging of the annular gap between the pressure tube and the calandria tube. Experience of testing ultrasonic transducers in very high (106 Roentgens/hour)(R/h) radiation fields is reviewed. (author)

  19. CANDU development

    International Nuclear Information System (INIS)

    Evolution of the 950 MW(e) CANDU reactor is summarized. The design was specifically aimed at the export market. Factors considered in the design were that 900-1000 MW is the maximum practical size for most countries; many countries have warmer condenser cooling water than Canada; the plant may be located on coastal sites; seismic requirements may be more stringent; and the requirements of international, as well as Canadian, standards must be satisfied. These considerations resulted in a 600-channel reactor capable of accepting condenser cooling water at 320C. To satisfy the requirement for a proven design, the 950 MW CANDU draws upon the basic features of the Bruce and Pickering plants which have demonstrated high capacity factors

  20. OpenFOAM Analysis of CANDU-6 Moderator Flow

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyoung Tae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Chang, Se-Myong [Kunsan National University, Gunsan (Korea, Republic of)

    2015-10-15

    In this study OpenFOAM (Open Field Operation and Manipulation), an open source CFD solver, is used to simulate the three-dimensional moderator flow in calandria tank of CANDU-6 reactor improving the computational efficiency by parallel computing which does not need any proprietary license. A prototype of CANDU-6 reactor is numerically analyzed about three-dimensional moderator flow in calandrian tank with OpenFOAM, an open source CFD code. The horizontal fuel channels in a CANDU-6 reactor (a pressurized heavy water reactor) are submerged in the heavy water (D{sub 2}O) pool which is contained by a cylindrical tank, calandria. Each fuel channel consists of concentric tubes: a Pressure Tube (PT) and a Calandria Tube (CT). And the CO{sub 2} gas is filled between these tubes. Consequently, a heat flux is rapidly transferred to the outer CT so that a film boiling may occur in CT. As a result, it is important to keep the subcooling in the moderator. It is one of the major concerns in the CANDU safety analyses to estimate the local subcooling margin of the moderator inside the calandria tank. Previous experimental studies showed that the film boiling would be unlikely to occur if the local moderator subcooling is sufficient. Therefore, an accurate prediction of the moderator temperature distribution in the calandria tank is needed to confirm the channel integrity. There have been numerous computational efforts to estimate the thermal hydraulics in the calandria tank using CFD codes. Hadaller et al. obtained a tube bank pressure drop model for tube bundle region of the calandria tank and implemented it into the MODTURC{sub C}LAS code. Yoon et al. used the CFX code to develop a CFD model with a porous media approach for the core region. However, it is known that porous media modeling provide only average values of flow velocities and temperatures and do not give any information about local flow variables near tube solid walls, which are necessary to implement accurate heat

  1. Measurements of elastic modulus in Zr alloys for CANDU applications

    International Nuclear Information System (INIS)

    Measurements of elastic modulus as a function of temperature from 20 to 400°C were carried out on specimens of Zr-2.5Nb, Zircaloy-4, Zircaloy-2 and Excel Zr alloy using an ultrasonic resonance technique. The specimens were machined from CANDU pressure tubes, a calandria tube and commercial sheet material. Effects of crystallographic texture, neutron irradiation and hydrogen on elastic modulus were investigated. The results show that elastic modulus of the Zr alloys (1) decreases with increasing temperature, (2) depends strongly on crystallographic texture, and (3) increases slightly with neutron irradiation. (author)

  2. Ultrasonic testing of the fracture toughness of Zr-Nb pressure tubes

    International Nuclear Information System (INIS)

    Bulk elastic properties were measured, using ultrasound, in thickness and circumferential directions of 13 Zr-Nb pressure tubes samples from CANDU nuclear reactors in the hope of finding a nondestructive means to evaluate fracture toughness. The longitudinal wave velocity in the thickness direction are found especially sensitive to changes in the α-Zr single-crystal c axis orientation distribution. This is verified by comparing measured values to predictions based on neutron diffraction measurements of the crystallographic orientation distribution and on the single-crystal elastic constants of α-Zr. Moreover, those velocities most sensitive to texture correlate best with the crack growth toughness of the pressure tubes. This led to the discovery of a correlation between the degree of alignment of the crystallographic c axes along the tube circumferential direction and crack growth toughness. The better is the alignment, the lower is crack growth toughness. Because of the measurement simplicity, the ultrasonic technique could be developed into a rugged industrial sensor

  3. Eccentric pressurized tube for measuring creep rupture

    International Nuclear Information System (INIS)

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

  4. Methodology Improvement of Reactor Physics Codes for CANDU Channels Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Myung Hyun; Choi, Geun Suk; Win, Naing; Aung, Tharndaing; Baek, Min Ho; Lim, Jae Yong [Kyunghee University, Seoul (Korea, Republic of)

    2010-04-15

    As the operational time increase, pressure tubes and calandria tubes in CANDU core encounter inevitably a geometrical deformation along the tube length. A pressure tube may be sagged downward within a calandria tube by creep from irradiation. This event can bring about a problem that is serious in integrity of pressure tube. A measurement of deflection state of in-service pressure tube is, therefore, very important for the safety of CANDU reactor. In this paper, evaluation of impacts on nuclear characteristic due to fuel channel deformation were aimed in order to improve nuclear design tools for concerning the local effects from abnormal deformations. It was known that sagged pressure tube can cause the eccentric configuration of fuel bundles in pressure tube by O.6cm maximum. In this case, adverse pin power distribution and reactivity balance can affect reactor safety under normal and accidental condition. Thermal and radiation-induced creep in pressure tube would expand a tube size. It was known that maximum expansion may be 5% in volume. In this case, more coolant make more moderation in the deformed channel resulting in the increase of reactivity. Sagging of pressure tube did not cause considerable change in K-inf values. However, expansion of the pressure tube made relatively large change in K-inf. Modeling of eccentric and enlarged configuration is not easy in preparation of input geometry at both HELlOS and MCNP. On the other hand, there is no way to consider this deformation in one-dimensional homogenization tool such as WIMS code. The way of handling this deformation was suggested as the correction method of expansion effect by adjusting the number density of coolant. The number density of heavy water coolant was set to be increased as the rate of expansion increase. This correction was done in the intact channel without changing geometry. It was found that this correction was very effective in the prediction of K-inf values. In this study, further

  5. The pressure tube inspection and integrity evaluation in Fugen

    International Nuclear Information System (INIS)

    Two hundred and twenty four pressure tubes are installed vertically in the reactor of the Fugen. Each pressure tube accommodates a fuel assembly and forms the pressure boundary of the primary cooling system. The operating pressure is 6.8 MPa and temperature 280degC. The pressure tube, made of Heat Treated Zr-2.5%Nb, is approximately five meter long, with 117.8 mm inner diameter and 4.3 mm wall thickness. The pressure tube is connected to upper and lower extension tubes of stainless steel with the rolled-joint technique. The soundness of the pressure tube, rolled joints, and upper/lower extension tubes must be checked in an appropriate and systematic manner. To satisfy the requirements, in-service inspections (ISIs) and post irradiation examinations (PIEs) of the pressure tubes have been carried out during the 24 year operation of the Fugen. Development of pressure tube inspection equipment started in 1971 for the ISIs. The first model of the equipment was developed and applied to the pre-service inspection of the pressure tubes in 1977. The measurement accuracy of the equipment was sufficient but the weight and size were too large to be set and handled in an irradiated environment. Thus, the design was modified to smaller the equipment in size, lighten to approx. 1/100 in weight, and realize to be handled with the refueling machine. The improved equipment was used in the 4th annual inspection in March 1984. Ultrasonic flaw detections, inner diameter measurements and inner surface visual inspection of pressure tubes were conducted. Up to the 17th annual inspection in 2002, 146 inspections in total were executed. The ultrasonic inspection detected no defect on the pressure tubes. The measured strain due to the irradiation creep of pressure tubes corresponded with the design values. To conduct the PIEs of the pressure tube materials, surveillance specimens were set in the special fuel assemblies and irradiated from the beginning of the reactor operation. Five PIEs

  6. Pickering NGS A: Assessment of calandria tube integrity following a sudden pressure tube failure

    International Nuclear Information System (INIS)

    The issue of calandria tube integrity following a sudden rupture of the pressure tube in Pickering NGS A reactor is addressed. Based on operating experience, only fish-mouth ruptures of the pressure tube are considered to be credible. The calandria tube response to the pressure tube break is delineated into three distinct stages, i.e. the initial transient response during the annulus filling stage, transient overpressurization and the final steady-state loading after bellows failure. The annulus response in the second stage is dominated by a waterhammer type overpressure transient with attenuation of this transient due to plastic straining of the calandria tube. The annulus pressure transients for various breaks and the sensitivity of the results to various parameters are presented. The strength margins of the calandria tube are evaluated to be relatively large. (author). 7 refs., 6 tabs., 6 figs

  7. CATHENA Code Assessment for Pressure Tube and Calandria Tube Contact Phenomena

    International Nuclear Information System (INIS)

    Canadian Algorithm for THErmalhydraulic Network Analysis (CATHENA), has been validated against full-scale Contact Boiling Experiments conducted using specific channel power, pressure, and moderator subcooling as pre-test conditions. The pressure tube (PT) and calandria tube (CT) temperatures, the extent of dryout and failures of the pressure tube or the calandria tube (if any) are the outcome of these experiments. Recently, an IAEA International Collaborative Standard Problem (ICSP) to provide contact boiling experimental data to participants for assessing the subcooling requirements for a heated pressure tube, plastically deforming into contact with the calandria tube during a postulated large break LOCA condition has been performed. The CATHENA code assessment results against the experimental data distributed for the ICSP are provided in this paper. The CATHENA code is used to simulate the experiment on pressure tube ballooning conducted at the AECL. The overall code's predictions show good agreements with the experimental data. The contact timing by the pressure tube ballooning is predicted accurately, however, it is found that the code largely underpredict the peak temperature at the pressure tube and the calandria tube. This discrepancy seems to be induced from multi-dimensional flow effects in the water tank. For more accurate calculations, detailed modeling of the water tank is required

  8. Life Assurance Strategy for CANDU NPP

    International Nuclear Information System (INIS)

    include design provisions to replace fuel channels and steam generators. Difficult to replace components such as reactor building structures and calandria/shield tank assembly are designed for much beyond 40 years. Given the performance of CND's to date and the successfully completed rehabilitations and the lessons learned from older plants, a newly committed CANDU will have an economic service life significantly longer than 40 years. The CANDU design life was initially set at thirty years. The key components of a CANDU nuclear steam plant are the calandria vessel, the fuel channels, the reactivity control mechanisms, and the primary heat transport components including piping and steam generators. The calandria vessel, a large stainless steel tank, experiences conditions of relatively low temperature and pressure and is designed for a very long life. Experience to date shows that of the remaining components, fuel channels and reactivity control mechanisms are replaceable. Given that other refurbishments and/or replacements can be done to existing plants, a minimum of 40 year operating life can be achieved. Large scale fuel channel replacement was dictated by Station Life Assurance rather than Life Extension considerations. This major rehabilitation program has been successfully implemented for three of the Pickering A reactors to achieve a minimum 40 year operating life. In this program steady flow of successful design and process improvements have contributed to the knowledge base and know how of the CANDU industry. Over the next few years, retuning of the fourth Pickering A unit and the first of the Bruce A units will be undertaken providing the opportunity for Life extension of these units. Steam Generators in most CANDU plants continue to perform, with relatively low tube failures and plugging rates. Remedial measures are being taken, with solutions being evaluated by Ontario Hydro to address current degradation problems due to tube fouling and sludge deposition. R

  9. CANDU refurbishment - managing the life cycle

    International Nuclear Information System (INIS)

    All utilities that operate a nuclear power plant have an integrated plan for managing the condition of the plant systems, structures and components. With a sound plant life management program, after about 25 years of operation, replacement of certain reactor core components can give an additional 25 to 30 years of operation. This demonstrates the long-term economic strength of CANDU technology and justifies a long-term commitment to nuclear power. Indeed, replacement of pressure tubes and feeders with the most recent technology will also lead to increased capacity factors - due to reduced requirements for feeder inspections and repair, and eliminating the need for fuel channel spacer relocation which have caused additional and longer maintenance outages. Continuing the operation of CANDU units parallels the successful life extensions of reactors in other countries and provides the benefits of ongoing reliable operation, at an existing plant location, with the continued support of the host community. The key factors for successful, optimum management of the life cycle are: ongoing, effective plant life management programs; careful development of refurbishment scope, taking into account system condition assessments and a systematic safety review; and, a well-planned and well-executed retubing and refurbishment outage, where safety and risk management is paramount to ensure a successful project The paper will describe: the benefits of extended plant life; the outlook for refurbishment; the life management and refurbishment program; preparations for retubing of the reactor core; and, enhanced performance post-retubing. Given the potential magnitude of the program over the next 10 years, AECL will maintain a lead role providing overall support for retubing and plant Life Cycle Management programs and the CANDU Owners Group will provide a framework for collaboration among its Members. (author)

  10. Research on method of pressure grouting piling of driven tube

    Institute of Scientific and Technical Information of China (English)

    Dianqi PAN; Zupei ZHANG; Diancai PAN; Yong CHEN; Maosen TAN

    2006-01-01

    The pressure grouting pile of driven tube can improve the load bearing capacity of the single pile from the mechanism of pressure grouting pile of driven tube. On the basis of analyzing the mechanism, the authors designed the machines and tools of pressure grouting, determined the operating manufacture and technology parameter on the pressure grouting secondly. The result shows that the pressure grouting pile of driven tube not only changes the pile type but also reduce the length of the pile and its engineering cost, it enhances the load bearing capacity of single pile an the same time.

  11. Requirements for class 1C, 2C, and 3C pressure-retaining components and supports in CANDU nuclear power plants

    International Nuclear Information System (INIS)

    This Standard applies to pressure-retaining components of CANDU nuclear power plants that have a code classification of Class 1C, 2C or 3C. These are pressure-retaining components where, because of the design concept, the rules of the ASME Boiler and Pressure Vessel Code do not exist, are not applicable, or are not sufficient. The Standard provides rules for the design, fabrication, installation, examination and inspection of these components and supports. It provides rules intended to ensure the pressure-retaining integrity of components, not the operability. It also provides rules for the support of fueling machines. The Standard applies only to new construction prior to the plant being declared in service

  12. Resistance welding of tubes at low regidual pressure jn tube cavity

    International Nuclear Information System (INIS)

    The procedure of butt resistance welding of boilers in diameter of 32 mm at low residual pressure in tube cavities has been studied. It is shown that the creation of low residual pressure in tube cavity makes it possible to produce qualitative joints of tubes of the 20, 12Kh1MF, 12Kh18N12T steels. The maximum relative deformation in the butt zone should be in the range of 0.5...0.6

  13. CANDU nuclear power system

    International Nuclear Information System (INIS)

    This report provides a summary of the components that make up a CANDU reactor. Major emphasis is placed on the CANDU 600 MW(e) design. The reasons for CANDU's performance and the inherent safety of the system are also discussed

  14. X-ray diffraction residual stress measurement in the rolled-joint zone of Zr - 2.5 % Nb pressure tube

    International Nuclear Information System (INIS)

    The in-service experience of Zr - 2.5 % Nb pressure tubes in CANDU-type nuclear reactors has demonstrated very good performance over a long period of time. However, analyses done by AECL specialists on most failure cases, showed that a big percentage of defects are manufacturing defects, which appear mostly at the beginning of the rolled-joint zone. It has been observed that a correct rolling ensures an acceptable distribution of residual stress, but an incorrect one leads to an accumulation of big values of residual stress. This determines a preferential radial orientation of hydrides, which during operation in the reactor can produce DHC. To ensure a suitable performance of the Zr - 2.5 % Nb pressure tubes in the CANDU reactor, it is very important to have a correct rolling as mentioned in the procedure. This work presents a methodology for the measurement of the stressing state in the surfaces layers of the rolled-joint zone. The X-ray diffraction method can also be used for establishing the residual stress distribution across the tub wall, in order to ensure a good performance at Cernavoda nuclear plant. The results obtained for the investigated tube have led to the conclusion that the rolling process was correctly applied in this case, the values obtained for the residual stress being in good agreement with those accepted in literature. (Author) 2 Figs., 2 Tabs

  15. ASSERT/NUCIRC commissioning for CANDU 6 fuel channel CCP analysis

    International Nuclear Information System (INIS)

    CANDU PHWR fuel channel pressure tubes will expand or creep under long-term (aging process) influence of temperature, pressure, and neutron flux. This diametral pressure tube creep will influence the critical channel power (CCP), or conditions that lead to dryout. In order to provide safety analysis models to quantify the effect of diametral pressure tube creep on CCP, a COG (AECL/NBP/HQ) project is underway to commission the ASSERT and NUCIRC codes to establish reliable production tools for the assessment of CANDU6 CCP in nominal (uncrept) and crept pressure tube fuel channels. This paper gives an overview of the background and objectives of the project along with a brief introduction into the subchannel analysis code ASSERT and the 1-D thermalhydraulics code NUCIRC. This project is a multistage endeavour, for which the first stage results are presented. A detailed cross-comparison of the 1-D (NUCIRC) and subchannel (ASSERT) models of pressure drop (ΔP) and critical heat flux (CHF) has been undertaken and has led to several enhancements and refinements to the respective models. These results are presented in addition to results of ASSERT commissioning against NUCIRC for a matrix of ΔP and dryout cases in a nominal pressure tube, which are based upon Gentilly 2 and Point Lepreau site area. Additionally, the initial results of an assessment, using ASSERT, of the effects of creep on ΔP are presented. In concluding, the status and future directions for ASSERT/NUCIRC CANDU 6 CCP analysis project are summarized. (author). 2 refs., 12 figs

  16. Deadly pressure pneumothorax after withdrawal of misplaced feeding tube

    DEFF Research Database (Denmark)

    Andresen, Erik Nygaard; Frydland, Martin; Usinger, Lotte

    2016-01-01

    BACKGROUND: Many patients have a nasogastric feeding tube inserted during admission; however, misplacement is not uncommon. In this case report we present, to the best of our knowledge, the first documented fatality from pressure pneumothorax following nasogastric tube withdrawal. CASE PRESENTATION......: An 84-year-old Caucasian woman with dysphagia and at risk of aspiration underwent routine insertion of a nasogastric feeding tube; however, shortly after insertion she developed respiratory distress. A chest X-ray showed the tube had been misplaced into our patient's right lung. The tube was removed......, but our patient died less than an hour after withdrawal. The autopsy report stated that cause of death was tension pneumothorax, which developed following withdrawal of the misplaced feeding tube. CONCLUSIONS: The indications for insertion of nasogastric feeding tubes are many and the procedure...

  17. Boussignac continuous positive airway pressure for weaning with tracheostomy tubes

    NARCIS (Netherlands)

    Dieperink, Willem; Aarts, Leon P. H. J.; Rodgers, Michael G. G.; Delwig, Hans; Nijsten, Maarten W. N.

    2008-01-01

    Background: In patients who are weaned with a tracheostomy tube ( TT), continuous positive airway pressure ( CPAP) is frequently used. Dedicated CPAP systems or ventilators with bulky tubing are usually applied. However, CPAP can also be effective without a ventilator by the disposable Bous-signac C

  18. Comparison of the intracuff pressures of three different tracheostomy tubes.

    Science.gov (United States)

    Nishiyama, Tomoki

    2005-01-01

    The purpose of this study was to compare the cuff pressures of three tracheostomy tubes, MERA sofit CLEAR, Blue Line Tracheostomy Tube, and Tracheosoft. Each tracheostomy tube with an internal diameter of 7.0 mm was put into a plastic column. The cuff was then inflated with air to seal the column, and the column was filled with water. The air in the cuff was withdrawn gradually and the cuff pressure at the point of water leakage was measured. Six columns of different size were used. In columns with an internal diameter of 18-21 mm, the water leakage pressure was lower in the following order: MERA sofit CLEAR sofit CLEAR was found to maintain most safely the lowest intracuff pressure to seal the trachea among the three tracheostomy tubes tested. PMID:16032458

  19. Measurement of internal diameter of pressure tubes in pressurized heavy water reactors using ultrasonics

    International Nuclear Information System (INIS)

    The Pressure Tube in Pressurized Heavy Water Reactors (PHWRs) undergoes dimensional changes due to the effects of creep and growth as it is subjected to high pressure and temperature, which causes Pressure Tubes to permanently increase in length and diameter and to sag because of weight of fuel and coolant (heavy water) contained in it. These dimensional changes are due to prolonged stresses under high temperature and radiation. Pressure Tube stresses are evaluated for both beginning and end of life for accounting the Pressure Tube dimensional changes that occur during its design life. At the beginning of life, the initial wall thickness and un-irradiated material properties are applied. At the end of life, Pressure Tube diameter and length increases, while wall thickness decreases. Material strength also increases during that period. The increase in Pressure Tube diameter results in squeezing of garter spring spacer between the pressure and calandria Tubes. It also causes unacceptable heat removal from the fuel due to an increased amount of primary coolant that bypasses the fuel bundles. This reduces the critical channel power at constant flow. Hence the periodic monitoring of pressure Tube diameter is important for these reasons. This is also required as per the applicable codes and standards for In-Service Inspection of PHWRs. Mechanical measurement from ID of the Tube during periodic monitoring is not practically feasible due to high radiation and inaccessibility. This necessitates the development of NDT technique using Ultrasonics for periodic in-situ measurement of ID of pressure Tubes with a BARC made remotely operated drive system called BARCIS (BARC Channel Inspection system). The development of Ultrasonic based ID measurement techniques and their actual applications in PHWRs Pressure tubes are being discussed in this paper. (author)

  20. Modelling and simulation of dynamic characteristics of CANDU-SCWR

    International Nuclear Information System (INIS)

    Owing to the thermal properties of supercritical water and features of heat transfer correlation under supercritical pressure, a detailed thermal-hydraulic model with movable boundary of is developed for CANDU-SCWR (Supercritical Water-Cooled Reactor). Steady-state results of the model agree well with the design data. The dynamic responses of CANDU-SCWR to different disturbances are simulated and characteristics are analyzed. A dynamic model for ACR is also developed using CATHENA. Differences between dynamic characteristics of CANDU-SCWR and those of ACR are highlighted and investigated. It is concluded that CANDU-SCWR has a larger time constant, but with a higher response amplitude. (author)

  1. Fracture analysis of axially cracked pressure tube of pressurized heavy water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Krishnan, S.; Bhasin, V.; Mahajan, S.C. [Bhabha Atomic Research Centre, Bombay (India)] [and others

    1997-04-01

    Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300{degrees}C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered.

  2. Fracture analysis of axially cracked pressure tube of pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Three Dimensional (313) finite element elastic plastic fracture analysis was done for through wall axially cracked thin pressure tubes of 220 MWe Indian Pressurized Heavy Water Reactor. The analysis was done for Zr-2 and Zr-2.5Nb pressure tubes operating at 300 degrees C and subjected to 9.5 Mpa internal pressure. Critical crack length was determined based on tearing instability concept. The analysis included the effect of crack face pressure due to the leaking fluid from tube. This effect was found to be significant for pressure tubes. The available formulae for calculating J (for axially cracked tubes) do not take into account the effect of crack face pressure. 3D finite element analysis also gives insight into variation of J across the thickness of pressure tube. It was observed that J is highest at the mid-surface of tube. The results have been presented in the form of across the thickness average J value and a peak factor on J. Peak factor on J is ratio of J at mid surface to average J value. Crack opening area for different cracked lengths was calculated from finite element results. The fracture assessment of pressure tubes was also done using Central Electricity Generating Board R-6 method. Ductile tearing was considered

  3. TEM examination of irradiated zircaloy-2 pressure tube material

    International Nuclear Information System (INIS)

    In the present work, microstructure of the zircaloy-2 pressure tube material irradiated in the Indian Pressurized Heavy Water RAPP-1. Reactor (PHWR) has been examined for the first time using transmission electron microscope (TEM). The samples were obtained from a zircaloy-2 pressure tube, which had been in operation in the high flux region of Rajasthan Atomic Power Station Unit -1, for a period for 6.77 effective full power years (EFPYs) and expected to have a cumulative radiation damage of about 3 dpa. In this study irradiated microstructure has been characterized and compared it with the microstructure of the unirradiated pressure tube samples. The effect of irradiation on the hydriding behaviour is also studied. (author)

  4. Long-term passive CANDU containment response after a design-basis accident

    International Nuclear Information System (INIS)

    A passive CANDU reg-sign containment system, currently being developed, is aimed at limiting the consequences of a postulated accident, by ensuring the structural integrity of the containment building and limiting fission-product release to within siting dose limits, without operator action or reliance on ac power for up to 3 d. All main functions of the containment system, i.e. energy removal, hydrogen mitigation, and fission-product retention, are to be accomplished passively. The passive CANDU containment relies on the passive emergency water system (PEWS) for energy removal after an accident and on passive autocatalytic recombiners (PAR) for hydrogen removal. The key feature of this concept, is a recirculating, buoyancy-driven flow through the recombiners and the tube banks of the PEWS. This paper presents preliminary design calculations for the PEWS tank and tube banks and a simulation of the long-term passive containment response, based on the current CANDU-6 containment, to a large loss-of-coolant/loss-of- emergency coolant injection (LOCA/LOECI) using the GOTHIC code. It is shown that a 1500-M3 PEWS tank, connected to tube banks with a total surface area of 1800 m2, can limit the second pressure peak to about 300 kPa(a) if a recirculating flow is established in the containment building. The PEWS tank water is boiling in the long term, and the peak containment temperature is 114 degrees C. 6 refs., 4 figs

  5. Deadly pressure pneumothorax after withdrawal of misplaced feeding tube

    DEFF Research Database (Denmark)

    Andresen, Erik Nygaard; Frydland, Martin; Usinger, Lotte

    2016-01-01

    BACKGROUND: Many patients have a nasogastric feeding tube inserted during admission; however, misplacement is not uncommon. In this case report we present, to the best of our knowledge, the first documented fatality from pressure pneumothorax following nasogastric tube withdrawal. CASE PRESENTATION......, but our patient died less than an hour after withdrawal. The autopsy report stated that cause of death was tension pneumothorax, which developed following withdrawal of the misplaced feeding tube. CONCLUSIONS: The indications for insertion of nasogastric feeding tubes are many and the procedure...... is considered harmless; however, if the tube is misplaced there is good reason to be cautious on removal as this can unmask puncture of the pleura eliciting pneumothorax and, as this case report shows, result in an ultimately deadly tension pneumothorax....

  6. Creep-rupture tests of internally pressurized Inconel 702 tubes

    Science.gov (United States)

    Gumto, K. H.

    1973-01-01

    Seamless Inconel 702 tubes with 0.375-in. outside diameter and 0.025-in. wall thickness were tested to failure at temperatures from 1390 to 1575 F and internal helium pressures from 700 to 1800 psi. Lifetimes ranged from 29 to 1561 hr. The creep-rupture strength of the tubes was about 70 percent lower than that of sheet specimens. Larson-Miller correlations and photomicrographs of some specimens are presented.

  7. Desktop Severe Accident Graphic Simulator Module for CANDU6 : PSAIS

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. Y.; Song, Y. M. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The ISAAC ((Integrated Severe Accident Analysis Code for CANDU Plant) code is a system level computer code capable of performing integral analyses of potential severe accident progressions in nuclear power plants, whose main purpose is to support a Level 2 probabilistic safety assessment or severe accident management strategy developments. The code has the capability to predict a severe accident progression by modeling the CANDU6- specific systems and the expected physical phenomena based on the current understanding of the unique accident progressions. The code models the sequence of accident progressions from a core heatup, pressure tube/calandria tube rupture after an uncovery from inside and outside, a relocation of the damaged fuel to the bottom of the calandria, debris behavior in the calandria, corium quenching after a debris relocation from the calandria to the calandria vault and an erosion of the calandria vault concrete floor, a hydrogen burn, and a reactor building failure. Along with the thermal hydraulics, the fission product behavior is also considered in the primary system as well as in the reactor building.

  8. Temperature effect of dynamic anisotropic elastic constants of Zr-2.5Nb pressure tube by resonant ultrasound spectroscopy

    International Nuclear Information System (INIS)

    Dynamic anisotropic elastic constants of CANDU Zr-2.5Nb pressure tube materials were determined by high temperature resonant ultrasound spectroscopy (RUS). The resonance frequencies were measured using a couple of alumina waveguides and wide-band ultrasonic transducers in a small furnace. The rectangular parallelepiped specimens were fabricated along with the longitudinal, radial and transverse direction of the pressure tube. The initial estimates for RUS were obtained from the orientation distribution function by X-ray pole figure and elastic stiffness of single crystal zirconium. A nine elastic stiffness tensor for orthotropic symmetry was determined in the range of room temperature ∼500 deg. C. As the temperature increases, the elastic constant tensor, cij gradually decreases. Higher elastic constants along the transverse direction compared to those along the longitudinal or radial direction are similar to the case of Young's modulus or shear modulus. A crossing of elastic constants along the longitudinal direction and radial direction was observed near 120-150 deg. C. This fact could correlate to the crossing characteristics of c44 and c66 of a zirconium single crystal in the temperature range

  9. Pressure Loss across Tube Bundles in Two-phase Flow

    Energy Technology Data Exchange (ETDEWEB)

    Sim, Woo Gun; Banzragch, Dagdan [Hannam Univ., Daejon (Korea, Republic of)

    2016-03-15

    An analytical model was developed by Sim to estimate the two-phase damping ratio for upward two-phase flow perpendicular to horizontal tube bundles. The parameters of two-phase flow, such as void fraction and pressure loss evaluated in the model, were calculated based on existing experimental formulations. However, it is necessary to implement a few improvements in the formulations for the case of tube bundles. For the purpose of the improved formulation, we need more information about the two-phase parameters, which can be found through experimental test. An experiment is performed with a typical normal square array of cylinders subjected to the two-phase flow of air-water in the tube bundles, to calculate the two-phase Euler number and the two-phase friction multiplier. The pitch-to-diameter ratio is 1.35 and the diameter of cylinder is 18mm. Pressure loss along the flow direction in the tube bundles is measured with a pressure transducer and data acquisition system to calculate the two-phase Euler number and the two-phase friction multiplier. The void fraction model by Feenstra et al. is used to estimate the void fraction of the two-phase flow in tube bundles. The experimental results of the two phase friction multiplier and two-phase Euler number for homogeneous and non-homogeneous two-phase flows are compared and evaluated against the analytical results given by Sim's model.

  10. CANDU 9 design

    International Nuclear Information System (INIS)

    AECL has made significant design improvements in the latest CANDU nuclear power plant (NPP) - the CANDU 9. The CANDU 9 operates with the energy efficient heavy water moderated reactor and natural uranium fuel and utilizes proven technology. The CANDU 9 NPP design is similar to the world leading CANDU 6 but is based upon the single unit adaptation of the 900 MWe class reactors currently operating in Canada as in integrated four-unit configurations. The evolution of the CANDU family of heavy water reactors (HAIR) is based on a continuous product improvement approach. Proven equipment and systems from operating stations are standardized and used in new products. As a result of the flexibility of the technology, evolution of the current design will ensure that any new requirements can be met, and there is no need to change the basic concept. This paper will provide an overview for some of the key features of the CANDU 9 NPP such as nuclear systems and equipment, advanced control and computer systems, safety design and protection features, and plant layout. The safety enhancements and operability improvements implemented in this design are described and some of the advantages that can be expected by the operating utility are highlighted. (author)

  11. CATASTROPHE FRACTURE OF THIN-WALL PRESSURE TUBES

    Institute of Scientific and Technical Information of China (English)

    魏德敏; 杨桂通

    2002-01-01

    Catastrophe theory was used to investigate the fracture behavior of thin-wall cylindrical tubes subjected to nternal explosive pressure. Based on the energy theory and catastrophe theory, a cusp catastrophe model for the fracture was established, and a critical condition associated with the model is given.

  12. Assessment of ASSERT-PV for prediction of post-dryout heat transfer in CANDU bundles

    International Nuclear Information System (INIS)

    Highlights: • Assessment of the new Canadian subchannel code ASSERT-PV 3.2 for PDO sheath temperature prediction. • CANDU 28-, 37- and 43-element bundle PDO experiments. • Prediction improvement of ASSERT-PV 3.2 over previous code versions. • Sensitivity study of the effect of PDO model options. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of subchannel flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against PDO tests performed during five full-size CANDU bundle experiments conducted between 1992 and 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element bundles. A total of 10 PDO test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for existing CANDU reactors. Code predictions of maximum PDO fuel-sheath temperature were compared against measurements from the SL PDO tests to quantify the code's prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, separate-effects sensitivity studies quantified the contribution of each PDO model change or enhancement to the improvement in PDO heat transfer prediction. Overall, the assessment demonstrated significant improvement in prediction of PDO sheath temperature in horizontal fuel channels containing CANDU bundles

  13. Assessment of ASSERT-PV for prediction of critical heat flux in CANDU bundles

    International Nuclear Information System (INIS)

    Highlights: • Assessment of the new Canadian subchannel code ASSERT-PV 3.2 for CHF prediction. • CANDU 28-, 37- and 43-element bundle CHF experiments. • Prediction improvement of ASSERT-PV 3.2 over previous code versions. • Sensitivity study of the effect of CHF model options. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against five full-scale CANDU bundle experiments conducted in 1990s and in 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element (CANFLEX) bundles. A total of 15 CHF test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for CANDU reactors. Code predictions of channel dryout power and axial and radial CHF locations were compared against measurements from the SL CHF tests to quantify the code prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, the sensitivity studies evaluated the contribution of each CHF model change or enhancement to the improvement in CHF prediction. Overall, the assessment demonstrated significant improvement in prediction of channel dryout power and axial and radial CHF locations in horizontal fuel channels containing CANDU bundles

  14. Assessment of ASSERT-PV for prediction of critical heat flux in CANDU bundles

    Energy Technology Data Exchange (ETDEWEB)

    Rao, Y.F., E-mail: raoy@aecl.ca; Cheng, Z., E-mail: chengz@aecl.ca; Waddington, G.M., E-mail: waddingg@aecl.ca

    2014-09-15

    Highlights: • Assessment of the new Canadian subchannel code ASSERT-PV 3.2 for CHF prediction. • CANDU 28-, 37- and 43-element bundle CHF experiments. • Prediction improvement of ASSERT-PV 3.2 over previous code versions. • Sensitivity study of the effect of CHF model options. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against five full-scale CANDU bundle experiments conducted in 1990s and in 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element (CANFLEX) bundles. A total of 15 CHF test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for CANDU reactors. Code predictions of channel dryout power and axial and radial CHF locations were compared against measurements from the SL CHF tests to quantify the code prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, the sensitivity studies evaluated the contribution of each CHF model change or enhancement to the improvement in CHF prediction. Overall, the assessment demonstrated significant improvement in prediction of channel dryout power and axial and radial CHF locations in horizontal fuel channels containing CANDU bundles.

  15. Assessment of ASSERT-PV for prediction of post-dryout heat transfer in CANDU bundles

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, Z., E-mail: chengz@aecl.ca; Rao, Y.F., E-mail: raoy@aecl.ca; Waddington, G.M., E-mail: waddingg@aecl.ca

    2014-10-15

    Highlights: • Assessment of the new Canadian subchannel code ASSERT-PV 3.2 for PDO sheath temperature prediction. • CANDU 28-, 37- and 43-element bundle PDO experiments. • Prediction improvement of ASSERT-PV 3.2 over previous code versions. • Sensitivity study of the effect of PDO model options. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of subchannel flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against PDO tests performed during five full-size CANDU bundle experiments conducted between 1992 and 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element bundles. A total of 10 PDO test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for existing CANDU reactors. Code predictions of maximum PDO fuel-sheath temperature were compared against measurements from the SL PDO tests to quantify the code's prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, separate-effects sensitivity studies quantified the contribution of each PDO model change or enhancement to the improvement in PDO heat transfer prediction. Overall, the assessment demonstrated significant improvement in prediction of PDO sheath temperature in horizontal fuel channels containing CANDU bundles.

  16. Plant life management strategies for pressurized heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Suk; Kwon, Sang Chul; Choo, Ki Nam; Ahn, Sang Bok; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-06-01

    This technical report reviewed aging mechanism of the major components of CANDU 6 reactor such as pressure tubes, calandria tube, end fitting, fuel channel spacer and calandria. Furthermore, the surveillance methodology was described for monitoring and inspection of these core components. Based on the in-reactor performances data such as delayed hydride cracking, leak-before-break, enhanced deformation-creep and growth, the life management of pressure tubes was illustrated in this report. (author). 19 refs., 11 figs., 2 tabs.

  17. Marketing CANDU internationally

    International Nuclear Information System (INIS)

    The market for CANDU reactor sales, both international and domestic, is reviewed. It is reasonable to expect that between five and ten reactors can be sold outside Canada before the end of the centry, and new domestic orders should be forthcoming as well. AECL International has been created to market CANDU, and is working together with the Canadian nuclear industry to promote the reactor and to assemble an attractive package that can be offered abroad. (L.L.)

  18. Periodic inspection of CANDU nuclear power plant containment components

    International Nuclear Information System (INIS)

    This Standard is one in a series intended to provide uniform requirements for CANDU nuclear power plants. It provides requirements for the periodic inspection of containment components including the containment pressure suppression systems

  19. Microhole High-Pressure Jet Drill for Coiled Tubing

    Energy Technology Data Exchange (ETDEWEB)

    Ken Theimer; Jack Kolle

    2007-06-30

    Tempress Small Mechanically-Assisted High-Pressure Waterjet Drilling Tool project centered on the development of a downhole intensifier (DHI) to boost the hydraulic pressure available from conventional coiled tubing to the level required for high-pressure jet erosion of rock. We reviewed two techniques for implementing this technology (1) pure high-pressure jet drilling and (2) mechanically-assisted jet drilling. Due to the difficulties associated with modifying a downhole motor for mechanically-assisted jet drilling, it was determined that the pure high-pressure jet drilling tool was the best candidate for development and commercialization. It was also determined that this tool needs to run on commingled nitrogen and water to provide adequate downhole differential pressure and to facilitate controlled pressure drilling and descaling applications in low pressure wells. The resulting Microhole jet drilling bottomhole assembly (BHA) drills a 3.625-inch diameter hole with 2-inch coil tubing. The BHA consists of a self-rotating multi-nozzle drilling head, a high-pressure rotary seal/bearing section, an intensifier and a gas separator. Commingled nitrogen and water are separated into two streams in the gas separator. The water stream is pressurized to 3 times the inlet pressure by the downhole intensifier and discharged through nozzles in the drilling head. The energy in the gas-rich stream is used to power the intensifier. Gas-rich exhaust from the intensifier is conducted to the nozzle head where it is used to shroud the jets, increasing their effective range. The prototype BHA was tested at operational pressures and flows in a test chamber and on the end of conventional coiled tubing in a test well. During instrumented runs at downhole conditions, the BHA developed downhole differential pressures of 74 MPa (11,000 psi, median) and 90 MPa (13,000 psi, peaks). The median output differential pressure was nearly 3 times the input differential pressure available from the

  20. The CANDU contribution to environmentally friendly energy production

    International Nuclear Information System (INIS)

    National prosperity is based on the availability of affordable, energy supply. However, this need is tempered by a complementary desire that the energy production and utilization will not have a major impact on the environment. The CANDU energy system, including a next generation of CANDU designs, is a major primary energy supply option that can be an important part of an energy mix to meet Canadian needs. CANDU nuclear power plants produce energy in the form of medium pressure steam. The advanced version of the CANDU design can be delivered in unit modules ranging from 400 to 1200 MWe. This Next Generation of CANDU designs features lower cost, coupled with robust safety margins. Normally this steam is used to drive a turbine and produce electricity. However, a fraction of this steam (large or small) may alternatively be used as process steam for industrial consumption. Options for such steam utilization include seawater desalination, oil sands extraction and heating. The electricity may be delivered to an electrical grid or alternatively used to produce quantities of hydrogen. Hydrogen is an ideal clean transportation fuel because its use only produces water. Thus, a combination of CANDU generated electricity and hydrogen distribution for vehicles is an available, cost-effective route to dramatically reduce emissions from the transportation sector. The CANDU energy system contributes to environmental protection and the prevention of climate change because of its very low emission. The CANDU energy system does not produce any NOx, SOx or greenhouse gas (notably CO2) emissions during operation. In addition, the CANDU system operates on a fully closed cycle with all wastes and emissions fully monitored, controlled and managed throughout the entire life cycle of the plant. The CANDU energy system is an environmentally friendly and flexible energy source. It can be an effective component of a total energy supply package, consistent with Canadian and global climate

  1. Enhancing the moderator effectiveness as a heat sink during loss-of-coolant accidents in CANDU-PHW reactors using glass-peened surfaces

    Energy Technology Data Exchange (ETDEWEB)

    Nitheanandan, T.; Tiede, R.W.; Sanderson, D.B. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada); Fong, R.W.L.; Coleman, C.E

    1998-08-01

    The horizontal fuel channel concept is a distinguishing feature of the CANDU-PHW reactor. Each fuel channel consists of a Zr-2.5Nb pressure tube and a Zircaloy-2 calandria tube, separated by a gas filled annulus. The calandria tube is surrounded by heavy-water moderator that also provides a backup heat sink for the reactor core. This heat sink (about 10 mm away from the hot pressure tube) ensures adequate cooling of fuel in the unlikely event of a loss-of-coolant accident (LOCA). One of the ways of enhancing the use of the moderator as a heat sink is to improve the heat-transfer characteristics between the calandria tube and the moderator. This enhancement can be achieved through surface modifications to the calandria tube which have been shown to increase the tube's critical heat flux (CHF) value. An increase in CHIF could be used to reduce moderator subcooling requirements for CANDU fuel channels or increase the margin to dryout. A series of experiments was conducted to assess the benefits provided by glass-peening the outside surface of calandria tubes for postulated LOCA conditions. In particular, the ability to increase the tube's CHF, and thereby reduce moderator subcooling requirements was assessed. Results from the experiments confirm that glass-peening the outer surface of a tube increases its CHF value in pool boiling. This increase in CHF could be used to reduce moderator subcooling requirements for CANDU fuel channels by at least 5 degrees C. (author)

  2. Detection of blister formation and evaluation of pressure tube/calandria tube contact location by ultrasonic velocity ratio measurement technique

    International Nuclear Information System (INIS)

    Presence of hydrogen in zircaloy pressure tube affects the velocity of ultrasound propagation. Both longitudinal wave velocity (VL) and shear wave velocity (VS) are affected depending on the concentration of hydrogen. Velocity ratio (VL/VS) changes as per the concentrations of hydrogen in different locations along the length of pressure tube. A hydride blister which forms at the pressure tube and calandria tube contact point is a distinct zone containing hydrogen 2-3 order of magnitude more than the parent matrix and hence, can be detected by sharp change in velocity ratio. (author)

  3. Development of pressure tube inspection equipment for the Fugen (ATR)

    International Nuclear Information System (INIS)

    A remote-controlled in-service inspection device has been developed for inspecting the pressure tubes of the Fugen, which is a heavy-water-moderated, boiling light-water-cooled pressure-tube-type reactor. The equipment is capable of performing three kinds of inspection: ultrasonic flaw detection, measurement of inside diameter and visual inspection of the internal surface. To reduce the radiation exposure of inspectors, the three kinds of detectors, with their associated electronics and drive mechanisms for vertical and rotating movements, are housed in the inspection tool assembly, which can be mounted on or removed from the pressure tubes by remote control using a refuelling machine. The ultrasonic technique has been adopted for measurement of the internal diameter in order to shorten the inspection time. The detectors, TV camera and electronic components used in the inspection tool assembly were selected on the basis of irradiation test results. Before inspection of the Fugen reactor, the total system was tested on a mock-up pressure tube to confirm its functions, performance, durability and reliability. The test results were: (1) the ultrasonic flaw detector can detect an artificial flaw of 2.0 mm in length, 0.1 mm in width and 0.1 mm in depth with S/N=7 dB; (2) the inside diameter measurement system can measure the inside diameter, ranging from 117.5 to 119.5 mm, with an accuracy of +-20 μm; (3) an artificial flaw of 2.0 mm in length, 0.1 mm in width and 0.1 mm in depth can be observed by the internal surface observation system. The equipment was used for the inspection of ten pressure tubes of the Fugen reactor during the May 1984 annual inspection. No degradation of the performance of the equipment was observed even after 55 hours of inspection under a maximum dose rate of 2.5x105 R/h. Based on these results, the functions and performance of the equipment in practical use were fully confirmed. (author)

  4. A Model to Reproduce the Response of the Gaseous Fission Product Monitor (GFPM) in a CANDU{sup R} 6 Reactor (An Estimate of Tramp Uranium Mass in a Candu Core)

    Energy Technology Data Exchange (ETDEWEB)

    Mostofian, Sara; Boss, Charles [AECL Atomic Energy of Canada Limited, 2251 Speakman Drive, Mississauga Ontario L5K 1B2 (Canada)

    2008-07-01

    In a Canada Deuterium Uranium (Candu) reactor, the fuel bundles produce gaseous and volatile fission products that are contained within the fuel matrix and the welded zircaloy sheath. Sometimes a fuel sheath can develop a defect and release the fission products into the circulating coolant. To detect fuel defects, a Gaseous Fission Product Monitoring (GFPM) system is provided in Candu reactors. The (GFPM) is a gamma ray spectrometer that measures fission products in the coolant and alerts the operator to the presence of defected fuel through an increase in measured fission product concentration. A background fission product concentration in the coolant also arises from tramp uranium. The sources of the tramp uranium are small quantities of uranium contamination on the surfaces of fuel bundles and traces of uranium on the pressure tubes, arising from the rare defected fuel element that released uranium into the core. This paper presents a dynamic model that reproduces the behaviour of a GFPM in a Candu 6 plant. The model predicts the fission product concentrations in the coolant from the chronic concentration of tramp uranium on the inner surface of the pressure tubes (PT) and the surface of the fuel bundles (FB) taking into account the on-power refuelling system. (authors)

  5. Sectional replacement of high pressure feedwater heater tubing

    Energy Technology Data Exchange (ETDEWEB)

    Bolton, J.A.; Bowes, P.D. [TransAlta Utilities Corp., Duffield, Alberta (Canada). Plant Engineering Services

    1994-12-31

    TransAlta Utilities is a Canadian Corporation which owns and operates the coal fired Sundance Generating Station located in central Alberta. Sundance is fitted with vertical channel down, carbon steel tubed, high pressure feedwater heaters. The primary mode of failure of these HP feedwater heaters on the six generating units is steam inlet area tube erosion and vibration damage. This damage is initiated with the deterioration of the desuperheating inlet shroud and backing plate, primarily due to thermal fatigue, thus allowing direct impingement of high velocity steam and entrained condensate upon the tubing. Topics discussed are: review of the design and conditions of the heater which allowed re-conditioning; cutting, lifting and supporting of the shell at an elevation sufficient to allow free access of the entire desuperheating zone; damage observed within the desuperheating and drains cooler zones; bundle reconditioning through damage tube section replacement and support plate repair techniques; design/installation of the desuperheating, drains-cooling zone shrouds and backing plates; benefits that this type of approach may offer; conclusions.

  6. A statistical method for draft tube pressure pulsation analysis

    Science.gov (United States)

    Doerfler, P. K.; Ruchonnet, N.

    2012-11-01

    Draft tube pressure pulsation (DTPP) in Francis turbines is composed of various components originating from different physical phenomena. These components may be separated because they differ by their spatial relationships and by their propagation mechanism. The first step for such an analysis was to distinguish between so-called synchronous and asynchronous pulsations; only approximately periodic phenomena could be described in this manner. However, less regular pulsations are always present, and these become important when turbines have to operate in the far off-design range, in particular at very low load. The statistical method described here permits to separate the stochastic (random) component from the two traditional 'regular' components. It works in connection with the standard technique of model testing with several pressure signals measured in draft tube cone. The difference between the individual signals and the averaged pressure signal, together with the coherence between the individual pressure signals is used for analysis. An example reveals that a generalized, non-periodic version of the asynchronous pulsation is important at low load.

  7. CANDU in-reactor quantitative visual-based inspection techniques

    Science.gov (United States)

    Rochefort, P. A.

    2009-02-01

    This paper describes two separate visual-based inspection procedures used at CANDU nuclear power generating stations. The techniques are quantitative in nature and are delivered and operated in highly radioactive environments with access that is restrictive, and in one case is submerged. Visual-based inspections at stations are typically qualitative in nature. For example a video system will be used to search for a missing component, inspect for a broken fixture, or locate areas of excessive corrosion in a pipe. In contrast, the methods described here are used to measure characteristic component dimensions that in one case ensure ongoing safe operation of the reactor and in the other support reactor refurbishment. CANDU reactors are Pressurized Heavy Water Reactors (PHWR). The reactor vessel is a horizontal cylindrical low-pressure calandria tank approximately 6 m in diameter and length, containing heavy water as a neutron moderator. Inside the calandria, 380 horizontal fuel channels (FC) are supported at each end by integral end-shields. Each FC holds 12 fuel bundles. The heavy water primary heat transport water flows through the FC pressure tube, removing the heat from the fuel bundles and delivering it to the steam generator. The general design of the reactor governs both the type of measurements that are required and the methods to perform the measurements. The first inspection procedure is a method to remotely measure the gap between FC and other in-core horizontal components. The technique involves delivering vertically a module with a high-radiation-resistant camera and lighting into the core of a shutdown but fuelled reactor. The measurement is done using a line-of-sight technique between the components. Compensation for image perspective and viewing elevation to the measurement is required. The second inspection procedure measures flaws within the reactor's end shield FC calandria tube rolled joint area. The FC calandria tube (the outer shell of the FC) is

  8. An integrated CANDU system

    International Nuclear Information System (INIS)

    Twenty years of experience have shown that the early choices of heavy water as moderator and natural uranium as fuel imposed a discipline on CANDU design that has led to outstanding performance. The integrated structure of the industry in Canada, incorporating development, design, supply, manufacturing, and operation functions, has reinforced this performance and has provided a basis on which to continue development in the future. These same fundamental characteristics of the CANDU program open up propsects for further improvements in economy and resource utilization through increased reactor size and the development of the thorium fuel cycle

  9. Fuel for advanced CANDU reactors

    International Nuclear Information System (INIS)

    The CANDU reactor system has proven itself to be a world leader in terms of station availability and low total unit energy cost. In 1985 for example, four of the top ten reactor units in the world were CANDU reactors operating in South Korea and Canada. This excellent operating record requires an equivalent performance record of the low-cost, natural uranium fuel. Future CANDU reactors will be an evolution of the present design. Engineering work is under way to refine the existing CANDU 600 and to incorporate state-of-the-art technology, reducing the capital cost and construction schedule. In addition, a smaller CANDU 300 plant has been designed using proven CANDU 600 technology and components but with an innovative new plant layout that makes it cost competitive with coal fired plants. For the long term, work on advanced fuel cycles and major system improvements is underway ensuring that CANDU plants will stay competitive well into the next century

  10. Simulation-based reactor control design methodology for CANDU 9

    Energy Technology Data Exchange (ETDEWEB)

    Kattan, M.K.; MacBeth, M.J. [Atomic Energy of Canada Limited, Saskatoon, Saskatchewan (Canada); Chan, W.F.; Lam, K.Y. [Cassiopeia Technologies Inc., Toronto, Ontario (Canada)

    1996-07-01

    The next generation of CANDU nuclear power plant being designed by AECL is the 900 MWe CANDU 9 station. This design is based upon the Darlington CANDU nuclear power plant located in Ontario which is among the world leading nuclear power stations for highest capacity factor with the lowest operation, maintenance and administration costs in North America. Canadian-designed CANDU pressurized heavy water nuclear reactors have traditionally been world leaders in electrical power generation capacity performance. This paper introduces the CANDU 9 design initiative to use plant simulation during the design stage of the plant distributed control system (DCS), plant display system (PDS) and the control centre panels. This paper also introduces some details of the CANDU 9 DCS reactor regulating system (RRS) control application, a typical DCS partition configuration, and the interfacing of some of the software design processes that are being followed from conceptual design to final integrated design validation. A description is given of the reactor model developed specifically for use in the simulator. The CANDU 9 reactor model is a synthesis of 14 micro point-kinetic reactor models to facilitate 14 liquid zone controllers for bulk power error control, as well as zone flux tilt control. (author)

  11. Trends in the capital costs of CANDU generating stations

    International Nuclear Information System (INIS)

    This paper consolidates the actual cost experience gained by Atomic Energy of Canada Limited, Ontario Hydro, and other Canadian electric utlities in the planning, design and construction of CANDU-PHWR (CANada Deuterium Uranium-Pressurized Heavy Water Reactor) generating stations over the past 30 years. For each of the major CANDU-PHWR generating stations in operation and under construction in Canada, an analysis is made to trace the evolution of the capital cost estimates. Major technical, economic and other parameters that affect the cost trends of CANDU-PHWR generating stations are identified and their impacts assessed. An analysis of the real cost of CANDU generating stations is made by eliminating interest during construction and escalation, and the effects of planned deferment of in-service dates. An historical trend in the increase in the real cost of CANDU power plants is established. Based on the cost experience gained in the design and construction of CANDU-PHWR units in Canada, as well as on the assessment of parameters that influence the costs of such projects, the future costs of CANDU-PHWRs are presented

  12. Intergranular susceptibility in failures of high pressure tubes

    International Nuclear Information System (INIS)

    This work addresses the influence of metallurgical susceptibility to intergranular cracking on the repeated cracking and failure of thick wall curved steel tubes from a petrochemical reactor. These tubes are made of HP-4 steel, bent and heat treated, and then subjected to autofrettage. Internal pressure is around 250 MPa. All failures are characterized by strongly branched, mostly circumferential multiple intergranular cracks. Most cracks initiated in the outer surface, in contact with steam; these were related to stress corrosion cracking (SCC). Some cracks initiated in pre defects in the inner surface, in contact with a polymer, and in the mid thickness of the tube wall. This study includes the assessment of deformation and temperature induced embrittlement mechanisms, measurement of longitudinal residual stresses, and mechanical testing included tensile, Charpy and SCC tests. Susceptibility to intergranular cracking was experimentally assessed by recreating conditions of embrittlement by thermal treatments and tensile testing. Samples with 0, 3 and 5% plastic deformations were subjected to 24 h thermal treatments between 300 and 400 deg. C. Under the conditions of previous plastic deformation due to bending and autofrettage it was possible to recreate intergranular embrittlement at service temperatures, a phenomenon similar to temper embrittlement. The process of forming the bent created localized yielding and large longitudinal residual stresses. Recovery measures, mostly relying on thermal treatments, were defined

  13. CANDU, building the future

    Energy Technology Data Exchange (ETDEWEB)

    Stern, F. [Stern Laboratories (Canada)

    1997-07-01

    The CEO of Stern Laboratories delivered a speech on the problems and challenges facing the nuclear industry. The CANDU system is looked at as the practical choice for the future of our energy source. The people of the industry must be utilized and respected to deliver to the best of their ability.

  14. CANDU market prospects

    International Nuclear Information System (INIS)

    This 1994 survey of prospective markets for CANDU reactors discusses prospects in Turkey, Thailand, the Philippines, Korea, Indonesia, China and Egypt, and other opportunities, such as in fuel cycles and nuclear safety. It was concluded that foreign partners would be needed to help with financing

  15. Magnetic pressure in electromagnetic tube forming with echelon coil

    Institute of Scientific and Technical Information of China (English)

    ZHAO Zhi-heng; YU Hai-ping; LI Chun-feng; LI Zhong

    2008-01-01

    The effects of geometrical characteristics of echelon coil on the magnetic pressure distribution and their contribution to the final shape of parts were focused and investigated through experiments and numerical simulation using FEM software ANSYS.The results show that the geometrical characteristics of echelon coil play a key role in controlling the magnetic pressure acting on the tube.They show a hump·like distribution near the interface between bigger diameter region and transition region of echelon coil,and affect the final shape of tubular parts then.With the reduction of relative diameter,the magnetic pressure in smaller diameter region decreases and its distribution gradient in transition region increases.With the augment of relative length,the magnetic pressure increases in bigger diameter region,while it almost remains constant in smaller diameter region,and the gradient in transition region enhances sharply.The distribution of magnetic pressure in the axial direction of tube agrees well with the profile of specimen.

  16. CANDU steam generator life management: laboratory data and plant experience

    International Nuclear Information System (INIS)

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

  17. R and D in support of CANDU plant life management

    International Nuclear Information System (INIS)

    One of the keys to the long-term success of CANDUs is a high capacity factor over the station design life. Considerable R and D in underway at AECL to develop technologies for assessing, monitoring and mitigating the effect of plant ageing and for improving plant performance and extending plant life. To achieve longer service life and to realize high capacity factor from CANDU stations, AECL is developing new technologies to enhance fuel channel and steam generator inspection capabilities, to monitor system health, and to allow preventive maintenance and cleaning (e.g., on-line chemical cleaning processes that produce small volumes of wastes). The life management strategy for fuel channels and steam generators requires a program to inspect components on a routine basis to identify mechanisms that could potentially affect fitness-for-service. In the case of fuel channels, the strategy includes inspections for dimensional changes, flaw detection, and deuterium concentration. New techniques are been developed to enhance these inspection capabilities; examples include accurate measurement of the gap between a pressure tube and its calandria tube and rapid full-length inspections of steam generator tubes for all known flaw types. Central to life management of components are Fitness-for-Service Guidelines (FFSG) that have been developed with the CANDU Owners Group (COG) that provide a standardized method to assess the potential for propagation of flaws detected during in-service inspections, and assessment of any change in fracture characteristics of the material. FFSG continue to be improved with the development of new technologies such as the capability to credit relaxation of stresses due to creep and non-rejectable flaws in pressure tubes. Effective management of plant systems throughout their lifetime requires much more than data acquisition and display - it requires that system health is continually monitored and managed. AECL has developed a system Health Monitor

  18. Remote ultrasonic characterisation of an irradiated pressure tube from RAPS-II

    International Nuclear Information System (INIS)

    The Rajasthan Atomic Power Station Unit-2 (RAPS-2) has reached a stage of operation where the contacting pressure tubes are suspect to failure as a result of irradiation creep and displacement of the garter springs, the hot pressure tube coming in contact with the cold calandria tube. To study and assess the safety of these pressure tubes, two channels believed to be in contact with the calandria tubes, have been removed from the reactor for detailed full length post irradiation examination. Some of the test results are presented. 2 refs., 3 figs., 1 tab

  19. Incentives for improvement of CANDU

    International Nuclear Information System (INIS)

    CANDU is a relatively young technology which has demonstrated many achievements as an electrical power generation system. These achievements include an unsurpassed safety record, high annual and lifetime capacity factors, low electricity cost and a broad range of other performance strengths which together indicate that the CANDU technology is fundamentally sound. Known capabilities not yet fully exploited, such as advanced fuel cycle options, indicate that CANDU technology will continue to pay strong dividends on research, development and design investment. This provides a strong incentive for the improvement of CANDU on a continuing basis

  20. Pressure loads and structural response of the BNL high-temperature detonation tube

    OpenAIRE

    Shepherd, Joseph E.

    1992-01-01

    The high-temperature detonation tube facility being designed at Brookhaven National Laboratory must withstand dynamic pressure loads. These loads are associated with both detonations and deflagration-to-detonation transition (DDT). The present report documents the results of computations of the pressure loads and structural response. Structural response considerations indicate that radial motion of the tube is sufficiently rapid that the tube actualkly responds to the peak pressure behi...

  1. CANDU at the crossroads

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    1990-11-01

    ''Ready for the challenge of the 90s'' was the theme of this year's gathering of the Canadian Nuclear Association held in Toronto, 3-6 June. What that challenge really entails is whether the CANDU system will survive as the last remaining alternative to the light water reactor in the world reactor market, or whether it will decline into oblivion along with the Advanced Gas Cooled reactor and so many other technically excellent systems which have fallen along the way. The fate of the CANDU system will not be determined by its technical merits, nor by its impeccable safety record. It will be determined by public perceptions and by the deliberations of an Environmental Assessment Panel established by the Government of Ontario. The debate at the Association meeting is reported. (author).

  2. Heavy-water-moderated pressure-tube reactor safety

    International Nuclear Information System (INIS)

    Several countries have heavy-water-moderated, pressure-tube reactors either in commercial operation or in late prototype stages. The supporting safety research and development includes such areas as the thermohydraulics of circuit depressurization, heat transfer from the fuel, heat rejection to the moderator from dry fuel, fuel and sheath behaviour, and fuel channel integrity. We review the work done in Canada, Great Britain, Italy and Japan, and describe some of the experimental tests underlaying the methods of accident analysis. The reactors have safety systems which, in the event of an accident, are able to shut down the reactor, keep the fuel cooled, and contain any released radioactivity. We summarize the characteristics of these safety systems (shutdown, emergency core cooling, and containment) in the various reactors, and discuss other reactor characteristics which either prevent accidents or reduce their potential demand on the safety systems. (author)

  3. Middle Ear Pressure Regulation - Complementary Action of the Mastoid and Eustachian Tube

    DEFF Research Database (Denmark)

    Gaihede, Michael; Jacobsen, Henrik; Tveterås, Kjell;

    :: In some cases, MEP counter-regulation presented as Eustachian tube openings with steep and fast pressure changes toward 0 Pa, whereas in others, gradual and slow pressure changes presented related to the mastoid; these changes sometimes crossed 0 Pa into opposite pressures. In many cases...... was related to continuous regulation of smaller pressures, whereas the tube was related to intermittent regulation of higher pressures....

  4. Middle Ear Pressure Regulation - Complementary Action of the Mastoid and Eustachian Tube

    DEFF Research Database (Denmark)

    Gaihede, Michael; Dirckx, Joris J J; Jacobsen, Henrik;

    2010-01-01

    :: In some cases, MEP counter-regulation presented as Eustachian tube openings with steep and fast pressure changes toward 0 Pa, whereas in others, gradual and slow pressure changes presented related to the mastoid; these changes sometimes crossed 0 Pa into opposite pressures. In many cases...... was related to continuous regulation of smaller pressures, whereas the tube was related to intermittent regulation of higher pressures....

  5. The Feasibility of Multidimensional CFD Applied to Calandria System in the Moderator of CANDU-6 PHWR Using Commercial and Open-Source Codes

    Directory of Open Access Journals (Sweden)

    Hyoung Tae Kim

    2016-01-01

    Full Text Available The moderator system of CANDU, a prototype of PHWR (pressurized heavy-water reactor, has been modeled in multidimension for the computation based on CFD (computational fluid dynamics technique. Three CFD codes are tested in modeled hydrothermal systems of heavy-water reactors. Commercial codes, COMSOL Multiphysics and ANSYS-CFX with OpenFOAM, an open-source code, are introduced for the various simplified and practical problems. All the implemented computational codes are tested for a benchmark problem of STERN laboratory experiment with a precise modeling of tubes, compared with each other as well as the measured data and a porous model based on the experimental correlation of pressure drop. Also the effect of turbulence model is discussed for these low Reynolds number flows. As a result, they are shown to be successful for the analysis of three-dimensional numerical models related to the calandria system of CANDU reactors.

  6. Joint studies on large CANDU

    International Nuclear Information System (INIS)

    CANDU PHWRs have demonstrated generic benefits which will be continued in future designs. These include economic benefits due to low operating costs, business potential, strategic benefits due to fuel cycle flexibility and operational benefits. These benefits have been realized in Korea through the operation of Wolsong 1, resulting in further construction of PHWRs at the same site. The principal benefit, low electricity cost, is due to the high capacity factor and the low fuel cost for CANDU. The CANDU plant at Wolsong has proven to be a safe, reliable and economical electricity producer. The ability of PHWR to burn natural uranium ensures security of fuel supply. Following successful Technology Transfer via the Wolsong 2,3 and 4 project, future opportunity exists between Korea and Canada for continuing co-operation in research and development to improve the technology base, for product development partnerships, and business opportunities in marketing and building PHWR plants in third countries. High reliability, through excellent design, well-controlled operation, efficient maintenance and low operating costs is critical to the economic viability of nuclear plants. CANDU plants have an excellent performance record. The four operating CANDU 6 plants, operated by four utilities in three countries, are world performance leaders. The CANDU 9 design, with higher output capacity, will help to achieve better site utilization and lower electricity costs. Being an evolutionary design, CANDU 9 assures high performance by utilizing proven systems, and component designs adapted from operating CANDU plants (Bruce B, Darlington and CANDU 6). All system and operating parameters are within the operating proven range of current plants. KAERI and AECL have an agreement to perform joint studies on future PHWR development. The objective of the joint studies is to establish the requirements for the design of future advanced CANDU PHWR including the utility need for design improvements

  7. Proof testing of CANDU concrete containment structures

    International Nuclear Information System (INIS)

    Prior to commissioning of a CANDU reactor, a proof pressure test is required to demonstrate the structural integrity of the containment envelope. The test pressure specified by AECB Regulatory Document R-7 (1991) was selected without a rigorous consideration of uncertainties associated with estimates of accident pressure and conatinment resistance. This study was undertaken to develop a reliability-based philosophy for defining proof testing requirements that are consistent with the current limit states design code for concrete containments (CSA N287.3).It was shown that the upodated probability of failure after a successful test is always less than the original estimate

  8. A pressurized drop-tube furnace for coal reactivity studies

    Science.gov (United States)

    Ouyang, Shan; Yeasmin, Hasina; Mathews, Joseph

    1998-08-01

    The design and characterization of a pressurized drop-tube furnace for investigation of coal devolatilization, gasification, and combustion are presented. The furnace is designed for high-temperature, isothermal operation in a developing laminar flow regime. It can be operated at pressures up to 1600 kPa, and temperatures up to 1673 K, with variable reaction time, particle feeding rate, and with inert and various oxidizing atmospheres. Particle residence times can be varied between ˜0.02 and ˜10 s depending upon operating conditions and positions of injection and sampling probes. Observations ports are available for sample collections and for optical investigation of the reactions or temperature measurements. Characterization of gas temperature in the furnace shows that, although the gas temperature profile in the furnace is affected by the water-cooled injection probe, the furnace is able to achieve isothermal operation in a developing laminar flow regime. Results from a series of brown coal devolatilization tests demonstrated the suitability of the furnace for experiments in coal research.

  9. Advancing CANDU experience to the world steam generator market

    International Nuclear Information System (INIS)

    Tube degradation in certain recirculating nuclear steam generators has provided a market for steam generator replacement. Prior to this need, B and W supplied over 200 steam generators for CANDU nuclear plants. With this experience, and implementing extensive research and development improvements in material selection, design enhancements, and new manufacturing and analytical methods, B and W has supplied or secured orders for the replacement of 26 steam generators. Along with plans for new replacement orders, B and W will continue to supply steam generators for future CANDU plants. This paper will review the progression of B and W's CANDU experience to meet the replacement steam generator market, and examine the continuous improvements required for today's increasingly demanding nuclear specifications. (author). 1 tab., 4 figs

  10. Leakage Characteristics of Dual-Cannula Fenestrated Tracheostomy Tubes during Positive Pressure Ventilation: A Bench Study.

    Science.gov (United States)

    Berlet, Thomas; Marchon, Mathias

    2016-01-01

    This study compared the leakage characteristics of different types of dual-cannula fenestrated tracheostomy tubes during positive pressure ventilation. Fenestrated Portex® Blue Line Ultra®, TRACOE® twist, or Rüsch® Traceofix® tracheostomy tubes equipped with nonfenestrated inner cannulas were tested in a tracheostomy-lung simulator. Transfenestration pressures and transfenestration leakage rates were measured during positive pressure ventilation. The impact of different ventilation modes, airway pressures, temperatures, and simulated static lung compliance settings on leakage characteristics was assessed. We observed substantial differences in transfenestration pressures and transfenestration leakage rates. The leakage rates of the best performing tubes were Careful tracheostomy tube selection permits the use of fenestrated tracheostomy tubes in patients receiving positive pressure ventilation immediately after stoma formation and minimises the risk of complications caused by transfenestration gas leakage, for example, subcutaneous emphysema. PMID:27073395

  11. A Model to Reproduce the Response of the Gaseous Fission Product Monitor (GFPM) in a CANDUR 6 Reactor (An Estimate of Tramp Uranium Mass in a Candu Core)

    International Nuclear Information System (INIS)

    In a Canada Deuterium Uranium (Candu) reactor, the fuel bundles produce gaseous and volatile fission products that are contained within the fuel matrix and the welded zircaloy sheath. Sometimes a fuel sheath can develop a defect and release the fission products into the circulating coolant. To detect fuel defects, a Gaseous Fission Product Monitoring (GFPM) system is provided in Candu reactors. The (GFPM) is a gamma ray spectrometer that measures fission products in the coolant and alerts the operator to the presence of defected fuel through an increase in measured fission product concentration. A background fission product concentration in the coolant also arises from tramp uranium. The sources of the tramp uranium are small quantities of uranium contamination on the surfaces of fuel bundles and traces of uranium on the pressure tubes, arising from the rare defected fuel element that released uranium into the core. This paper presents a dynamic model that reproduces the behaviour of a GFPM in a Candu 6 plant. The model predicts the fission product concentrations in the coolant from the chronic concentration of tramp uranium on the inner surface of the pressure tubes (PT) and the surface of the fuel bundles (FB) taking into account the on-power refuelling system. (authors)

  12. Characterization of magnetically impelled arc butt welded T11 tubes for high pressure applications

    OpenAIRE

    R. Sivasankari; V. Balusamy; P.R. Venkateswaran; G. Buvanashekaran; K Ganesh Kumar

    2015-01-01

    Magnetically impelled arc butt (MIAB) welding is a pressure welding process used for joining of pipes and tubes with an external magnetic field affecting arc rotation along the tube circumference. In this work, MIAB welding of low alloy steel (T11) tubes were carried out to study the microstructural changes occurring in thermo-mechanically affected zone (TMAZ). To qualify the process for the welding applications where pressure could be up to 300 bar, the MIAB welds are studied with variations...

  13. Estimation of Aging Effects on LOHS for CANDU-6

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Yong Ki; Moon, Bok Ja; Kim, Seoung Rae [Nuclear Engineering Service and Solution Co. Ltd., Daejeon (Korea, Republic of)

    2014-05-15

    To evaluate the Wolsong Unit 1's capacity to respond to large-scale natural disaster exceeding design, the loss of heat sink(LOHS) accident accompanied by loss of all electric power is simulated as a beyond design basis accident. This analysis is considered the aging effects of plant as the consequences of LOHS accident. Various components of primary heat transport system(PHTS) get aged and some of the important aging effects of CANDU reactor are pressure tube(PT) diametral creep, steam generator(SG) U-tube fouling, increased feeder roughness, and feeder orifice degradation. These effects result in higher inlet header temperatures, reduced flows in some fuel channels, and higher void fraction in fuel channel outlets. Fresh and aged models are established for the analysis where fresh model is the circuit model simulating the conditions at retubing and aged model corresponds to the model reflecting the aged condition at 11 EFPY after retubing. CATHENA computer code[1] is used for the analysis of the system behavior under LOHS condition. The LOHS accident is analyzed for fresh and aged models using CATHENA thermal hydraulic computer code. The decay heat removal is one of the most important factors for mitigation of this accident. The major aging effect on decay heat removal is the reduction of heat transfer efficiency by steam generator. Thus, the channel failure time cannot be conservatively estimated if aged model is applied for the analysis of this accident.

  14. Development of Evaluation Technology of the Integrity of HWR Pressure Tubes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y. S.; Jeong, Y. M.; Ahn, S. B. (and others)

    2005-03-15

    Major degradation of the feeder pipe is the thinning due to the flow accelerated corrosion and the cracking in the bent region due to the stress corrosion cracking. The feeder pipe in a PHWR is a pipe to supply the coolant to the pressure tube and the heated coolant to the steam generator for power generation. Approximately 380 pipes are installed on the inlet side and outlet side each with two bent regions in the 600 MW-class PHWR. After a leakage in the bent region of the feeder pipe, it is required to examine all the pipes in order to ensure the integrity of the pressure boundaries. It is not easy, however, to examine all the pipes with the conventional ultrasonic method, because of a high dose of radiation exposure and a limited accessibility to the pipe. In order to get rid of the limited accessibility, the ultrasonic guided wave method are developed for detection and evaluation of the cracks in the feeder pipe. The dispersion mode analysis was performed for the development of long-range guided wave inspection for the feeder pipe. An analytical approach for the straight pipe as well as numerical approach for the bent pipe with 2-D FFT were accomplished. A computer program for the calculation of the dispersion curves and wave structures was developed. Based on the dispersion curves and wave structure of the feeder pipe, candidates for the optimal parameters on the frequencies and vibration modes were selected. A time-frequency analysis methodology was developed for the mode identification of received ultrasonic signal. A high power tone-burst ultrasonic system has been setup for the generation of guided waves. Various artificial notches were fabricated on the bent feeder pipes for the experiment on the flaw detection. Considering the results of dispersion analysis and field condition, the torsional vibration mode, T(0,1) is selected for the first choice. An array of electromagnetic acoustic transducers (EMAT) was designed and fabricated for the generation of T

  15. CANDU operating experience

    International Nuclear Information System (INIS)

    The CANDU-PHW program is based upon 38 years of heavy water reactor experience with 35 years of operating experience. Canada has had 72 reactor years of nuclear-electric operations experience with 10 nuclear units in 4 generating stations during a period of 18 years. All objectives have been met with outstanding performance: worker safety, public safety, environmental emissions, reliable electricity production, and low electricity cost. The achievement has been realized through total teamwork involving all scientific disciplines and all project functions (research, design, manufacturing, construction, and operation). (auth)

  16. Towards a sustainable future using pressure tube reactor technology

    International Nuclear Information System (INIS)

    We describe the contribution nuclear energy will make to global energy needs based on the sound foundation of existing technology, infrastructure, natural resources and human knowledge, while meeting the requirements of security of supply (energy independence) and growing demand. Currently all reactors internationally operate on an unsustainable once-through nuclear fuel cycle using uranium fuel. Future decisions will be increasingly based on strategic considerations involving the complete nuclear fuel cycle, including requirements related to supply assurances, resource utilization, proliferation resistance and radioactive waste disposal. Pressure tube reactor (PTR) technology using fuel channels is uniquely suited to respond to the future needs because of its inherent technical characteristics and associated fuel cycle flexibility. PTR channel technology concepts have also continued to advance based on 50 years of continuous development and improvement, with strategic considerations involving the complete nuclear fuel cycle related to: Fuel Availability and Supply Assurances, Uranium, Plutonium and Thorium utilization, Waste Minimization, Proliferation Resistance (Safeguards) ,Assured Licensability, Improved Safety Cost, Competitiveness. We show how nuclear technology development and global sustainability is determined by R and D progress, with challenging technology goals for nuclear energy systems in the four areas of sustainability, economics, safety and reliability, and proliferation resistance and physical protection, leading naturally to the next phase of PTR channel development, namely the high efficiency Supercritical Water Reactor (SCWR). Aggressive targets have been set for R and D and advanced concepts, complementary to the approaches taken in India, which support enhanced safety, cost reduction, resource sustainability, and economical and efficient operation. (author)

  17. Study of creep collapse of tubes subject to external pressure at elevated temperature

    International Nuclear Information System (INIS)

    Intermediate heat exchanger (IHX) tubes of VHTR form the boundary between the primary and secondary coolants of the reactor. The tubes are subject to external pressures at a postulated secondary coolant depressurization accident, which might lead to creep collapse. Therefore, it is necessary to ensure the integrity against creep collapse by analysis. The objective of this work is to study a simplified analytical method for predicting collapse time of a curved tube subjected to an external pressure. The study is made based on the comparison of experimental collapse time of curved and straight tubes. Creep collapse tests were conducted under an elevated temperature and an external pressure. Test results showed that curved tubes had longer collapse time than straight tubes with the same cross sectional ovality. The simplified analytical method for a curved tube is proposed in this report, which is to compute collapse time of a straight tube with the same ovality. And in this method the computed time is considered as collapse time of the curved tube. The above test results show that this simplified method gives the conservative collapse time. And it is confirmed by additional IHX tube tests that the method is applicable to creep collapse analysis of IHX tubes

  18. Cost and schedule reduction for next-generation Candu

    International Nuclear Information System (INIS)

    construction costs through more efficient work planning and use of materials, through reduced re-work and through more precise configuration management. Full-scale exploitation of AECL's electronic engineering and project management tools enables further reductions in cost. The Candu fuel-channel reactor type offers inherent manufacturing and construction advantages through the application of a simple, low-pressure low-temperature reactor vessel along with modular fuel channel technology. This leads to cost benefits and total project schedule benefits. As a result, the targets which AECL has set for replication units - overnight capital cost of $1000 US/kW and total project schedule (engineering/manufacturing/construction/commissioning) of 48 months, have been shown to be achievable for the reference NG Candu design. (authors)

  19. Investigation of pressure drop in capillary tube for mixed refrigerant Joule-Thomson cryocooler

    Energy Technology Data Exchange (ETDEWEB)

    Ardhapurkar, P. M. [Mechanical Engineering Department, Indian Institute of Technology Bombay, Mumbai, MS 400 076 India and S. S. G. M. College of Engineering Shegaon, MS 444 203 (India); Sridharan, Arunkumar; Atrey, M. D. [Mechanical Engineering Department, Indian Institute of Technology Bombay, Mumbai, MS 400 076 (India)

    2014-01-29

    A capillary tube is commonly used in small capacity refrigeration and air-conditioning systems. It is also a preferred expansion device in mixed refrigerant Joule-Thomson (MR J-T) cryocoolers, since it is inexpensive and simple in configuration. However, the flow inside a capillary tube is complex, since flashing process that occurs in case of refrigeration and air-conditioning systems is metastable. A mixture of refrigerants such as nitrogen, methane, ethane, propane and iso-butane expands below its inversion temperature in the capillary tube of MR J-T cryocooler and reaches cryogenic temperature. The mass flow rate of refrigerant mixture circulating through capillary tube depends on the pressure difference across it. There are many empirical correlations which predict pressure drop across the capillary tube. However, they have not been tested for refrigerant mixtures and for operating conditions of the cryocooler. The present paper assesses the existing empirical correlations for predicting overall pressure drop across the capillary tube for the MR J-T cryocooler. The empirical correlations refer to homogeneous as well as separated flow models. Experiments are carried out to measure the overall pressure drop across the capillary tube for the cooler. Three different compositions of refrigerant mixture are used to study the pressure drop variations. The predicted overall pressure drop across the capillary tube is compared with the experimentally obtained value. The predictions obtained using homogeneous model show better match with the experimental results compared to separated flow models.

  20. Steam generator tube failures

    International Nuclear Information System (INIS)

    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

  1. Steam generator tube failures

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-04-01

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

  2. CFX analysis of the CANDU moderator thermal-hydraulics in the Stern Lab. Test Facility

    International Nuclear Information System (INIS)

    A numerical calculation with the commercial CFD code CFX is conducted for a test facility simulating the CANDU moderator thermal-hydraulics. Two kinds of moderator thermal-hydraulic tests at Stern Laboratories Inc. were performed in the full geometric configuration of the CANDU moderator circulating vessel, which is called a Calandria, housing a matrix of horizontal rod bundles simulating the Calandria tubes. The first of these tests is the pressure drop measurement of a cross flow in the horizontal rod bundles. The other is the local temperature measurement on the cross section of the horizontal cylinder vessel simulating the Calandria. In the present study the full geometric details of the Calandria are incorporated in the grid generation of the computational domain to which the boundary conditions for each experiment are applied. The numerical solutions are reviewed and compared with the available test data. It is shown that the present CFD prediction without the empirical correlation based on the pressure drop test is in good agreement with the test results. The prediction becomes more accurate, as the flow conditions become more turbulent with a higher Reynolds number. However, the temperature fluctuation is observed during iteration steps for a steady-state simulation of the thermal-hydraulic test. This result shows that the flow and temperature distribution inside the moderator tank may not be stable in the actual test

  3. Leakage Characteristics of Dual-Cannula Fenestrated Tracheostomy Tubes during Positive Pressure Ventilation: A Bench Study

    OpenAIRE

    Thomas Berlet; Mathias Marchon

    2016-01-01

    This study compared the leakage characteristics of different types of dual-cannula fenestrated tracheostomy tubes during positive pressure ventilation. Fenestrated Portex® Blue Line Ultra®, TRACOE® twist, or Rüsch® Traceofix® tracheostomy tubes equipped with nonfenestrated inner cannulas were tested in a tracheostomy-lung simulator. Transfenestration pressures and transfenestration leakage rates were measured during positive pressure ventilation. The impact of different ventilation modes, air...

  4. Development of High-Performance Pressure Tube Material for the Canadian SCWR Concept

    Science.gov (United States)

    Walters, L.; Donohue, S.

    2016-02-01

    The Canadian super-critical water-cooled reactor (SCWR) concept is moderated by using heavy water, while the coolant is light water at 25 MPa with an inlet temperature of 625 K and an outlet temperature of 900 K. The fuel assemblies reside in vertical pressure tubes that are the pressure boundary. The pressure tubes are insulated from the fuel assemblies and operate at temperatures near the moderator temperature, at 390 K. The zirconium alloy Excel has been selected as a candidate material for the pressure tube based on favorable properties such as high strength, resistance to radiation-induced diametral strain, and high terminal solid solubility. However, significant future effort will be required to obtain material properties and crack initiation mechanisms at super-critical water (SCW) conditions to verify that annealed Excel is a viable option as a pressure tube material in the Canadian SCWR.

  5. Technology transfer in CANDU marketing

    International Nuclear Information System (INIS)

    The author discusses how the CANDU system lends itself to technology transfer, the scope of CANDU technology transfer, and the benefits and problems associated with technology transfer. The establishment of joint ventures between supplier and client nations offers benefits to both parties. Canada can offer varying technology transfer packages, each tailored to a client nation's needs and capabilities. Such a package could include all the hardware and software necessary to develop a self-sufficient nuclear infrastructure in the client nation

  6. Leakage Characteristics of Dual-Cannula Fenestrated Tracheostomy Tubes during Positive Pressure Ventilation: A Bench Study

    Directory of Open Access Journals (Sweden)

    Thomas Berlet

    2016-01-01

    Full Text Available This study compared the leakage characteristics of different types of dual-cannula fenestrated tracheostomy tubes during positive pressure ventilation. Fenestrated Portex® Blue Line Ultra®, TRACOE® twist, or Rüsch® Traceofix® tracheostomy tubes equipped with nonfenestrated inner cannulas were tested in a tracheostomy-lung simulator. Transfenestration pressures and transfenestration leakage rates were measured during positive pressure ventilation. The impact of different ventilation modes, airway pressures, temperatures, and simulated static lung compliance settings on leakage characteristics was assessed. We observed substantial differences in transfenestration pressures and transfenestration leakage rates. The leakage rates of the best performing tubes were <3.5% of the delivered minute volume. At body temperature, the leakage rates of these tracheostomy tubes were <1%. The tracheal tube design was the main factor that determined the leakage characteristics. Careful tracheostomy tube selection permits the use of fenestrated tracheostomy tubes in patients receiving positive pressure ventilation immediately after stoma formation and minimises the risk of complications caused by transfenestration gas leakage, for example, subcutaneous emphysema.

  7. Leakage Characteristics of Dual-Cannula Fenestrated Tracheostomy Tubes during Positive Pressure Ventilation: A Bench Study.

    Science.gov (United States)

    Berlet, Thomas; Marchon, Mathias

    2016-01-01

    This study compared the leakage characteristics of different types of dual-cannula fenestrated tracheostomy tubes during positive pressure ventilation. Fenestrated Portex® Blue Line Ultra®, TRACOE® twist, or Rüsch® Traceofix® tracheostomy tubes equipped with nonfenestrated inner cannulas were tested in a tracheostomy-lung simulator. Transfenestration pressures and transfenestration leakage rates were measured during positive pressure ventilation. The impact of different ventilation modes, airway pressures, temperatures, and simulated static lung compliance settings on leakage characteristics was assessed. We observed substantial differences in transfenestration pressures and transfenestration leakage rates. The leakage rates of the best performing tubes were <3.5% of the delivered minute volume. At body temperature, the leakage rates of these tracheostomy tubes were <1%. The tracheal tube design was the main factor that determined the leakage characteristics. Careful tracheostomy tube selection permits the use of fenestrated tracheostomy tubes in patients receiving positive pressure ventilation immediately after stoma formation and minimises the risk of complications caused by transfenestration gas leakage, for example, subcutaneous emphysema.

  8. Computation and measurement of calandria tube sag in pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Calandria tubes and liquid injection shutdown system nozzles in a pressurized heavy water reactor are to sag due to irradiation creep and growth during plant operation. When the sag of calandria tube becomes bigger, the calandria tube possibly comes in contact with liquid injection shutdown system tube crossing beneath the calandria tube. The contact subsequently may cause the damage on the calandria tube resulting in unpredicted outage of the plant. It is therefore necessary to check the gap between the two tubes in order to periodically confirm no contact by using a proper measure during the plant life. An ultrasonic gap measurement probe assembly which can be inserted into two viewing ports of the calandria was developed in Korea and utilized to measure the sag between both tubes in the reactor. It was found that the centerlines of calandria tubes and liquid injection shutdown system tubes can be precisely detected by ultrasonic wave. The gaps between two tubes were easily obtained from the relative distance of the measured centerline elevations of the tubes. Based on the irradiation creep equation and the measurement data, a computer program to calculate the sags was also developed. With the computer program, the sag at the end of plant life was predicted

  9. Tube Plugging Criteria for the High-pressure Heaters of Ulchin NPP 3 and 4

    International Nuclear Information System (INIS)

    Power generation field urges nuclear power plants to reduce operating and maintaining costs to remain competitive. To reduce the cost by means of preventing the lowering thermal efficiency, the inspection of balance-of-plant heat exchanger, which was treated as not important work, becomes important. The tubing materials and tube thickness of heat exchangers in nuclear power plants are selected to withstand system temperature, pressure, and corrosion. But tubes have experienced leaks and failures and plugged based upon eddy current testing (ET) results. There are some problems for plugging the heat exchanger tubes since the criterion and its basis are not clearly described. For this reason, the criteria for the tube wall thickness are addressed in order to operate the heat exchangers in nuclear power plant without trouble during the cycle. The feed water heater is a kind of heat exchanger which raises the temperature of water supplied from the condenser. The heat source of high-pressure heaters is the extraction steam from the high-pressure turbine and moisture separator re-heater. If the tube wall of the heater is broken, the feed water flowing inside the tube intrudes to shell side. This forces the turbine to be stop in order to protect it. There are many codes and standards to be referred for calculating the minimum thickness of the heat exchanger tube in the designing stage. However, the codes and standards related to show the tube plugging criteria may not exist currently. In this paper, a method to establish the tube plugging criteria of BOP heat exchangers is introduced and the tube plugging criteria for the high pressure heaters of Ulchin NPP No. 3 and 4. This method relies on the similar plugging criteria used in the steam generator tubes

  10. The CANDU experience in Romania

    International Nuclear Information System (INIS)

    The CANDU program in Romania is now well established. The Cernavoda Nuclear Station presently under construction will consist of 5-CANDU 600 MWE Units and another similar size station is planned to be in operation in the next decade. Progress on the multi-unit station at Cernavoda was stalled for 18 months in 1982/83 as the Canadian Export Development Corporation had suspended their loan disbursements while the Romanian National debt was being rescheduled. Since resumption of the financing in August 1983 contracts worth almost 200M dollars have been placed with Canadian Companies for the supply of major equipment for the first two units. The Canadian design is that which was used in the latest 600 MWE CANDU station at Wolsong, Korea. The vast construction site is now well developed with the cooling water systems/channels and service buildings at an advanced stage of completion. The perimeter walls of the first two reactor buildings are already complete and slip-forming for the 3rd Unit is imminent. Many Romanian organizations are involved in the infrastructure which has been established to handle the design, manufacture, construction and operation of the CANDU stations. The Romanian manufacturing industry has made extensive preparations for the supply of CANDU equipment and components, and although a major portion of the first two units will come from Canada their intentions are to become largely self-supporting for the ensuing CANDU program. Quality assurance programs have been prepared already for many of the facilities

  11. CANDU safety under severe accidents

    International Nuclear Information System (INIS)

    The characteristics of the CANDU reactor relevant to severe accidents are set first by the inherent properties of the design, and second by the Canadian safety/licensing approach. Probabilistic safety assessment studies have been performed on operating CANDU plants, and on the 4 x 880 MW(e) Darlington station now under construction; furthermore a scoping risk assessment has been done for a CANDU 600 plant. They indicate that the summed severe core damage frequency is of the order of 5 x 10-6/year. CANDU nuclear plant designers and owner/operators share information and operational experience nationally and internationally through the CANDU Owners' Group (COG). The research program generally emphasizes the unique aspects of the CANDU concept, such as heat removal through the moderator, but it has also contributed significantly to areas generic to most power reactors such as hydrogen combustion, containment failure modes, fission product chemistry, and high temperature fuel behaviour. Abnormal plant operating procedures are aimed at first using event-specific emergency operating procedures, in cases where the event can be diagnosed. If this is not possible, generic procedures are followed to control Critical Safety Parameters and manage the accident. Similarly, the on-site contingency plans include a generic plan covering overall plant response strategy, and a specific plan covering each category of contingency

  12. Development of nuclear fuel. Development of CANDU advanced fuel bundle

    International Nuclear Information System (INIS)

    In order to develop CANDU advanced fuel, the agreement of the joint research between KAERI and AECL was made on February 19, 1991. AECL conceptual design of CANFLEX bundle for Bruce reactors was analyzed and then the reference design and design drawing of the advanced fuel bundle with natural uranium fuel for CANDU-6 reactor were completed. The CANFLEX fuel cladding was preliminarily investigated. The fabricability of the advanced fuel bundle was investigated. The design and purchase of the machinery tools for the bundle fabrication for hydraulic scoping tests were performed. As a result of CANFLEX tube examination, the tubes were found to be meet the criteria proposed in the technical specification. The dummy bundles for hydraulic scoping tests have been fabricated by using the process and tools, where the process parameters and tools have been newly established. (Author)

  13. Radiological Characteristics of decommissioning waste from a CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Dong Keun; Choi, Heui Joo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ahmed, Rizwan; Heo, Gyun Young [Dept. of Nuclear Engineering, Kyung Hee University, Yongin (Korea, Republic of)

    2011-11-15

    The radiological characteristics for waste classification were assessed for neutron-activated decommissioning wastes from a CANDU reactor. The MCNP/ORIGEN2 code system was used for the source term analysis. The neutron flux and activation cross-section library for each structural component generated by MCNP simulation were used in the radionuclide buildup calculation in ORIGEN2. The specific activities of the relevant radionuclides in the activated metal waste were compared with the specified limits of the specific activities listed in the Korean standard and 10 CFR 61. The time-average full-core model of Wolsong Unit 1 was used as the neutron source for activation of in-core and ex-core structural components. The approximated levels of the neutron flux and cross-section, irradiated fuel composition, and a geometry simplification revealing good reliability in a previous study were used in the source term calculation as well. The results revealed the radioactivity, decay heat, hazard index, mass, and solid volume for the activated decommissioning waste to be 1.04 x 10{sup 16} Bq, 2.09 x 10{sup 3} W, 5.31 x 10{sup 14} m{sup 3}-water, 4.69 x 10{sup 5} kg, and 7.38 x 10{sup 1} m{sup 3}, respectively. According to both Korean and US standards, the activated waste of the pressure tubes, calandria tubes, reactivity devices, and reactivity device supporters was greater than Class C, which should be disposed of in a deep geological disposal repository, whereas the side structural components were classified as low- and intermediate-level waste, which can be disposed of in a land disposal repository. Finally, this study confirmed that, regardless of the cooling time of the waste, 15% of the decommissioning waste cannot be disposed of in a land disposal repository. It is expected that the source terms and waste classification evaluated through this study can be widely used to establish a decommissioning/disposal strategy and fuel cycle analysis for CANDU reactors.

  14. Fracture toughness of irradiated Zr–2.5Nb pressure tube from Indian PHWR

    International Nuclear Information System (INIS)

    Fracture toughness of irradiated Zr–2.5Nb alloy pressure tube, fabricated by the cold pilgering and stress relieving route, was evaluated using disk compact tension type specimens. These specimens were punched out from the irradiated pressure tube (S-07), which was in service for about 8 effective full power years of reactor operation in the Kakrapar Atomic Power Station-2 (KAPS-2). The tests were carried out remotely inside a lead shielded enclosure. Crack growth during the test was measured using the direct current potential drop technique. The irradiated pressure tube showed low fracture toughness at 25 °C. The fracture toughness increased with increase in temperature up to 250 °C but was practically unaffected with further increase in temperature up to 300 °C. This paper discusses the fracture behavior of irradiated Indian pressure tube material and compares it with other data available

  15. In-service inspection of zircaloy pressure tube of CIRUS reactor

    International Nuclear Information System (INIS)

    The pressurized water loop (PWL) of Cirus uses a 10 meter long zircaloy tube of 57.9 mm ID and 5.4 mm wall thickness. The loop has been used for irradiation testing of various experimental fuel pins since 1972. As part of the refurbishment programme, the condition of Zircaloy-2 pressure tube of the pressurized water loop was investigated by Eddy current and ultrasonic testing. The eddy current probe was balanced over a portion of the tube and the differential signals were recorded for the entire length of the tube. For ultrasonic flaw scanning, gadgets were fabricated and scanning was carried out to evaluate the condition of irradiated pressure tube. For ultrasonic testing an annular probe holder matching to the internal diameter of the zircaloy tube was used for immersion scanning. The probe holder fitted with 10 MHz line focused ultrasonic probes inclined at 28 deg in axial and circumferential directions. A normal spot focused probe was also used to measure wall thickness and detection of laminar flaws. Axial and circumferential grooves of 3% wall thickness depth on ID and OD were used as standard calibration defects. The eddy current and ultrasonic tests did not detect any defect of unacceptable size in the zircaloy pressure tube. (author)

  16. Enhanced candu 6 reactor: status

    International Nuclear Information System (INIS)

    The CANDU 6 power reactor is visionary in its approach, renowned for its on-power refuelling capability and proven over years of safe, economical and reliable power production. Developed by Atomic Energy of Canada Limited (AECL), the CANDU 6 design offers excellent performance utilizing state-of-the-art technology. The first CANDU 6 plants went into service in the early 1980s as leading edge technology and the design has been continuously advanced to maintain superior performance with an outstanding safety record. The first set of CANDU 6 plants - Gentilly 2 and Point Lepreau in Canada, Embalse in Argentina and Wolsong- Unit 1 in Korea - have been in service for more than 22 years and are still producing electricity at peak performance; to the end of 2004, their average Lifetime Capacity Factor was 83.2%. The newer CANDU 6 units in Romania (Cernavoda 1), Korea (Wolsong-Units 2, 3 and 4) and Qinshan (Phase III- Units 1 and 2) have also been performing at outstanding levels. The average lifetime Capacity Factor of the 10 CANDU 6 operating units around the world has been 87% to the end of 2004. Building on these successes, AECL is committed to the further development of this highly successful design, now focussing on meeting customers' needs for reduced costs, further improvements to plant operation and performance, enhanced safety and incorporating up-to-date technology, as warranted. This has resulted in AECL embarking on improving the CANDU 6 design through an upgraded product termed the ''Enhanced CANDU 6'' (EC6), which incorporates several attractive but proven features that make the CANDU 6 reactor even more economical, safer and easier to operate. Some of the key features that are being incorporated into the EC6 include increasing the plant's power output, shortening the overall project schedule, decreasing the capital cost, dealing with obsolescence issues, optimizing maintenance outages and incorporating lessons learnt through feedback obtained from the

  17. Stress and integrity analysis of steam superheater tubes of a high pressure boiler

    OpenAIRE

    Daniel Leite Cypriano Neves; Jansen Renato de Carvalho Seixas; Ediberto Bastos Tinoco; Adriana da Cunha Rocha; Ibrahim de Cerqueira Abud

    2004-01-01

    Sources that can lead to deterioration of steam superheater tubes of a high pressure boiler were studied by a stress analysis, focused on internal pressure and temperature experienced by the material at real operating conditions. Loss of flame control, internal deposits and unexpected peak charge are factors that generate loads above the design limit of tube materials, which can be subjected to strain, buckling, cracks and finally rupture in service. To evaluate integrity and dependability of...

  18. Second International Conference on CANDU Fuel

    International Nuclear Information System (INIS)

    Thirty-four papers were presented at this conference in sessions dealing with international experience and programs relating to CANDU fuel; fuel manufacture; fuel behaviour; fuel handling, storage and disposal; and advanced CANDU fuel cycles. (L.L.)

  19. Experimental Study for Pressure Drop of Viscoelastic Fluids through Periodically Sudden COnverging Diverging Tube

    Institute of Scientific and Technical Information of China (English)

    1996-01-01

    The purpose of this work is to investigate experimentally the pressure drop of drag-reducing polymer solutions in the fully developed flow region in a periodically sudden converging-diverging tube as well as in a straight tube.Testing fluids were aqueous polyacrylamide solutions with concentration ranged from 200 to 1400 w.p.p.m.,and the Reynolds number ranged from 5×103to 7×104.Drag-reducing phenomenon is found to exist in the straight tube flow with the viscoelastic fluid,while in the periodically sudden converging-diverging tube the friction factor is insensitive to the concentration of solution.

  20. A general computing code devoted to the analysis of bending vibrations specific to the CANDU type fuel channel

    International Nuclear Information System (INIS)

    It is known that circulation of the coolant through the pressure tube of a CANDU type reactor initiates and maintains bending vibrations in: individual fuel elements, fuel cluster, cluster column and in the pressure tube. The driving forces are either aleatory, due to turbulent flow, or harmonical due to the pressure pulsations from the circulation pumps. The vibrations induced by laminar flow in case of excessive intensities may induce both a acceleration of the fretting wear phenomena in the fuel elements and pressure tubes and a premature aging of the latter. In these conditions an important problem in the cluster design is that of obtaining, based on knowledge of laminar flow frequency structure, the eigenfrequencies for the four categories of oscillatory systems mentioned above and thus to avoid by construction the resonance phenomenon or at least to diminish its impairing effects. An activity of comparative analysis in different fuel cluster types is underway at INR Pitesti, a special attention being of course directed toward their vibrational behavior. The paper presents a general computational code devoted to characterization of bending vibration for: individual fuel elements, fuel element cluster, pressure tube loaded or not with fuel clusters and filled or not with coolant; fuel channel. During the presentation of the work the computing code will be run for demonstration

  1. Quality Products - The CANDU Approach

    International Nuclear Information System (INIS)

    The prime focus of the CANDU concept (natural uranium fuelled-heavy water moderated reactor) from the beginning has economy, heavy water losses and radiation exposures also were strong incentives for ensuring good design and reliable equipment. It was necessary to depart from previously accepted commercial standards and to adopt those now accepted in industries providing quality products. Also, through feedback from operating experience and specific design and development programs to eliminate problems and improve performance, CANDU has evolved into today's successful product and one from which future products will readily evolve. Many lessons have been learned along the way. On the one hand, short cuts of failures to understand basic requirements have been costly. On the other hand, sound engineering and quality equipment have yielded impressive economic advantages through superior performance and the avoidance of failures and their consequential costs. The achievement of lifetime economical performance demands quality products, good operation and good maintenance. This paper describes some of the basic approaches leading to high CANDU station reliability and overall excellent performance, particularly where difficulties have had to be overcome. Specific improvements in CANDU design and in such CANDU equipment as heat transport pumps, steam generators, valves, the reactor, fuelling machines and station computers, are described. The need for close collaboration among designers, nuclear laboratories, constructors, operators and industry is discussed. This paper has reviewed some of the key components in the CANDU system as a means of indicating the overall effort that is required to provide good designs and highly reliable equipment. This has required a significant investment in people and funding which has handsomely paid off in the excellent performance of CANDU stations. The close collaboration between Atomic Energy of Canada Limited, Canadian industry and the

  2. Analysis of the Internal Pressure in Tube Hydroforming and Its Experimental Investigation

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    The internal pressure of the process was studied theoretically and experimentally. The external load character and internal stress character of tube hydroforming were discussed first. Then, according to the characters,the function and classification of internal pressure were presented in general. Base on the stress analysis, its effect on the yield criterion and calculation formula were also researched and derived. To verify the correction of the theoretical analysis and derived formula, experiments with different internal pressures were carried out and the result was compared and discussed. It demonstrates that internal pressure plays an important role in tube hydroforming and theory and formula discussed and derived by this paper are feasible in practice.

  3. Development of advanced CANDU PHWR -Development of the advanced CANDU technology-

    International Nuclear Information System (INIS)

    The target of this project is to assess the feasibility of improving PHWR and to establish the parameter of the improved concept and requirements for developing it. To set up the requirements for the Improved Pressurized Heavy Water Reactor: (1) Design requirements of PHWR main systems and Safety Design Regulatory Requirements for Safety Related System i.e. Reactor Shutdown System, Emergency Core Cooling System and Containment System were prepared. (2) Licensing Basis Documents were summarized and Safety Analysis Regulatory. Requirements were reviewed and analyzed. To estimate the feasibility of improving PHWR and to establish the main parameters of the concept of new PHWR in future: (1) technical level/developing trend of PHWR in Korea through Wolsong 2, 3 and 4 design experience and Technical Transfer Program was investigated to analyze the state of basic technology and PHWR improvement potential. (2) CANDU 6 design improvement tendency, CANDU 3 design concept and CANDU 9 development state in other country was analyzed. (3) design improvement items to apply to the reactors after Wolsong 2, 3 and 4 were selected and Plant Design Requirements and Conceptual Design Description were prepared and the viability of improved HWR was estimated by analyzing economics, performance and safety. (4) PHWR technology improving research and development plan was established and international joint study initiated for main design improvement items

  4. Future generations of CANDU: advantages and development with passive safety

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) advances water reactor and CANDLT technology using an evolutionary development strategy. This strategy ensures that innovations are based firmly on current experience and keeps our development programs focused on one reactor concept, reducing risks, development costs, and product development cycle times. It also assures our customers that our products will never become obsolete or unsupported, and the continuous line of water reactor development is secure and supported into the future. Using the channel reactor advantage of modularity, the subdivided core has the advantage of passive safety by heat removal to the low- pressure moderator. With continuous improvements, the Advanced CANDU Reactor TM (ACR-1000TM) concept will likely remain highly competitive for a number of years and leads naturally to the next phase of CANDU development, namely the Generation IV CANDU -SCWR concept. This is conventional water technology, since supercritical boilers and turbines have been operating for some time in coal-fired power plants. Significant cost, safety, and performance advantages would result from the CANDU-SCWR concept, plus the flexibility of a range of plant sizes suitable for both small and large electric grids, and the ability for co-generation of electric power, process heat, and hydrogen. In CANDU-SCWR, novel developments are included in the primary circuit layout and channel design. The R and D in Canada is integrated with the Generation IV international Forum (GIF) plans, and has started on examining replaceable insulating liners that would ensure channel life, and on providing completely passive reactor decay heat removal directly to the moderator heat sink without forced cooling. In the interests of sustainability, hydrogen production by a CANDU- SCWR is also be included as part of the system requirements, where the methods for hydrogen production will depend on the outlet temperature of the reactor

  5. Results of experimental tests simulating supply pressure decrease in a K process tube

    Energy Technology Data Exchange (ETDEWEB)

    Toyoda, K.G.; Calkin, J.F.

    1957-11-13

    Simultaneous reduction of coolant to several or all reactor tubes raises concern not only for the adequacy of protection in the individual process tube but also the reactor as a whole. In event of such flow reduction, the heat generation does not decrease until at least 1.4 seconds have elapsed following the accident. Thus, the water temperature from each tube will rise, and result in an increase in the bulk water temperature. If the increase in bulk water temperature is such that saturation temperature at the top of downcomer is reached, pressurization may occur at that point and exceed the maximum recommended working pressure limit (approximately 1 to 2 psig). The purpose of this report is to present experimental data on a series of tests which were made to simulate flow reductions to a K type process tube by simulated front header pressure decreases.

  6. ACR technology for CANDU enhancements

    International Nuclear Information System (INIS)

    The ACR-1000 design retains many essential features of the original CANDU plant design. As well as further-enhanced safety, the design also focuses on operability and maintainability, drawing on valuable customer input and OPEX. The engineering development of the ACR-1000 design has been accompanied by a research and confirmatory testing program. This program has extended the database of knowledge on the CANDU design. The ACR-1000 design has been reviewed by the Canadian regulator, the Canadian Nuclear Safety Commission (CNSC) which concluded that there are no fundamental barriers to licensing the ACR-1000 design in Canada after completing three phases of the pre-project design review. The generic PSAR for the ACR-1000 design was completed in September 2009. The PSAR contains the ACR-1000 design details, the safety and design methodology, and the safety analysis that demonstrate the ACR-1000 safety case and compliance with Canadian and international regulatory requirements and expectations. The ACR technology developed during the ACR-1000 Engineering Program and the supporting development testing has had a major impact beyond the ACR program itself: Improved CANDU components and systems; Enhanced engineering processes and engineering tools, which lead to better product quality, and better project efficiency; and Improved operational performance. This paper provides a summary of technology arising from the ACR program that has been incorporated into new CANDU designs such as the EC6, or can be applied for servicing operating CANDU reactors. (author)

  7. Ultrasonic testing of pressure contact welded joints of heterogeneous tubes

    International Nuclear Information System (INIS)

    A method of ultrasonic testing of welded joints of tubes of heterogeneous 12Kh1MF and 1Kh18N12T steels is described. The tubes are 32 to 57 mm in diameter with the walls 4 to 6 mm thick. A prism of a serial inclined converter rated at 5 MHz has been used for testing. The testing has been conducted by a singly - and doubly reflected beam at the incident angle of 50 deg. The sensitivity margin of the converter is 35 dB at a 6 to 9 dB signal/noise ratio. 25 specimens have been tested. The test results have shown that amplide of echosignal in a defective sample is by 2 dB higher as compared with the reference signal. Criteria according to which a sample is considered to be defective are given

  8. Advanced eddy current technique for measurement of annular gap between pressure tube and calandria tube in Indian Pressurized Heavy Water Reactors (PHWRs)

    International Nuclear Information System (INIS)

    In Indian Pressurised Heavy Water Reactors (PHWRs), the PT (pressure tube) is designed to be nominally concentric with the encircling CT (calandria tube). Due to various factors PT becomes eccentric with respect to CT over the life of reactor. If this becomes excessive, hot PT will come in contact with cold CT. Such a cold spot could act as potential location for initiating blister formation and premature failure of PT. Hence it is important to periodically measure annular gap between PT and CT. An advanced eddy current technique has been successfully developed and incorporated in BARCIS (BARC Channel Inspection System) for measurement of PT-CT gap. (author). 4 refs., 3 figs

  9. Development of CANDU ECCS performance evaluation methodology and guides

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Kwang Hyun; Park, Kyung Soo; Chu, Won Ho [Korea Maritime Univ., Jinhae (Korea, Republic of)

    2003-03-15

    The objectives of the present work are to carry out technical evaluation and review of CANDU safety analysis methods in order to assist development of performance evaluation methods and review guides for CANDU ECCS. The applicability of PWR ECCS analysis models are examined and it suggests that unique data or models for CANDU are required for the following phenomena: break characteristics and flow, frictional pressure drop, post-CHF heat transfer correlations, core flow distribution during blowdown, containment pressure, and reflux rate. For safety analysis of CANDU, conservative analysis or best estimate analysis can be used. The main advantage of BE analysis is a more realistic prediction of margins to acceptance criteria. The expectation is that margins demonstrated with BE methods would be larger that when a conservative approach is applied. Some outstanding safety analysis issues can be resolved by demonstration that accident consequences are more benign than previously predicted. Success criteria for analysis and review of Large LOCA can be developed by top-down approach. The highest-level success criteria can be extracted from C-6 and from them, the lower level criteria can be developed step-by-step, in a logical fashion. The overall objectives for analysis and review are to verify radiological consequences and frequency are met.

  10. In vitro evaluation of the method effectiveness to limit inflation pressure cuffs of endotracheal tubes

    Directory of Open Access Journals (Sweden)

    Rafael de Macedo Coelho

    2016-04-01

    Full Text Available ABSTRACT BACKGROUND AND OBJECTIVE: Cuffs of tracheal tubes protect the lower airway from aspiration of gastric contents and facilitate ventilation, but may cause many complications, especially when the cuff pressure exceeds 30 cm H2O. This occurs in over 30% of conventional insufflations, so it is recommended to limit this pressure. In this study we evaluated the in vitro effectiveness of a method of limiting the cuff pressure to a range between 20 and 30 cm H2O. METHOD: Using an adapter to connect the tested tube to the anesthesia machine, the relief valve was regulated to 30 cm H2O, inflating the cuff by operating the rapid flow of oxygen button. There were 33 trials for each tube of three manufacturers, of five sizes (6.5-8.5, using three times inflation (10, 15 and 20 s, totaling 1485 tests. After inflation, the pressure obtained was measured with a manometer. Pressure >30 cm H2O or <20 cm H2O were considered failures. RESULTS: There were eight failures (0.5%, 95% CI: 0.1-0.9%, with all by pressures <20 cm H2O and after 10 s inflation (1.6%, 95% CI: 0 5-2.7%. One failure occurred with a 6.5 tube (0.3%, 95% CI: -0.3 to 0.9%, six with 7.0 tubes (2%, 95% CI: 0.4-3.6%, and one with a 7.5 tube (0.3%, 95% CI: -0.3 to 0.9%. CONCLUSION: This method was effective for inflating tracheal tube cuffs of different sizes and manufacturers, limiting its pressure to a range between 20 and 30 cm H2O, with a success rate of 99.5% (95% CI: 99.1-99.9%.

  11. THE EFFECTS OF AREA CONTRACTION ON SHOCK WAVE STRENGTH AND PEAK PRESSURE IN SHOCK TUBE

    Directory of Open Access Journals (Sweden)

    A. M. Mohsen

    2012-06-01

    Full Text Available This paper presents an experimental investigation into the effects of area contraction on shock wave strength and peak pressure in a shock tube. The shock tube is an important component of the short duration, high speed fluid flow test facility, available at the Universiti Tenaga Nasional (UNITEN, Malaysia. The area contraction was facilitated by positioning a bush adjacent to the primary diaphragm section, which separates the driver and driven sections. Experimental measurements were performed with and without the presence of the bush, at various diaphragm pressure ratios, which is the ratio of air pressure between the driver (high pressure and driven (low pressure sections. The instantaneous static pressure variations were measured at two locations close to the driven tube end wall, using high sensitivity pressure sensors, which allow the shock wave strength, shock wave speed and peak pressure to be analysed. The results reveal that the area contraction significantly reduces the shock wave strength, shock wave speed and peak pressure. At a diaphragm pressure ratio of 10, the shock wave strength decreases by 18%, the peak pressure decreases by 30% and the shock wave speed decreases by 8%.

  12. A model for analyzing CANDU-6 SDS No.2 poison injection system

    International Nuclear Information System (INIS)

    In CANDU reactor one of the two reactor shutdown systems is the liquid poison injection system which injects the highly pressurized liquid neutron poison into the moderator tank via small holes on the nozzle pipes. To ensure the safe shutdown of a reactor it is necessary for the poison curtains generated by jets provide quick, and enough negative reactivity to the reactor during the early stage of the accident. In order to produce the neutron cross section necessary to perform this work, the poison concentration distribution during the transient is necessary. In this study, a set of models for analyzing the transient poison concentration induced by this high pressure poison injection jet activated upon the reactor trip in a CANDU-6 reactor moderator tank has been developed and used to generate the poison concentration distribution of the poison curtains induced by the high pressure jets injected into the vacant region between the pressure tube banks. The poison injection rate through the jet holes drilled on the nozzle pipes is obtained by a 1-D transient hydrodynamic code called. ALITRIG, and this injection rate is used to provide the inlet boundary condition to a 3-D CFD model of the moderator tank based on CFX4.3, a CFD code, to simulate the formation of the poison jet curtain inside the moderator tank. For the validation, an attempt was made to validate this model against a poison injection experiment performed at Bhabha Atomic Research Center (BARC) of India. The interim progress will be presented and the validation analysis result is discussed

  13. Build your own Candu reactor

    International Nuclear Information System (INIS)

    The author discusses the marketing of Candu reactors, particularly the export trade. Future sales will probably be of the nuclear side of a station only, thus striking a compromise between licensing and 'turnkey' sales. It is suggested that AECL might have made more money in the past had it not given the right to manufacture Candu fuel away to Canadian industry. Future sales to certain potential customers may be limited by the requirement of strict safeguards, which will almost certainly never be relaxed. (N.D.H.)

  14. Neutronics-thermalhydraulics coupling in a CANDU SCWR

    Science.gov (United States)

    Adouki, Pierre

    In order to implement new nuclear technologies as a solution to the growing demand for energy, 10 countries agreed on a framework for international cooperation in 2002, to form the Generation IV International Forum (GIF). The goal of the GIF is to design the next generation of nuclear reactors that would be cost effective and would enhance safety. This forum has proposed several types of Generation IV reactors including the Supercritical Water-Cooled Reactor (SCWR). The SCWR comes in two main configurations: pressure vessel SCWR and pressure tube SCWR (PT-SCWR). In this study, the CANDU SCWR (a PT-SCWR) is considered. This reactor is oriented vertically and contains 336 channels with a length of 5 m. The target coolant inlet and outlet temperatures are 350 Celsius and 625 Celsius, respectively. The coolant flows downwards, and the reactor power is 2540 MWth. Various fuel designs have been considered in order not to exceed the linear element rating. However, the dependency between the core power and thermalhydraulics parameters results in the necessity to use a neutronics/thermalhydaulics coupling scheme to determine the core power and the thermalhydraulics parameters. The core power obtained has a power peaking factor of 1.4. The bundle power distribution for all channels has a peak at the third bundle from the inlet, but the value of this peak increases with the channel power. The heat-transfer coefficient and the specific-heat capacity have a peak at the same location in a channel, and this location shifts toward the inlet as the channel power increases. The exit coolant temperature increases with the channel power, while the exit coolant density and pressure decrease with the channel power. Also, higher channel powers lead to higher fuel and cladding temperatures. Moreover, as the coupling method is applied, the effective multiplication factor and the values of thermalhydaulics parameters oscillate as they converge.

  15. Pressure tests to assess the significance of defects in boiler and superheater tubing

    International Nuclear Information System (INIS)

    Internal pressure tests on 9 per cent Cr-1 per cent Mo steel tubing containing artificial defects demonstrated that the resultant loss of strength was less than a simple calculation based on the reduced tube thickness would suggest. Bursting tests on tubes containing longitudinal defects of varying length, depth and acuity showed notch strengthening at ambient temperature and at 5500C. A flow stress concept developed for simple bursting tests was shown to apply to creep conditions at 5500C. Results of creep and short-term bursting tests show that the length as well as the depth of the defect is an important factor affecting the life of bursting strength of the tubes. Defects less than 10 per cent of the tube thickness were found to have an insignificant effect. (author)

  16. Studies on an improved indigenous pressure wave generator and its testing with a pulse tube cooler

    Science.gov (United States)

    Jacob, S.; Karunanithi, R.; Narsimham, G. S. V. L.; Kranthi, J. Kumar; Damu, C.; Praveen, T.; Samir, M.; Mallappa, A.

    2014-01-01

    Earlier version of an indigenously developed Pressure Wave Generator (PWG) could not develop the necessary pressure ratio to satisfactorily operate a pulse tube cooler, largely due to high blow by losses in the piston cylinder seal gap and due to a few design deficiencies. Effect of different parameters like seal gap, piston diameter, piston stroke, moving mass and the piston back volume on the performance is studied analytically. Modifications were done to the PWG based on analysis and the performance is experimentally measured. A significant improvement in PWG performance is seen as a result of the modifications. The improved PWG is tested with the same pulse tube cooler but with different inertance tube configurations. A no load temperature of 130 K is achieved with an inertance tube configuration designed using Sage software. The delivered PV power is estimated to be 28.4 W which can produce a refrigeration of about 1 W at 80 K.

  17. Review of the Safety Concern Related to CANDU Moderator Temperature Distribution and Status of KAERI Moderator Circulation Test (MCT) Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Rhee, Bo W.; Kim, Hyoung T. [Severe Accident and PHWR Safety Research Division, Daejeon (Korea, Republic of); Kim, Tongbeum [University of the Witwatersrand, Johannesburg (South Africa); Im, Sunghyuk [KAIST, Daejeon (Korea, Republic of)

    2015-10-15

    Following a large break LOCA and before Emergency Coolant Injection (ECI) initiation, pressure tubes (PT) significantly heat up as a result of the initial power pulse and degraded coolant flow. Consequently, some pressure tubes balloon and come into contact with the calandria tubes (CT). Following the PT/CT contact, the pressure tubes cool as they transfer some of the absorbed heat to the moderator via conduction at contact locations. As long as sustained calandria tube dryout does not occur, the calandria tube surface temperature remains below the creep threshold temperature and no further deformation is expected. Consequently, a sufficient condition to ensure fuel channel integrity following a large LOCA, is the avoidance of sustained calandria tubes dryout. If the moderator available subcooling at the onset of a large LOCA is greater than the subcooling requirements, a sustained calandria tube dryout is avoided. The subcooling requirements are determined from a set of experiments known as fuel channel contact experiments. The difference between available subcooling and required subcooling is called subcooling margins. The moderator flow circulation patterns are complicated slow flows that significantly vary from buoyancy dominated to inertia dominated patterns. Accurate predictions of flow patterns are essential for accurate calculation of moderator temperature distributions and the related moderator subcooling. Following a large break LOCA and before Emergency Coolant Injection (ECI) initiation, pressure tubes (PT) significantly heat up as a result of the initial power pulse and degraded coolant flow. Consequently, some pressure tubes balloon and come into contact with the calandria tubes (CT). Following the PT/CT contact, the pressure tubes cool as they transfer some of the absorbed heat to the moderator via conduction at contact locations. As long as sustained calandria tube dryout does not occur, the calandria tube surface temperature remains below the creep

  18. Creep-rupture tests of internally pressurized Hastelloy-X tubes

    Science.gov (United States)

    Gumto, K. H.; Colantino, G. J.

    1973-01-01

    Seamless Hastelloy-X tubes with 0.375-in. outside diameter and 0.025-in. wall thickness were tested to failure at temperatures from 1400 to 1650 F and internal helium pressures from 800 to 1800 psi. Lifetimes ranged from 58 to 3600 hr. The creep-rupture strength of the tubes was from 20 to 40 percent lower than that of sheet specimens. Larson-Miller correlations and photomicrographs of some specimens are presented.

  19. Tracheal tube and laryngeal mask cuff pressure during anaesthesia - mandatory monitoring is in need

    Directory of Open Access Journals (Sweden)

    Møller Ann M

    2010-12-01

    Full Text Available Abstract Background To prevent endothelium and nerve lesions, tracheal tube and laryngeal mask cuff pressure is to be maintained at a low level and yet be high enough to secure air sealing. Method In a prospective quality-control study, 201 patients undergoing surgery during anaesthesia (without the use of nitrous oxide were included for determination of the cuff pressure of the tracheal tubes and laryngeal masks. Results In the 119 patients provided with a tracheal tube, the median cuff pressure was 30 (range 8 - 100 cm H2O and the pressure exceeded 30 cm H2O (upper recommended level for 54 patients. In the 82 patients provided with a laryngeal mask, the cuff pressure was 95 (10 - 121 cm H2O and above 60 cm H2O (upper recommended level for 56 patients and in 34 of these patients, the pressure exceeded the upper cuff gauge limit (120 cm H2O. There was no association between cuff pressure and age, body mass index, type of surgery, or time from induction of anaesthesia to the time the cuff pressure was measured. Conclusion For maintenance of epithelia flow and nerve function and at the same time secure air sealing, this evaluation indicates that the cuff pressure needs to be checked as part of the procedures involved in induction of anaesthesia and eventually checked during surgery.

  20. Development and validation of a model for CANDU-6 SDS2 poison injection analysis

    International Nuclear Information System (INIS)

    In CANDU-6 reactor there are two independent reactor shutdown systems. The shutdown system no. 2(SDS2) injects the liquid poison into the moderator tank by high pressure via small holes on the 6 nozzle pipes and stops the nuclear chain reaction. To ensure the safe shutdown of a reactor loaded with either DUPIC or SEU fuels it is necessary for the poison curtains generated by jets provide quick, and enough negative reactivity to the reactor during the early stage of the accident. In order to produce the neutron cross section necessary to perform this work, the poison concentration distribution during the transient is necessary. The motivation for this work arose from the fact that the computer code package for performing this task is not transfered to Korea yet. In this study, a set of models for analyzing the transient poison concentration induced by this high pressure poison injection jet activated upon the reactor trip in a CANDU-6 reactor moderator tank has been developed and used to generate the poison concentration distribution of the poison curtains induced by the high pressure jets injected into the vacant region between the pressure tube banks. The poison injection rate through the jet holes drilled on the nozzle pipes is obtained by a 1-D transient hydrodynamic code called, ALITRIG, and this injection rate is used to provide the inlet boundary condition to a 3-D CFD model of the moderator tank based on CFX4.3, a commercial CFD code developed by AEA technology, to simulate the formation of the poison jet curtain inside the moderator tank. For the validation, a simulation for a generic CANDU-6 SDS2 design poison jet growth experiment was made to evaluate this model's capability against experiment. As no concentration field was measured and only the growth of the poison jet height was obtained by high speed camera, the validation was limited as such. The result showed that if one assume the jet front corresponds to 200 ppm of the poison the model succeed to

  1. Transfer of a cold atmospheric pressure plasma jet through a long flexible plastic tube

    Science.gov (United States)

    Kostov, Konstantin G.; Machida, Munemasa; Prysiazhnyi, Vadym; Honda, Roberto Y.

    2015-04-01

    This work proposes an experimental configuration for the generation of a cold atmospheric pressure plasma jet at the downstream end of a long flexible plastic tube. The device consists of a cylindrical dielectric chamber where an insulated metal rod that serves as high-voltage electrode is inserted. The chamber is connected to a long (up to 4 m) commercial flexible plastic tube, equipped with a thin floating Cu wire. The wire penetrates a few mm inside the discharge chamber, passes freely (with no special support) along the plastic tube and terminates a few millimeters before the tube end. The system is flushed with Ar and the dielectric barrier discharge (DBD) is ignited inside the dielectric chamber by a low frequency ac power supply. The gas flow is guided by the plastic tube while the metal wire, when in contact with the plasma inside the DBD reactor, acquires plasma potential. There is no discharge inside the plastic tube, however an Ar plasma jet can be extracted from the downstream tube end. The jet obtained by this method is cold enough to be put in direct contact with human skin without an electric shock. Therefore, by using this approach an Ar plasma jet can be generated at the tip of a long plastic tube far from the high-voltage discharge region, which provides the safe operation conditions and device flexibility required for medical treatment.

  2. Analysis of the pressure tube failure at Pickering NGS A unit 2

    International Nuclear Information System (INIS)

    About noon on the 1st August 1983, the pressure tube in fuel channel G16 of the Pickering NGS A unit 2 reactor developed a critical through-wall crack and failed by fast fracture after 342 days of continuous full power operation. Following removal of the fuel, a TV inspection inside the fuel channel revealed an axial crack in the bottom of the pressure tube approximately 2 metres long, in which two fuel pencils were lodged. After extracting the fuel pencils, the fuel channel was removed and shipped to the Atomic Energy of Canada Limited's Chalk River Nuclear Laboratories for detailed examination to determine the cause of failure. Examination of failures normally takes a course of looking at the fracture and gradually refining the work into finer detail to determine the actual origin of the failure. In this case, several other aspects also needed to be examined. The position of the garter spring was very important, as was examination of the calandria tube, which was subsequently removed. During the inspection several other fuel channels in Pickering A, Bruce A and NPD reactors were inspected and some removed for further assessment at CRNL. All these aspects came together to outline the cause and mechanism of failure. The following gives a very brief review of the salient features of the examination of Zircaloy 2 and zirconium-niobium pressure tubes and the implication for operation of subsequent reactors which have zirconium-niobium pressure tubes

  3. Boiling on a tube bundle: heat transfer, pressure drop and flow patterns

    International Nuclear Information System (INIS)

    The complexity of the two-phase flow in a tube bundle presents important problems in the design and understanding of the physical phenomena taking place. The working conditions of an evaporator depend largely on the dynamics of the two-phase flow that in turn influence the heat exchange and the pressure drop of the system. A characterization of the flow dynamics, and possibly the identification of the flow pattern in the tube bundle, is thus expected to lead to a better understanding of the phenomena and to reveal on the mechanisms governing the tube bundle. Therefore, the present study aims at providing further insights into two-phase bundle flow through a new visualization system able to provide for the first time a view of the flow in the core of a tube bundle. In addition, the measurement of the light attenuation of a laser beam through the two-phase flow and measurement of the high frequency pressure fluctuations with a piezo-electric pressure transducer are used to characterize the flow. The design and the validation of this new instrumentation also provided a method for the detection of dry-out in tube bundles. This was achieved by a laser attenuation technique, flow visualization, and estimation of the power spectrum of the pressure fluctuation. The current investigation includes results for two different refrigerants, R134a and R236fa, three saturations temperatures Tsat = 5, 10 and 15 °C, mass velocities ranging from 4 to 40 kg/sm² in adiabatic and diabatic conditions (several heat fluxes). Measurement of the local heat transfer coefficient and two-phase frictional pressure drop were obtained and utilized to improve the current prediction methods. The heat transfer and pressure drop data were supported by extensive characterization of the two-phase flow, which was to improve the understanding of the two-phase flow occurring in tube bundles. (author)

  4. Comparative evaluation of intraocular pressure changes subsequent to insertion of laryngeal mask airway and endotracheal tube.

    Directory of Open Access Journals (Sweden)

    Ghai B

    2001-07-01

    Full Text Available AIMS: To evaluate the intraocular pressure and haemodynamic changes subsequent to insertion of laryngeal mask airway and endotracheal tube. SUBJECTS AND METHODS: The study was conducted in 50 adult patients. A standard general anaesthesia was administered to all the patients. After 3 minutes of induction of anaesthesia baseline measurements of heart rate, non-invasive blood pressure and intraocular pressure were taken following which patients were divided into two groups: laryngeal mask airway was inserted in group 1 and tracheal tube in group 2. These measurements were repeated at 15-30 second, every minute thereafter up to 5 minutes after airway instrumentation. RESULTS: A statistically significant rise in heart rate, systolic blood pressure, diastolic blood pressure and intraocular pressure was seen in both the groups subsequent to insertion of laryngeal mask airway or endotracheal tube. Mean maximum increase was statistically more after endotracheal intubation than after laryngeal mask airway insertion. The duration of statistically significant pressure responses was also longer after endotracheal intubation. CONCLUSION: Laryngeal mask airway is an acceptable alternative technique for ocular surgeries, offering advantages in terms of intraocular pressure and cardiovascular stability compared to tracheal intubation.

  5. Development of Evaluation Technology of the Integrity of HWR Pressure Tubes

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Y. M.; Kim, Y. S.; Im, K. S.; Kim, K. S.; Ahn, S. B

    2007-06-15

    Zr-2.5Nb pressure tubes are one of the most critical structural components governing the lifetime of the heavy water reactors to carry fuel bundles and heavy coolant water inside. Since they are being degraded during their operation in reactors due to dimensional changes caused by creep and irradiation growth, neutron irradiation and delayed hydride cracking, it is required to evaluate their degradation by conducting material testing and examinations on the highly irradiated pressure tubes in hot cells and to keep tracking of their degradation behavior with operation time, which are the aim of this project.

  6. Evaluation and analysis of critical crack length of irradiated pressure tubes from Indian pressurised heavy water reactors

    International Nuclear Information System (INIS)

    Results of fracture toughness KJic were computed from transverse tensile properties of reactor operated pressure tubes, and axial critical crack length values derived from KJic are presented. Similarly fracture resistance curves derived from tensile properties of reactor operated pressure tubes and axial critical crack length values computed therefrom are presented. Under normal operating condition of the reactors the pressure tubes experience temperatures ranging from 250 deg C to 300 deg C. In occurrence of contact between pressure tube and calandria tube, the contact region may not be expected to have a mean through wall temperature below 200 deg C. The axial critical crack length of three reactor operated pressure tubes, therefore were evaluated in the temperature range 200 deg C to 300 deg C. The significance of the magnitude of the evaluated critical crack length is discussed. (author)

  7. Development of poison injection code-COPJET for high pressure liquid poison injection in pressure tube type heavy water reactor

    International Nuclear Information System (INIS)

    Shut Down System-2 (SDS-2) in advanced vertical pressure tube type reactor, provides rapid reactor shutdown by high pressure injection of a neutron absorbing liquid called poison, into the moderator in the calandria. Poison inside the calandria is distributed by poison jets issued from holes provided in the injection tubes. Effectiveness of the system depends on the rate and spread of the poison in the moderator. In this study, a transient one-dimensional (1-D) hydraulic code, COPJET is developed, to predict the performance of system by predicting poison jet length with time. Validation of the COPJET is done with the data available in literature. Thereafter, it is applied for poison jet length prediction of advanced vertical pressure type reactor. (author)

  8. Is Anesthesiologist’s experience important while inflating the endotracheal tube cuff with the right pressure?

    Directory of Open Access Journals (Sweden)

    Nesrin Turan

    2010-12-01

    Full Text Available Objectives: Cuff pressure in endotracheal tubes should be in the range of 26–30 cm H2O. In this study we aimed to examine whether anesthesiologist’s experience is important while inflating the endotracheal tube correctly after the intubation.Materials and methods: The patients who were included to the study were intubated after the induction of general anesthesia. The patients were divided into 4 groups according to the training year of the anesthesia research assistant resident inflating the endotracheal tube (ET cuff. Group I (n=64 the cuff pressure which were inflated by the first year residents; Group II (n=92 the cuff pressure which were inflated by the 2nd year residents; Group III (n=144 the cuff pressure which were inflated by the 3rd year residents; Group IV (n=93 the cuff pressure which were inflated by the 4th year residents were measured by manometer.Results: When we compared the cases in which the cuff pressure were between 26-30cm H2O we found that the best results were in Group II and respectively in Group III and IV and the worst results were in Group I. The difference between Group II and Group I were statistically significant (p<0.05.Conclusion: We believe that manometer should be used ET for cuff pressure setting and monitoring. J Clin Exp Invest 2010; 1(3: 195-198

  9. Tracheal tube and laryngeal mask cuff pressure during anaesthesia - mandatory monitoring is in need

    DEFF Research Database (Denmark)

    Rokamp, K.Z.; Secher, N.H.; Møller, Ann;

    2010-01-01

    patients. In the 82 patients provided with a laryngeal mask, the cuff pressure was 95 (10 - 121) cm H2O and above 60 cm H2O (upper recommended level) for 56 patients and in 34 of these patients, the pressure exceeded the upper cuff gauge limit (120 cm H2O). There was no association between cuff pressure......ABSTRACT: BACKGROUND: To prevent endothelium and nerve lesions, tracheal tube and laryngeal mask cuff pressure is to be maintained at a low level and yet be high enough to secure air sealing. METHOD: In a prospective quality-control study, 201 patients undergoing surgery during anaesthesia (without...... the use of nitrous oxide) were included for determination of the cuff pressure of the tracheal tubes and laryngeal masks. RESULTS: In the 119 patients provided with a tracheal tube, the median cuff pressure was 30 (range 8 - 100) cm H2O and the pressure exceeded 30 cm H2O (upper recommended level) for 54...

  10. Tracheal tube and laryngeal mask cuff pressure during anaesthesia - mandatory monitoring is in need

    DEFF Research Database (Denmark)

    Rokamp, K.Z.; Secher, N.H.; Møller, Ann;

    2010-01-01

    ABSTRACT: BACKGROUND: To prevent endothelium and nerve lesions, tracheal tube and laryngeal mask cuff pressure is to be maintained at a low level and yet be high enough to secure air sealing. METHOD: In a prospective quality-control study, 201 patients undergoing surgery during anaesthesia (without...... the use of nitrous oxide) were included for determination of the cuff pressure of the tracheal tubes and laryngeal masks. RESULTS: In the 119 patients provided with a tracheal tube, the median cuff pressure was 30 (range 8 - 100) cm H2O and the pressure exceeded 30 cm H2O (upper recommended level) for 54...... patients. In the 82 patients provided with a laryngeal mask, the cuff pressure was 95 (10 - 121) cm H2O and above 60 cm H2O (upper recommended level) for 56 patients and in 34 of these patients, the pressure exceeded the upper cuff gauge limit (120 cm H2O). There was no association between cuff pressure...

  11. Candu advanced fuel cycles: key to energy sustainability

    International Nuclear Information System (INIS)

    reactor designs, allowing operation today on currently available fuels and switching to other fuelling options as market conditions change. This establishes an important freedom from future resource constraints without depending on future commercialization of challenging and expensive technologies such as fast breeder reactors, yet, once these are commercially available, Candu and fast breeder fuel cycles are complementary and can achieve a highly advantageous synergism. This paper examines the fuel cycle options which Candu reactor technology can accommodate, including the use of slightly enriched uranium, direct use of spent pressurized water reactor fuel in Candu (dupic), burning recovered uranium, mixed plutonium and uranium oxides or actinides and the use of thorium based fuel cycles. These options provide Candu reactors with the most flexible fuelling of any reactor type, which are readily adaptable to meeting future variations in energy markets, regardless of what these may be

  12. Creep properties of electric resistance welded boiler tubes under internal pressure

    International Nuclear Information System (INIS)

    Creep rupture tests on electric resistance welded (ERW) tubular specimens of carbon steel and 1% Cr-0.5% Mo steel and burst tests on thickness-deviated tubular specimens of carbon steel are described. Also, changes of structures and mechanical properties of 1% Cr-0.5% Mo steel tubes after exposure to 5500C for up to 10,000 hours under a tensile hoop stress of 108 MPa are described. The creep rupture properties of ERW boiler tubes were proved to be quite comparable to those of seamless tubes, and the slightest deviation in wall thickness was shown to affect the internal pressure rupture behavior. Changes of structures at welded portion of ERW 1% Cr-0.5% Mo steel tubes were as same as those of base metal

  13. Low-frequency pressure wave propagation in liquid-filled, flexible tubes. (A)

    DEFF Research Database (Denmark)

    Bjørnø, Leif; Bjelland, C.

    1992-01-01

    . The complex, frequency-dependent moduli of relevant tube materials have been measured in a series of experiments using three different experimental procedures, and the data obtained are compared. The three procedures were: (1) ultrasonic wave propagation, (2) transversal resonance in bar samples, and (3......A model has been developed for propagation of low-frequency pressure waves in viscoelastic tubes with distensibility of greater importance than compressibility of the liquid. The dispersion and attenuation are shown to be strongly dependent on the viscoelastic properties of the tube wall......) moduli determined by stress wave transfer function measurements in simple extension experiments. The moduli are used in the model to produce realistic dispersion relations and frequency dependent attenuation. Signal transfer functions between positions in the liquid-filled tube can be synthesized from...

  14. Francis turbine draft tube modelling for prediction of pressure fluctuations on prototype

    Science.gov (United States)

    Alligné, S.; Landry, C.; Favrel, A.; Nicolet, C.; Avellan, F.

    2015-12-01

    The prediction of pressure fluctuations amplitudes on Francis turbine prototype is a challenge for hydro-equipment industry since it is subjected to guarantees to ensure smooth and reliable operation of the hydro units. The European FP7 research project Hyperbole aims to setup a methodology to transpose the pressure fluctuations measured on the reduced scale model to the prototype generating units. This paper presents this methodology which relies on an advanced modelling of the draft tube cavitation flow, and focuses on the transposition to the prototype of the draft tube model parameters identified on the reduced scale model. Different modelling assumptions of the draft tube are considered and their influence on the eigenmodes and the forced response of the system are presented.

  15. Thorium utilization in Candu reactors

    International Nuclear Information System (INIS)

    In this study, means of thorium utilization in a CANDU reactor are considered. A once through thorium-DUPIC cycle is analyzed in detail. CANDU has the best neutron economy among the commercially available power reactors, which makes it suitable for many different fuel cycle options. A review of the available fuel cycles is also done in the scope of this study to select an economically viable cycle which does not impose profound changes in the neutronic properties of the core that require remodeling of core and related systems. To create a good model ot the CANDU core for the necessary calculations, the steady state properties of CANDU reactor are analyzed. It is assumed that approximation ot refueling as moving the bundles at a constant velocity is valid. This approximation leads to a corollary; The average cross sections of two adjacent bidirectionally refueled channels are independent of axial location. This is also veritied. A result of this corollary the CANDU core can be modeled only in radial direction in cylindirical geometry. The steady state CANDU core model is prepared using the actual power values and these values are sought in the results. The control systems which effect the neutron flux shape are introduced into the model later in the form of additional absorption cross section and lower diffusion coefficient. The results are in good agreement with the actual values. Several different thorium-DUPIC fuel bundle configurations are considered and the one with 12 Th02 elements in the third ring is found to have similar burnup dependent cross-sections and location infinite multiplication factors. Using the model created, the bundle is tested also in the tull core model and it is tound that this bundle configuration complies with the current refueling scheme. That is, no changes are necessary in the refuelind rate or the control systems. A higher conversion ratio of 0.82 is attained, while the excess reactivity of the core is found to decrease by 0.01 Ak

  16. The small (or large) modular CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Meneley, D.; Harvel, G. [Univ. of Ontario Inst. of Tech., Oshawa, Ontario (Canada)

    2013-07-01

    This presentation outlines the design for small (or large) modular CANDU. The origins of this work go back many years to a comment by John Foster, then President of AECL CANDU. Foster noted that the CANDU reactor, with its many small fuel channels, was like a wood campfire. To make a bigger fire, just throw on some more logs (channels). If you want a smaller fire, just use fewer logs. The design process is greatly simplified.

  17. Evaluated Plan Stress Of Weld In Pressure Tube Using X Ray Diffraction Technique

    International Nuclear Information System (INIS)

    X ray diffraction is a fundamental technique measuring stress, this technique has determined crystal strain in materials, from that determined stress in materials. This paper presents study of evaluating plane stress of weld in pressure tube, using modern XRD apparatus: X Pert Pro. (author)

  18. Pre and post garter spring repositioning ultrasonic inspection of pressure tubes

    International Nuclear Information System (INIS)

    This paper present a description of the ultrasonic cracked hydride blister detections system used for pre and post inspection of pressure tubes during garter spring repositioning in CNE (Embalse Nuclear Power Station). Ultrasonic system setup configuration, transducers characteristics, blister detection head, calibration of parameters, operating procedure, records of ultrasonic inspections and evaluation. (author)

  19. Irradiation Effect on the Mechanical Property for Wolsong 1 Pressure Tube

    International Nuclear Information System (INIS)

    Summary: Need for accurate prediction model for PT dimensional change - One of key data to determine power decrease and lifetime; • Limit for PT measured data - Necessity for collaboration of HWR countries; • Irradiation Effect on PT Mechanical Property - Basic data of pressure tube behavior for the irradiation - More research on the microstructure change for the irradiation

  20. Intercomparison of techniques for inspection and diagnostics of heavy water reactor pressure tubes. Additional information

    International Nuclear Information System (INIS)

    The reports from Argentina, Canada, India, Korea and Romania are presented concerning the projects carried out under the Coordinated Research Program (CRP) I3.30.10 of the International Agency for Atomic Energy - Vienna related to 'Intercomparison of Techniques for Pressure Tube Inspection and Diagnostics'

  1. Systems analysis of the CANDU 3 Reactor

    International Nuclear Information System (INIS)

    This report presents the results of a systems failure analysis study of the CANDU 3 reactor design; the study was performed for the US Nuclear Regulatory Commission. As part of the study a review of the CANDU 3 design documentation was performed, a plant assessment methodology was developed, representative plant initiating events were identified for detailed analysis, and a plant assessment was performed. The results of the plant assessment included classification of the CANDU 3 event sequences that were analyzed, determination of CANDU 3 systems that are ''significant to safety,'' and identification of key operator actions for the analyzed events

  2. Numerical model for thermal and mechanical behaviour of a CANDU 37-element bundle

    International Nuclear Information System (INIS)

    Prediction of transient fuel bundle deformations is important for assessing the integrity of fuel and the surrounding structural components under different operating conditions including accidents. For numerical simulation of the interactions between fuel bundle and pressure tube, a reliable numerical bundle model is required to predict thermal and mechanical behaviour of the fuel bundle assembly under different thermal loading conditions. To ensure realistic representations of the bundle behaviour, this model must include all of the important thermal and mechanical features of the fuel bundle, such as temperature-dependent material properties, thermal viscoplastic deformation in sheath, fuel-to-sheath interactions, endplate constraints and contacts between fuel elements. In this paper, we present a finite element based numerical model for predicting macroscopic transient thermal-mechanical behaviour of a complete 37-element CANDU nuclear fuel bundle under accident conditions and demonstrate its potential for being used to investigate fuel bundle to pressure tube interaction in future nuclear safety analyses. This bundle model has been validated against available experimental and numerical solutions and applied to various simulations involving steady-state and transient loading conditions. (author)

  3. Two-Phase Critical Discharge of Initially Saturated or Subcooled Water Flowing in Sharp-Edgred Tubes at High Pressure

    Institute of Scientific and Technical Information of China (English)

    1995-01-01

    The transient critical flow experiment with sharp-deged tubes as the break geometries is conducted in high pressure convective circulation test loop of Xi'an Jiantong University.The initial Steady operation pressure is up to 22.0MPa.An empirical correlation was made to obtain the critical mass flow rates,the critical pressure ratio and the thermal nonequilibrium number were correlated as the functions of the tube length to tube diameter ratio L/D.The predicted critical mass flow rate gets a higher accureacy for short tubes with L/D 12.

  4. Thermal-hydraulic instabilities in pressure tube graphite - moderated boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tsiklauri, G.; Schmitt, B.

    1995-09-01

    Thermally induced two-phase instabilities in non-uniformly heated boiling channels in RBMK-1000 reactor have been analyzed using RELAP5/MOD3 code. The RELAP5 model of a RBMK-1000 reactor was developed to investigate low flow in a distribution group header (DGH) supplying 44 fuel pressure tubes. The model was evaluated against experimental data. The results of the calculations indicate that the period of oscillation for the high power tube varied from 3.1s to 2.6s, over the power range of 2.0 MW to 3.0 MW, respectively. The amplitude of the flow oscillation for the high powered tube varied from +100% to -150% of the tube average flow. Reverse flow did not occur in the lower power tubes. The amplitude of oscillation in the subcooled region at the inlet to the fuel region is higher than in the saturated region at the outlet. In the upper fuel region and outlet connectors the flow oscillations are dissipated. The threshold of flow instability for the high powered tubes of a RBMK reactor is compared to Japanese data and appears to be in good agreement.

  5. Spatial resolution of thin-walled high-pressure drift tubes

    CERN Document Server

    Davkov, V I; Tikhomirov, V O; Smirnov, S Y; Gregor, I; Senger, P; Naumann, L; Myalkovskiy, V V; Mouraviev, S V; Peshekhonov, V D; Russakovich, N A; Rufanov, I A; Rembser, C

    2011-01-01

    A small prototype detector based on high pressure thin-walled tubes (straws) has been developed and its parameters have been studied on a bench at JINR, Dubna, and SPS at CERN. The inner diameter of the straws is 9.53 mm. The pressure of the active gas mixture Ar/CO(2) (80/20) was varied from 1 to 5 bar. The best spatial resolution achieved in this pressure range is similar to 40 mu m. Both the high efficiency and high rate capability are retained. (C) 2011 Published by Elsevier B.V.

  6. Boiling on a tube bundle: heat transfer, pressure drop and flow patterns

    International Nuclear Information System (INIS)

    The complexity of two-phase flow boiling on a tube bundle presents many challenges to the understanding of the physical phenomena taking place. It is important to quantify these numerous heat flow mechanisms in order to better describe the performance of tube bundles as a function of the operational conditions. In the present study, the bundle boiling facility at the Laboratory of Heat and Mass Transfer (LTCM) was modified to obtain high-speed videos to characterise the two-phase regimes and some bubble dynamics of the boiling process. It was then used to measure heat transfer on single tubes and in bundle boiling conditions. Pressure drop measurements were also made during adiabatic and diabatic bundle conditions. New enhanced boiling tubes from Wolverine Tube Inc. (Turbo-B5) and the Wieland-Werke AG (Gewa-B5) were investigated using R134a and R236fa as test fluids. The tests were carried out at saturation temperatures Tsat of 5 °C and 15 °C, mass flow rates from 4 to 35 kg/m2s and heat fluxes from 15 to 70 kW/m2, typical of actual operating conditions. The flow pattern investigation was conducted using visual observations from a borescope inserted in the middle of the bundle. Measurements of the light attenuation of a laser beam through the intertube two-phase flow and local pressure fluctuations with piezo-electric pressure transducers were also taken to further help in characterising the complex flow. Pressure drop measurements and data reduction procedures were revised and used to develop new, improved frictional pressure drop prediction methods for adiabatic and diabatic two-phase conditions. The physical phenomena governing the enhanced tube evaporation process and their effects on the performance of tube bundles were investigated and insight gained. A new method based on a theoretical analysis of thin film evaporation was used to propose a new correlating parameter. A large new database of local heat transfer coefficients were obtained and then utilised

  7. Optimization of thorium-uranium content in a 54-element fuel bundle for use in a CANDU-SCWR

    International Nuclear Information System (INIS)

    A new 54-element fuel bundle design has been proposed for use in a pressure-tube supercritical water-cooled reactor, a pre-conceptual evolution of existing CANDU reactors. Pursuant to the goals of the Generation IV International Forum regarding advancement in nuclear fuel cycles, optimization of the thorium and uranium content in each ring of fuel elements has been studied with the objectives of maximizing the achievable fuel utilization (burnup) and total thorium content within the bundle, while simultaneously minimizing the linear element ratings and coolant void reactivity. The bundle was modeled within a reactor lattice cell using WIMS-AECL, and the uranium and thorium content in each ring of fuel elements was optimized using a weighted merit function of the aforementioned criteria and a metaheuristic search algorithm. (author)

  8. Characterization of magnetically impelled arc butt welded T11 tubes for high pressure applications

    Directory of Open Access Journals (Sweden)

    R. Sivasankari

    2015-09-01

    Full Text Available Magnetically impelled arc butt (MIAB welding is a pressure welding process used for joining of pipes and tubes with an external magnetic field affecting arc rotation along the tube circumference. In this work, MIAB welding of low alloy steel (T11 tubes were carried out to study the microstructural changes occurring in thermo-mechanically affected zone (TMAZ. To qualify the process for the welding applications where pressure could be up to 300 bar, the MIAB welds are studied with variations of arc current and arc rotation time. It is found that TMAZ shows higher hardness than that in base metal and displays higher weld tensile strength and ductility due to bainitic transformation. The effect of arc current on the weld interface is also detailed and is found to be defect free at higher values of arc currents. The results reveal that MIAB welded samples exhibits good structural property correlation for high pressure applications with an added benefit of enhanced productivity at lower cost. The study will enable the use of MIAB welding for high pressure applications in power and defence sectors.

  9. Effect of Ovality on Maximum External Pressure of Helically Coiled Steam Generator Tubes with a Rectangular Wear

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Dong In; Lim, Eun Mo; Huh, Nam Su [Seoul National Univ. of Science and Technology, Seoul (Korea, Republic of); Choi, Shin Beom; Yu, Je Yong; Kim, Ji Ho; Choi, Suhn [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    A structural integrity of steam generator tubes of nuclear power plants is one of crucial parameters for safe operation of nuclear power plants. Thus, many studies have been made to provide engineering methods to assess integrity of defective tubes of commercial nuclear power plants considering its operating environments and defect characteristics. As described above, the geometric and operating conditions of steam generator tubes in integral reactor are significantly different from those of commercial reactor. Therefore, the structural integrity assessment of defective tubes of integral reactor taking into account its own operating conditions and geometric characteristics, i. e., external pressure and helically coiled shape, should be made to demonstrate compliance with the current design criteria. Also, ovality is very specific characteristics of the helically coiled tube because it is occurred during the coiling processes. The wear, occurring from FIV (Flow Induced Vibration) and so on, is main degradation of steam generator tube. In the present study, maximum external pressure of helically coiled steam generator tube with wear is predicted based on the detailed 3-dimensional finite element analysis. As for shape of wear defect, the rectangular shape is considered. In particular, the effect of ovality on the maximum external pressure of helically coiled tubes with rectangular shaped wear is investigated. In the present work, the maximum external pressure of helically coiled steam generator tube with rectangular shaped wear is investigated via detailed 3-D FE analyses. In order to cover a practical range of geometries for defective tube, the variables affecting the maximum external pressure were systematically varied. In particular, the effect of tube ovality on the maximum external pressure is evaluated. It is expected that the present results can be used as a technical backgrounds for establishing a practical structural integrity assessment guideline of

  10. Endotracheal tube cuff pressure before, during, and after fixed-wing air medical retrieval.

    Science.gov (United States)

    Brendt, Peter; Schnekenburger, Marc; Paxton, Karen; Brown, Anthony; Mendis, Kumara

    2013-01-01

    Abstract Background. Increased endotracheal tube (ETT) cuff pressure is associated with compromised tracheal mucosal perfusion and injuries. No published data are available for Australia on pressures in the fixed-wing air medical retrieval setting. Objective. After introduction of a cuff pressure manometer (Mallinckrodt, Hennef, Germany) at the Royal Flying Doctor Service (RFDS) Base in Dubbo, New South Wales (NSW), Australia, we assessed the prevalence of increased cuff pressures before, during, and after air medical retrieval. Methods. This was a retrospective audit in 35 ventilated patients during fixed-wing retrievals by the RFDS in NSW, Australia. Explicit chart review of ventilated patients was performed for cuff pressures and changes during medical retrievals with pressurized aircrafts. Pearson correlation was calculated to determine the relation of ascent and ETT cuff pressure change from ground to flight level. Results. The mean (± standard deviation) of the first ETT cuff pressure measurement on the ground was 44 ± 20 cmH2O. Prior to retrieval in 11 patients, the ETT cuff pressure was >30 cmH2O and in 11 patients >50 cmH2O. After ascent to cruising altitude, the cuff pressure was >30 cmH2O in 22 patients and >50 cmH2O in eight patients. The cuff pressure was reduced 1) in 72% of cases prior to take off and 2) in 85% of cases during flight, and 3) after landing, the cuff pressure increased in 85% of cases. The correlation between ascent in cabin altitude and ETT cuff pressure was r = 0.3901, p = 0.0205. Conclusions. The high prevalence of excessive cuff pressures during air medical retrieval can be avoided by the use of cuff pressure manometers. Key words: cuff pressure; air medical retrieval; prehospital. PMID:23252881

  11. Pressure drop across a tube-bundle flow rectifier. Sekiso koshi no teiko keisu ni kansuru kenkyu

    Energy Technology Data Exchange (ETDEWEB)

    Hayashi, T.; Yoshino, F. (Tottori Universtiy, Tottori (Japan). Faculty of Engineering)

    1991-01-25

    The pressure drop coefficient of a tube-bundle flow rectifier made by piling tubes in parallel in the flow direction was obtained with experiments. A tube-bundle rectifier forms optional pressure drop distribution, has rectification effects, and is fit for the formation of the field of the homogeneous flow with large velocity gradient. Tubes may be piled in a staggered or a side-by side arrangement. As a duct, an acrylic pipe of 130 outer diameter, 119 inner diameter and 8,069 mm length was used. A tube-bundle rectifier is a cartridge of integral construction in which polypropylene straws of 6.53 inner diameter and 0.19 mm wall thickness are piled. The ratio of the flow passage sectional area against the whole sectional area is 9.3% in the staggered arrangement and 21.5% in the side-by side arrangement. While the opening ratio is small in the staggered arrangement, the pressure drop coefficient is not necessarily large in this arrangement; the coefficient depends on the tube length and the Reynolds number. In some cases, on the contrary, the pressure drop coefficient is larger in the side-by-side arrangement. It was also indicated that the approximation of the wire mesh equivalent pressure drop coefficient in the extremity, where the tube wall is as thick as the length of the tube-bundle rectifier, can be obtained with the pressure drop coefficient of plain weave wire mesh. 11 refs., 6 figs.

  12. Reconstruction of an acoustic pressure field in a resonance tube by particle image velocimetry.

    Science.gov (United States)

    Kuzuu, K; Hasegawa, S

    2015-11-01

    A technique for estimating an acoustic field in a resonance tube is suggested. The estimation of an acoustic field in a resonance tube is important for the development of the thermoacoustic engine, and can be conducted employing two sensors to measure pressure. While this measurement technique is known as the two-sensor method, care needs to be taken with the location of pressure sensors when conducting pressure measurements. In the present study, particle image velocimetry (PIV) is employed instead of a pressure measurement by a sensor, and two-dimensional velocity vector images are extracted as sequential data from only a one- time recording made by a video camera of PIV. The spatial velocity amplitude is obtained from those images, and a pressure distribution is calculated from velocity amplitudes at two points by extending the equations derived for the two-sensor method. By means of this method, problems relating to the locations and calibrations of multiple pressure sensors are avoided. Furthermore, to verify the accuracy of the present method, the experiments are conducted employing the conventional two-sensor method and laser Doppler velocimetry (LDV). Then, results by the proposed method are compared with those obtained with the two-sensor method and LDV.

  13. Experimental investigation of pressure fluctuations caused by a vortex rope in a draft tube

    Science.gov (United States)

    Kirschner, O.; Ruprecht, A.; Göde, E.; Riedelbauch, S.

    2012-11-01

    In the last years hydro power plants have taken the task of power-frequency control for the electrical grid. Therefore turbines in storage hydro power plants often operate outside their optimum. If Francis-turbines and pump-turbines operate at off-design conditions, a vortex rope in the draft tube can develop. The vortex rope can cause pressure oscillations. In addition to low frequencies caused by the rotation of the vortex rope and the harmonics of these frequencies, pressure fluctuations with higher frequencies can be observed in some operating points too. In this experimental investigation the flow structure and behavior of the vortex rope movement in the draft tube of a model pump-turbine are analyzed. The investigation focuses on the correlation of the pressure fluctuation frequency measured at the draft tube wall with the movement of the vortex rope. The movement of the vortex rope is analyzed by the velocity field in the draft tube which was measured with particle image velocimetry. Additionally, the vortex rope movement has been analyzed with the captures of high-speed-movies from the cavitating vortex rope. Besides the rotation of the vortex rope due to pressure fluctuation with low frequencies the results of the measurement also show a correlation between the rotation of the elliptical or deformed rope cross-section and the higher frequency pressure pulsation. An approximation shows that the frequencies of the pressure fluctuation and the movement of the vortex rope are also connected with the velocity of the flow. Taking into account the size and position of the cavitating vortex core as well as the velocity at the position of the surface of the cavitating vortex core the time-period of the rotation of the vortex core can be approximated. The results show that both, the low frequency pressure fluctuation and the higher frequency pressure fluctuation are correlating with the vortex rope movement. With this estimation, the period of the higher frequency

  14. Experimental Study on Heat Transfer and Pressure Drop of Micro-Sized Tube Heat Exchanger

    Institute of Scientific and Technical Information of China (English)

    王秋香; 戴传山

    2014-01-01

    A micro-sized tube heat exchanger (MTHE) was fabricated, and its performance in heat transfer and pres-sure drop was experimentally studied. The single-phase forced convection heat transfer correlation on the sides of the MTHE tubes was proposed and compared with previous experimental data in the Reynolds number range of 500-1 800. The average deviation of the correlation in calculating the Nusselt number was about 6.59%. The entrance effect in the thermal entrance region was discussed. In the same range of Reynolds number, the pressure drop and friction coefficient were found to be considerably higher than those predicted by the conventional correlations. The product of friction factor and Reynolds number was also a constant, but much higher than the conventional.

  15. Stress and integrity analysis of steam superheater tubes of a high pressure boiler

    Energy Technology Data Exchange (ETDEWEB)

    Neves, Daniel Leite Cypriano; Seixas, Jansen Renato de Carvalho [PETROBRAS, Duque de Caxias, RJ (Brazil). Refinaria Duque de Caxias (REDUC). Mantencao Industrial]. E-mail: dcypriano@petrobras.com.br; Tinoco, Ediberto Bastos [PETROBRAS, Rio de Janeiro, RJ (Brazil). Centro de Pesquisas (CENPES). Engenharia de Equipamento Basico; Rocha, Adriana da Cunha; Abud, Ibrahim de Cerqueira [Instituto Nacional de Tecnologia (INT), Rio de Janeiro, RJ (Brazil). Lab. de Metalografia e de Dureza

    2004-03-01

    Sources that can lead to deterioration of steam superheater tubes of a high pressure boiler were studied by a stress analysis, focused on internal pressure and temperature experienced by the material at real operating conditions. Loss of flame control, internal deposits and unexpected peak charge are factors that generate loads above the design limit of tube materials, which can be subjected to strain, buckling, cracks and finally rupture in service. To evaluate integrity and dependability of these components, the microstructure of selected samples along the superheater was studied by optical microscopy. Associated with this analysis, dimensional inspection, nondestructive testing, hardness measurement and deposit examination were made to determine the resultant material condition after twenty three years of operation. (author)

  16. Stress and integrity analysis of steam superheater tubes of a high pressure boiler

    Directory of Open Access Journals (Sweden)

    Neves Daniel Leite Cypriano

    2004-01-01

    Full Text Available Sources that can lead to deterioration of steam superheater tubes of a high pressure boiler were studied by a stress analysis, focused on internal pressure and temperature experienced by the material at real operating conditions. Loss of flame control, internal deposits and unexpected peak charge are factors that generate loads above the design limit of tube materials, which can be subjected to strain, buckling, cracks and finally rupture in service. To evaluate integrity and dependability of these components, the microstructure of selected samples along the superheater was studied by optical microscopy. Associated with this analysis, dimensional inspection, nondestructive testing, hardness measurement and deposit examination were made to determine the resultant material condition after twenty three years of operation.

  17. Exergoeconomic optimization of coaxial tube evaporators for cooling of high pressure gaseous hydrogen during vehicle fuelling

    DEFF Research Database (Denmark)

    Jensen, Jonas Kjær; Rothuizen, Erasmus Damgaard; Markussen, Wiebke Brix

    2014-01-01

    Gaseous hydrogen as an automotive fuel is reaching the point of commercial introduction. Development of hydrogen fuelling stations considering an acceptable fuelling time by cooling the hydrogen to -40 C has started. This paper presents a design study of coaxial tube ammonia evaporators for three...... different concepts of hydrogen cooling, one onestage and two two-stage processes. An exergoeconomic optimization is imposed to all three concepts to minimize the total cost. A numerical heat transfer model is developed in Engineer Equation Solver, using heat transfer and pressure drop correlations from the...... open literature. With this model the optimal choice of tube sizes and circuit numbers are found for all three concepts. The results show that cooling with a two-stage evaporator after the pressure eduction valve yields the lowest total cost, 45 % lower than the highest, which is with a one...

  18. Break flow modeling for a steam generator tube rupture (SGTR) incident in a pressurized water reactor (PWR)

    International Nuclear Information System (INIS)

    The design-basis steam generator tube rupture (SGTR) scenario for the pressurized water reactor (PWR) postulates an instantaneous double-ended break of a steam generator (SG) U-tube. The flow rate through the broken U-tube depends on the primary-to-secondary side differential pressure in the affected SG, the primary coolant subcooling, and the break location along the U-tube. In this report, the RELAP5/MOD2 code's capability in predicting the SGTR break flow rate is assessed against experiments conducted on the Large Scale Test Facility (LSTF). The code is then used to predict break flow rate in the PWR for typical SGTR situations. It is shown that the code simulates well the break flow rates in the LSTF experiments for both single-phase and two-phase discharges, including two-phase critical flow discharge. The calculated PWR break flow rate takes a maximum for a break occurring at the lower end of the U-tube, on its cold leg side, because of the combined influence of tube-inlet fluid subcooling and frictional pressure drop along the broken tube. Modeling the tube frictional pressure drop is important to predict the break flow rate dependence on inlet fluid sub-cooling; simplified break flow modeling which applies a constant discharge coefficient less than unity, instead of modeling explicitly the tube frictional length, fails to predict the change in break flow rate accurately if the inlet subcooling varies for a wide range. (author)

  19. Detecting Nonlinearity in Pressure Data Inside the Draft Tube of a Real Francis Turbine

    OpenAIRE

    Sello, S.

    1995-01-01

    A general method for testing nonlinearity in time series is described and applied to measurements of different pressure data inside the draft tube surge of a real Francis turbine. Comparing the current original time series to an ensemble of surrogates time series, suitably constructed to mimic the linear properties of the original one, we was able to distinguish a linear stochastic from a nonlinear deterministic behaviour and, moreover, to quantify the degree of nonlinearity present in the re...

  20. Exact Solution of a Cylinder Tube Made of Metallic Foam Under Inner Pressure

    Institute of Scientific and Technical Information of China (English)

    ZHU Ai-yu; FAN Tian-you

    2008-01-01

    Exact solution of the stress and velocity fields of a cylinder tube of metallic foams under inner Pressure is given in which the Triantafillou and Gibson constitutive law(TG model)for the material is taken as a basis of the calculation.The nonlinear equation is turned linear equation by introducing a kinematics parameter.The differences between the full condensed materials and the effect of the relative densitv are also discussed.

  1. The formation and characteristics of hydride blisters in c.w. Zircaloy-2 pressure tubes

    International Nuclear Information System (INIS)

    Under the auspices of the IAEA, a consultants' meeting was arranged in Vienna, 1994 July 25-29, at which a Canadian delegation, consisting of AECL and Ontario Hydro Technologies personnel, presented information on their knowledge of the behaviour of hydride blisters in Zircaloy-2 pressure tubes. This document contains the 10 papers presented by the Canadian delegation to the meeting. It is believed that they represent a good reference document on hydride blister phenomena

  2. Some characteristics of the digitization pulses from high pressure neon-helium flash tubes

    International Nuclear Information System (INIS)

    Characteristics of the digitization output pulses from high pressure neon-helium flash tubes were studied under various operation conditions using square ultra-high voltage pulses. Properties reported by previous workers were compared. Two discharge mechanisms, the Townsend avalanche discharge and the streamer discharge, were observed to occur in sequence in some events. The output waveforms for both discharge mechanisms were studied in detail. The charge induced on a detecting probe was also estimated from the measured data. (Auth.)

  3. Very High Pressure Single Pulse Shock Tube Studies of Aromatic Species

    Energy Technology Data Exchange (ETDEWEB)

    Brezinsky, K.

    2006-11-28

    The principal focus of this research program is aimed at understanding the oxidation and pyrolysis chemistry of primary aromatic molecules and radicals with the goal of developing a comprehensive kinetic model at conditions that are relevant to practical combustion devices. A very high pressure single pulse shock tube is used to obtain experimental data over a wide pressure range in the high pressure regime, 5-1000 bars, at pre-flame temperatures for fuel pyrolysis and oxidation over a broad spectrum of equivalence ratios. Stable species sampled from the shock tube are analyzed using standard chromatographic techniques using GC/MS-PDD and GC/TCD-FID. Experimental data from the HPST (stable species profiles) and data from other laboratories (if available) are simulated using kinetic models (if available) to develop a comprehensive model that can describe aromatics oxidation and pyrolysis over a wide range of experimental conditions. The shock tube has been heated (1000C) recently to minimize effects due to condensation of aromatic, polycyclic and other heavy species. Work during this grant period has focused on 7 main areas summarized in the final technical report.

  4. Draft tube pressure pulsation predictions in Francis turbines with transient Computational Fluid Dynamics methodology

    Science.gov (United States)

    Melot, M.; Nennemann, B.; Désy, N.

    2014-03-01

    An automatic Computational Fluid Dynamics (CFD) procedure that aims at predicting Draft Tube Pressure Pulsations (DTPP) at part load is presented. After a brief review of the physics involved, a description of the transient numerical setup is given. Next, the paper describes a post processing technique, namely the separation of pressure signals into synchronous, asynchronous and random pulsations. Combining the CFD calculation with the post-processing technique allows the quantification of the potential excitation of the mechanical system during the design phase. Consequently it provides the hydraulic designer with a tool to specifically target DTPP and thus helps in the development of more robust designs for part load operation of turbines.

  5. CANDU reactors. Experience and innovation

    International Nuclear Information System (INIS)

    The title of this paper highlights two key considerations which must be properly balanced through good management in the evolution of any engineering product. Excessive reliance on experience will lead to product stagnation; excessive reliance on innovation will often lead to an unsatisfactory product, at least in the first generation of this product. To illustrate this balancing process, the paper reviews CANDU evolution and experience and the balance between proveness and innovation achieved through management of the evolution process from early prototypes to today's large-scale commercial units. A forecast of continuing evolutionary directions is included

  6. Tracheal Rupture due to Diffusion of Nitrous Oxide to Cuff of High-Volume, Low-Pressure Intubation Tube

    OpenAIRE

    Atalay, Canan; AYKAN, Şeyda; CAN, Abdullah; Doğan, Nazım

    2009-01-01

    Tracheal rupture is a rare complication of endotracheal intubation. Risk factors include short neck, repeated attempts due to failed intubation, inappropriate stylus, over-inflation of the cuff, poor positioning of the tube, inappropriate tube size, weakened membrane structure due to steroid use, chronic obstructive pulmonary disease, tracheomalacia, kyphosis, and use of nitric oxide during the operation. In this article, we suggest that high-volume, low-pressure tubes may not always provide ...

  7. Experimental study of heat transfer and pressure drop characteristics on shell-side of pin-fin tube oil cooler

    International Nuclear Information System (INIS)

    The comparative experimental study for one smooth tube oil cooler and three pin-fin tube oil coolers was performed by using lubricating oil as heat transfer medium. The experimental results indicate that in the range of experimental study, total heat transfer coefficient of pin-fin tube oil coolers is about 1.4-2 times higher than that of the smooth tube oil cooler. The heat transfer and pressure drop characteristics are greatly different for different structures of pin-fin tube oil coolers. The effects of the structure of pin-fin tube and shell-side flow path number are dominant to influence heat transfer and pressure drop characteristics of oil coolers. In the range of experimental study, large pin-fin height is conducive to the oil flow disturbance, but not conducive to the heat transfer on the tube-base heat transfer surface of pin-fin tube; single-pass pin-fin tube oil cooler offers high total heat transfer coefficient and volumetric heat transfer capacity, the global heat transfer performance and the friction characteristics are better than that of two-pass pin-fin tube oil cooler. (authors)

  8. LBB in Candu plants

    Energy Technology Data Exchange (ETDEWEB)

    Kozluk, M.J.; Vijay, D.K. [Ontario Hydro Nuclear, Toronto, Ontario (Canada)

    1997-04-01

    Postulated catastrophic rupture of high-energy piping systems is the fundamental criterion used for the safety design basis of both light and heavy water nuclear generating stations. Historically, the criterion has been applied by assuming a nonmechanistic instantaneous double-ended guillotine rupture of the largest diameter pipes inside of containment. Nonmechanistic, meaning that the assumption of an instantaneous guillotine rupture has not been based on stresses in the pipe, failure mechanisms, toughness of the piping material, nor the dynamics of the ruptured pipe ends as they separate. This postulated instantaneous double-ended guillotine rupture of a pipe was a convenient simplifying assumption that resulted in a conservative accident scenario. This conservative accident scenario has now become entrenched as the design basis accident for: containment design, shutdown system design, emergency fuel cooling systems design, and to establish environmental qualification temperature and pressure conditions. The requirement to address dynamic effects associated with the postulated pipe rupture subsequently evolved. The dynamic effects include: potential missiles, pipe whipping, blowdown jets, and thermal-hydraulic transients. Recent advances in fracture mechanics research have demonstrated that certain pipes under specific conditions cannot crack in ways that result in an instantaneous guillotine rupture. Canadian utilities are now using mechanistic fracture mechanics and leak-before-break assessments on a case-by-case basis, in limited applications, to support licensing cases which seek exemption from the need to consider the various dynamic effects associated with postulated instantaneous catastrophic rupture of high-energy piping systems inside and outside of containment.

  9. CANDU plant life management - An integrated approach

    International Nuclear Information System (INIS)

    Commercial versions of CANDU reactors were put into service starting more than 25 years ago. The first unit of Ontario Hydro's Pickering A station was put into service in 1971, and Bruce A in 1977. Most CANDU reactors, however, are only now approaching their mid-life of 15 to 20 years of operation. In particular, the first series of CANDU 6 plants which entered service in the early 1980's were designed for a 30 year life and are now approaching mid life. The current CANDU 6 design is based on a 40 year life as a result of advancement in design and materials through research and development. In order to assure safe and economic operation of these reactors, a comprehensive CANDU Plant Life Management (PLIM) program is being developed from the knowledge gained during the operation of Ontario Hydro's Pickering, Bruce, and Darlington stations, worldwide information from CANDU 6 stations, CANDU research and development programs, and other national and international sources. This integration began its first phase in 1994, with the identification of most of the critical systems structures and components in these stations, and a preliminary assessment of degradation and mechanisms that could affect their fitness for service for their planned life. Most of these preliminary assessments are now complete, together with the production of the first iteration of Life Management Plans for several of the systems and components. The Generic CANDU 6 PLIM program is now reaching its maturity, with formal processes to systematically identify and evaluate the major CSSCs in the station, and a plan to ensure that the plant surveillance, operation, and maintenance programs monitor and control component degradation well within the original design specifications essential for the plant life attainment. A Technology Watch program is being established to ensure that degradation mechanisms which could impact on plant life are promptly investigated and mitigating programs established. The

  10. Cobalt-60 production in CANDU power reactors

    International Nuclear Information System (INIS)

    The technology for cobalt-60 production in CANDU reactors, designed and developed by MDS Nordion and AECL, has been safely,economically and successfully employed in CANDU reactors with over 195 reactor years of production. Today over forty percent of the world's disposable medical supplies are made safer through sterilization using cobalt-60 sources from MDS Nordion. Over the past 40 years, MDS Nordion with its CANDU reactor owner partners, has safely and reliably shipped more than 500 million curies of cobalt-60 sources to customers around the world

  11. Thermal fatigue screening criteria for identifying susceptible piping components in CANDU stations

    Energy Technology Data Exchange (ETDEWEB)

    Schefski, C.; Chen, Q.; Pentecost, S. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2011-07-01

    In December 1987, a fatigue failure in a non-isolable section of a safety injection line at the Farley-2 plant prompted the U.S. Nuclear Regulatory Commission (NRC) to issue Bulletin 88-08 requiring that U.S. utilities review all non-isolable branch lines to determine if they are susceptible to thermal fatigue. The thermal fatigue incident at Farley-2 was caused by stresses in the pipe wall resulting from large-scale temperature fluctuations. Shortly after the Farley-2 event, several other incidents with through-wall cracks due to thermal fatigue had occurred in plant subsystems and piping configurations similar to the Farley-2 safety injection line. Thermal fatigue cracks have also occurred in piping configurations with different geometries, such as drain, residual heat removal, and shutdown cooling suction lines in various pressurized water reactors (PWR) and boiling water reactors (BWR). Thermal fatigue, caused by local thermal stratification phenomena, has received significant attention in the PWR and BWR communities in the past two decades. Although CANDU stations have experienced relatively few thermal fatigue failures; the impact of this known fatigue mechanism for CANDU designs has not been rigorously assessed. Screening and evaluation methodology, which has been developed by Electric Power Research Institute (EPRI) to identify locations susceptible to thermal cycling in PWR systems, has recently been modified under a CANDU Owners Group (COG) project for application in CANDU piping systems. This paper describes a new software tool for evaluating locations susceptible to thermal fatigue in CANDU piping systems in an effort to avoid failures that lead to costly plant shutdowns. The software, combined with engineering judgement, will assist CANDU station staff to focus their inspections on key components, therefore reducing dose, time and cost during outages. Computational Fluid Dynamics (CFD) was used to form the basis for expanding the range of validity in

  12. Vertical laryngeal position and oral pressure variations during resonance tube phonation in water and in air. A pilot study.

    Science.gov (United States)

    Wistbacka, Greta; Sundberg, Johan; Simberg, Susanna

    2016-10-01

    Resonance tube phonation in water (RTPW) is commonly used in voice therapy, particularly in Finland and Sweden. The method is believed to induce a lowering of the vertical laryngeal position (VLP) in phonation as well as variations of the oral pressure, possibly inducing a massage effect. This pilot study presents an attempt to measure VLP and oral pressure in two subjects during RTPW and during phonation with the free tube end in air. VLP is recorded by means of a dual-channel electroglottograph. RTPW was found to lower VLP in the subjects, while it increased during phonation with the tube end in air. RTPW caused an oral pressure modulation with a bubble frequency of 14-22 Hz, depending mainly on the depth of the tube end under the water surface. The results indicate that RTPW lowers the VLP instantly and creates oral pressure variations. PMID:26033381

  13. In vitro estimation of pressure drop across tracheal tubes during high-frequency percussive ventilation

    International Nuclear Information System (INIS)

    Tracheal tubes (TT) are used in clinical practice to connect an artificial ventilator to the patient's airways. It is important to know the pressure used to overcome tube impedance to avoid lung injury. Although high-frequency percussive ventilation (HFPV) has been increasingly used, the mechanical behavior of TT under HFPV has not yet been described. Thus, we aimed at characterizing in vitro the pressure drop across TT (ΔPTT) by identifying the model that best fits the measured pressure–flow (P– V-dot ) relationships during HFPV under different working pressures (PWork), percussive frequencies and mechanical loads. Three simple models relating ΔPTT and flow ( V-dot ) were tested. Model 1 is characterized by linear resistive [Rtube ⋅  V-dot (t)] and inertial [I ⋅ V¨(t)] terms. Model 2 takes into consideration Rohrer's approach [K1 ⋅  V-dot (t) + K2 ⋅  V-dot 2(t)] and inertance [I ⋅ V¨(t)]. In model 3 the pressure drop caused by friction is represented by the non-linear Blasius component [Kb ⋅  V-dot 1.75(t)] and the inertial term [I ⋅ V¨(t)]. Model 1 presented a significantly higher root mean square error of approximation than models 2 and 3, which were similar. Thus, model 1 was not as accurate as the latter, possibly due to turbulence. Model 3 presented the most robust resistance-related coefficient. Estimated inertances did not vary among the models using the same tube. In conclusion, in HFPV ΔPTT can be easily calculated by the physician using model 3. (paper)

  14. Highly sensitive contact pressure measurements using FBG patch in endotracheal tube cuff

    Science.gov (United States)

    Correia, R.; Blackman, O. R.; Hernandez, F. U.; Korposh, S.; Morgan, S. P.; Hayes-Gill, B. R.; James, S. W.; Evans, D.; Norris, A.

    2016-05-01

    A method for measuring the contact pressure between an endotracheal tube cuff and the trachea was designed and developed by using a fibre Bragg grating (FBG) based optical fibre sensor. The FBG sensor is encased in an epoxy based UV-cured cuboid patch and transduces the transversely loaded pressure into an axial strain that induces wavelength shift of the Bragg reflection. The polymer patch was created by using a PTFE based mould and increases tensile strength and sensitivity of the bare fibre FBG to pressure to 2.10×10-2 nm/kPa. The characteristics of the FBG patch allow for continuous measurement of contact pressure. The measurement of contact pressure was demonstrated by the use of a 3D printed model of a human trachea. The influence of temperature on the measurements is reduced significantly by the use of a second FBG sensor patch that is not in contact with the trachea. Intracuff pressure measurements performed using a commercial manometer agreed well with the FBG contact pressure measurements.

  15. Crystallographic Phases, Texture and Dislocation Densities of ZR2.5%NB Pressure Tubes at Different Stages of Manufacturing

    International Nuclear Information System (INIS)

    Neutron diffraction experiments have been performed on specimens produced from Zr2.5%Nb pressure tubes, in order to characterize the crystallographic phases, texture and dislocation densities at different stages of a new manufacturing schedule developed in Argentina. Experiments were performed on ENGIN-X a time-of-flight neutron strain scanner at the Isis Facility, UK, using an optimized measurement strategy that exploits the well-known crystallographic texture of the pressure tubes. (author)

  16. Monte Carlo Study on Gas Pressure Response of He-3 Tube in Neutron Porosity Logging

    Directory of Open Access Journals (Sweden)

    TIAN Li-li;ZHANG Feng;WANG Xin-guang;LIU Jun-tao

    2016-10-01

    Full Text Available Thermal neutrons are detected by (n,p reaction of Helium-3 tube in the compensated neutron logging. The helium gas pressure in the counting area influences neutron detection efficiency greatly, and then it is an important parameter for neutron porosity measurement accuracy. The variation law of counting rates of a near detector and a far one with helium gas pressure under different formation condition was simulated by Monte Carlo method. The results showed that with the increasing of helium pressure the counting rate of these detectors increased firstly and then leveled off. In addition, the neutron counting rate ratio and porosity sensitivity increased slightly, the porosity measurement error decreased exponentially, which improved the measurement accuracy. These research results can provide technical support for selecting the type of Helium-3 detector in developing neutron porosity logging.

  17. Ear tube insertion

    Science.gov (United States)

    Myringotomy; Tympanostomy; Ear tube surgery; Pressure equalization tubes; Ventilating tubes; Ear infection - tubes; Otitis - tubes ... trapped fluid can flow out of the middle ear. This prevents hearing loss and reduces the risk ...

  18. Ultrasonic estimation of hydride degradation of zirconium pressure tubes of RBMK fuel channel

    International Nuclear Information System (INIS)

    Fuel channels of nuclear reactors, which are major structural elements of a reactor core, have to meet strict requirements in terms of operational reliability. The middle part of the fuel channel, located in a graphite stack, is a tube made of a zirconium-2.5% niobium alloy. However, zirconium alloys can pick up hydrogen during operation as a consequence of corrosion reaction with water. Hydrogen redistributes easily at elevated temperatures migrating down a temperature or concentration gradient and up a stress gradient. When the terminal solid solubility is exceeded in a component such as a pressure tube that is highly stressed for long periods of time, delayed hydride cracking failures may occur. To estimate degradation of the zirconium alloy in the presence of hydrides, predetermined amounts of hydrogen were added to the sections of the fuel channel tubes by electrolytic deposition of a layer of hydride on the surface of the pressure tube material followed by dissolving the hydride layer by diffusion annealing at an elevated temperature. For estimation of the concentration of zirconium hydride platelets in the zirconium alloy test samples ultrasonic testing methods were proposed. The first method is based on precise measurement of velocity of longitudinal and shear wave at different directions and the second is based on the investigation of high frequency ultrasonic signals backscattered in a focal zone of an ultrasonic transducer. The experimental investigations were performed on the zirconium alloy samples of different concentration of hydrides in the immersion tank at a room temperature. The results obtained on testing samples using different excitation conditions and different types of ultrasonic waves are presented. (orig.)

  19. Value added services to CANDU plants

    International Nuclear Information System (INIS)

    Over the last decade or so, nuclear power plants, just like other types of electricity generating plants, have been facing a number of challenges. Depending on the operating environment of the utility, these challenges are forcing plant owners to examine all facets of the operating costs. Privatization, deregulation and economics of alternative electricity generation methods are exerting enormous pressure on nuclear power plants to streamline costs and improve their operational performance. CANDU reactors are no exception to these forces and face similar pressures. In particular, operating plants that are contemplating plant life extensions are being required to clearly demonstrate the economics of continued operation over other forms of power generation available to the utility. Improvement of capacity factors has the effect of increasing the revenues from the plant and as these revenues increase, the fixed portion of the plant costs including OM and A costs become a smaller percentage of the total revenues. Similar results can be achieved by aiming to reduce the plant OM and A costs. In reality, most well-planned intervention schemes directed at reducing OM and A costs tend to also increase the plant availability. Following plant turnover after commissioning, AECL has been supporting the CANDU owners and utilities with an assortment of products and services dealing with plant operations and outage management issues. AECL has taken the lead in arranging specialized resources, products and services by teaming with other complementary organizations to provide a complete suite of services. Recent examples of such support to operating CANDU plants will be described in the paper. AECL is responding to this changing business environment in two important ways. First, AECL is changing from simply providing a service to its clients towards providing value, something much more important. To this end, AECL is looking to other organizations to form alliances, partnerships and

  20. Future fuel cycle development for CANDU reactors

    International Nuclear Information System (INIS)

    The CANDU reactor has proven to be safe and economical and has demonstrated outstanding performance with natural uranium fuel. The use of on-power fuelling, coupled with excellent neutron economy, leads to a very flexible reactor system with can utilize a wide variety of fuels. The spectrum of fuel cycles ranges from natural uranium, through slightly enriched uranium, to plutonium and ultimately thorium fuels which offer many of the advantages of the fast breeder reactor system. CANDU can also burn the recycled uranium and/or the plutonium from fuel discharged from light water reactors. This synergistic relationship could obviate the need to re-enrich the reprocessed uranium and allow a simpler reprocessing scheme. Fule management strategies that will permit future fuel cycles to be used in existing CANDU reactors have been identified. Evolutionary design changes will lead to an even greater flexibility, which will guarantee the continued success of the CANDU system. (author)

  1. Experimental determination of thermal contact conductance between pressure and calandria tubes of Indian pressurised heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dureja, A.K., E-mail: akdureja@barc.gov.in [Reactor Design & Development Group, Bhabha Atomic Research Centre, Mumbai (India); Pawaskar, D.N.; Seshu, P. [Department of Mechanical Engineering, Indian Institute of Technology Bombay, Mumbai (India); Sinha, S.K. [Reactor Design & Development Group, Bhabha Atomic Research Centre, Mumbai (India); Sinha, R.K. [Department of Atomic Energy, OYC, Near Gateway of India, Mumbai (India)

    2015-04-01

    Highlights: • We established an experimental facility to measure thermal contact conductance between disc shaped specimens. • We measured thermal contact conductance between Zr-2.5Nb alloy pressure tube (PT) material and Zr-4 calandria tube (CT) material. • We concluded that thermal contact conductance is a linear function of contact pressure for interface of PT and CT up to 10 MPa contact pressure. • We concluded that thermal contact conductance is a weak function of interface temperature. - Abstract: Thermal contact conductance (TCC) is one of the most important parameters in determining the temperature distribution in contacting structures. Thermal contact conductance between the contacting structures depends on the mechanical properties of underlying materials, thermo-physical properties of the interstitial fluid and surface condition of the structures coming in contact. During a postulated accident scenario of loss of coolant with coincident loss of emergency core cooling system in a tube type heavy water nuclear reactor, the pressure tube is expected to sag/balloon and come in contact with outer cooler calandria tube to dissipate away the heat generated to the moderator. The amount of heat thus transferred is a function of thermal contact conductance and the nature of contact between the two tubes. An experimental facility was designed, fabricated and commissioned to measure thermal contact conductance between pressure tube and calandria tube specimens. Experiments were conducted on disc shaped specimens under axial contact pressure in between mandrels. Experimental results of TCC and a linear correlation as a function of contact pressure have been reported in this paper.

  2. CATHENA analysis of CANDU 6 steam generators for steam main break at a remote location

    International Nuclear Information System (INIS)

    CATHENA (Canadian Algorithm for THErmalhydraulic Network Analysis) is a nonequilibrium, two-phase, two fluid network analysis code that has been in use for over two decades by various groups in Canada and around the world. It is the primary Thermalhydraulics network analysis tool used by Atomic Energy Canada Ltd. (AECL) in the design, safety and licensing analysis of power and research reactors as well as test facilities. In the thermalhydraulic model, the liquid and vapor phases may have different pressures, velocities, and temperatures. The objective of the present paper is to present the detailed modeling of a CANDU 6 Steam Generator (SG) using the transient, thermalhydraulics network code CATHENA. The model represents the secondary side, primary side and the main steam system including the main steam line up to the assumed break location. The present model is designed such that the transient pressure drops across Tube Support Plates (TSP) could be extracted. The resistances of degraded/fouled TSPs were modeled by using the increased/reduced flow area of the TSPs. CATHENA then calculates the flow resistance in two-phase flow based on the area contraction/expansion at the TSPs. Three sets of simulations were performed; one with the degraded steam generator data provided by the utility users, and the other two with waterlanced (cleaned with high pressure water jet) TSPs. One run assumed the flow area increased by 25 percent, the other run assumed the flow area increased by 50 percent. on the hot side of the SG. No significant changes in the break discharge rates were observed between the runs. However, the steam generator downcomer flow for the waterlanced case did not reverse during the blowdown as was calculated for the degraded case. As expected, the pressure drop across the TSPs was decreased in the waterlanced cases comparing with degraded cases. The CATHENA simulation provides estimates of the velocity, density, and quality in the tube bundle as well as

  3. Experimental Investigation of Heat Transfer and Pressure Drop Characteristics of H-type Finned Tube Banks

    Directory of Open Access Journals (Sweden)

    Heng Chen

    2014-11-01

    Full Text Available H-type finned tube heat exchanger elements maintain a high capacity for heat transfer, possess superior self-cleaning properties and retain the ability to effect flue gas waste heat recovery in boiler renovations. In this paper, the heat transfer and pressure drop characteristics of H-type finned tube banks are studied via an experimental open high-temperature wind tunnel system. The effects of fin width, fin height, fin pitch and air velocity on fin efficiency, convective heat transfer coefficient, integrated heat transfer capacity and pressure drop are examined. The results indicate that as air velocity, fin height and fin width increase, fin efficiency decreases. Convective heat transfer coefficient is proportional to fin pitch, but inversely proportional to fin height and fin width. Integrated heat transfer capacity is related to fin efficiency, convective heat transfer coefficient and finned ratio. Pressure drop increases with the increase of fin height and fin width. Finally, predictive correlations of fin efficiency, Nusselt number and Euler Number are developed based on the experimental data.

  4. Development of a best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, M. K.; Lee, W. J. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2001-03-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool , i.e. RELAP5/MOD3. This scope of project is a fourth step of the whole project, applying the RELAP5/MOD3/CANDU+ version for the real CANDU plant LOCA Analysis and D2O leakage incident. There are three main models under investigation, i.e. Moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs, especially when CANDU LOCA is tested. Also, for Wolsung unit 1 D2O leakage incident analysis, the plant behavior is predicted with the newly developed version for the first 1000 seconds after onset of the incident, with the main interest aiming for system pressure, level control system, and thermal hydraulic transient behavior of the secondary system. The model applied for this particular application includes heat transfer model of nuclear fuel assembly, decay heat model, and MOV (Motor Operated Valve) model. Finally, the code maintenance work, mainly correcting the known errors, is presented. 12 refs., 26 figs., 3 tabs. (Author)

  5. Development of best estimate auditing code for CANDU thermal hydraulic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Hwnag, M

    2001-04-01

    The main purpose of this study is to develop a thermal hydraulic auditing code for the CANDU reactor, modifying the model of existing PWR auditing tool , i.e. RELAP5/MOD3. This scope of project is a fourth step of the whole project, applying the RELAP5/MOD3/CANDU+ version for the real CANDU plant LOCA Analysis and D2O leakage incident. There are three main models under investigation, i.e. Moody critical flow model, flow regime model of horizontal CANDU bundle, and fuel element heatup model when the stratification occurs, especially when CANDU LOCA is tested. Also, for Wolsung unit 1 D2O leakage incident analysis, the plant behavior is predicited with the newly developed version for the first 1000 seconds after onset of the incident, with the main interest aiming for system pressure, level control system, and thermal hydraulic transient behavior of the secondary system. The model applided for this particular application includes heat transfer model of nuclear fuel assembly, decay heat model, and MOV (Motor Operated Valve) model. Finally, the code maintenance work, mainly correcting the known errors, is presented.

  6. The third generation CANDU control room

    International Nuclear Information System (INIS)

    In CANDU stations, as in most complex industrial plants, the man/machine interface design has progressed through three generations. First Generation control rooms consisted entirely on fixed, discrete components (handswitches, indicator lights, strip chart, recorder, annunciator windows, etc.). Human factors input was based on intuitive common sense factors which varied considerably from one designer to another. Second Generation control rooms incorporated video display units and keyboards in the control panels. Computer information processing and display are utilized. There is systematic application of human factors through ergonomic and anthropometric standards and cookbooks. The human factors are applied mainly to the physical layout of the control panels and the physical manipulation performed by the operators. Third Generation control rooms exploit the dramatic performance/cost improvements in computer, electronic display and communication technologies of the 1980's. Further applications of human factors address the cognitive aspects of operator performance. At AECL, second generation control rooms were installed on CANDU stations designed in the mid 70s and early 80s. Third generation features will be incorporated in the CANDU 3 station design and future CANDU stations. There have been significant improvements in the man/machine interface in CANDU stations over the past three decades. The continuing rapid technological developments in computers and electronics coupled with an increasing understanding and application of human factors principles is leading to further enhancements. This paper outlines progress achieved in earlier stations and highlights the features of the CANDU 3rd generation control room. (author). 13 refs, 5 figs

  7. Drift effects in CANDU reactors

    International Nuclear Information System (INIS)

    The diffusion equation is an approximation to the transport equation which relies on the validity of Fick's law. Since this law is not explicitly integrated in the transport equation it can only be derived approximately using homogenization theories. However, such homogenization theories state that when the cell is not symmetric Fick's law breaks down due to the presence of an additional term to the neutron current, called the drift term. In fact, this term can be interpreted as a transport correction to Fick's law which tends to increase the neutron current in a direction opposite to that specified by the flux gradient. In this paper, we investigate how the presence of asymmetric liquid zone controllers will modify the flux distribution inside a CANDU core. 5 refs., 2 figs., 1 tab

  8. A design basis for the development of advanced CANDU control centres

    International Nuclear Information System (INIS)

    The basic design for current CANDU control centres was established in the early 1970's. Plants constructed since then have, for the most part, retained the same basic design. Several factors have led to the need to re-examine CANDU control centre design for plants to be built beyond the year 2000. These factors include the changing roles and responsibilities for the operations staff, an improved understanding of operational issues associated with supervisory control, an improved understanding of human error in operational situations, the opportunity for improved plant performance through the introduction of new technologies, and marketing pressures. This paper describes the proposed design bases for the development of advanced control centres to be implemented in CANDU plants beyond the year 2000. Four areas have been defined covering design goals, design principles, operational bases, and plant functional bases. (author)

  9. Simulation of CANDU Fuel Behaviour into In-Reactor LOCA Tests

    International Nuclear Information System (INIS)

    The purpose of this work is to simulate the behaviour of an instrumented, unirradiated, zircaloy sheathed UO2 fuel element assembly of CANDU type, subjected to a coolant depressurization transient in the X-2 pressurized water loop of the NRX reactor at the Chalk River Nuclear Laboratories in 1983. The high-temperature transient conditions are such as those associated with the onset of a loss of coolant accident (LOCA). The data and the information related to the experiment are those included in the OECD/NEA-IFPE Database (IFPE/CANDU-FIO-131 NEA-1783/01). As tool for this simulation is used the TRANSURANUS fuel performance code, developed at ITU, Germany, along with the corresponding fabrication and in-reactor operating conditions specific of the CANDU PHWR fuel. The results, analyzed versus the experimental ones, are encouraging and perfectible. (author)

  10. Failure assessment and evaluation of critical crack length for a fresh Zr-2 pressure tube of an Indian PHWR

    International Nuclear Information System (INIS)

    Fracture analysis of Zr-2 pressure tubes having through wall axial crack was done using finite element method. The analysis was done for tubes in as received condition. During reactor operation the mechanical properties of Zr-2 undergo changes. The analysis is valid for pressure tubes of newly commissioned reactors. The main aim of the study was to determine critical crack length of pressure tubes in normal operating conditions. Elastic plastic fracture analysis was done for different crack lengths to determine applied J-integral values. Tearing modulus instability concept was used to evaluate critical crack length. One of the important parameter studied was, the effect of crack face pressure, which leaking fluid exert on the crack faces/lips of through wall axial crack. Its effect was found to be significant for pressure tubes. It increases the applied J-integral values. Approximate analytical solutions which takes into account the plasticity ahead of crack tip, are available and widely used. These formulae do not take into account the crack face pressure. Since, for the present situation the effect of crack face pressure is significant hence, detailed finite analysis was necessary. Detailed 3D finite element analysis gives an insight into the variation of J-integral values over the thickness of pressure tube. It was found that J values are maximum at the middle layer of the tube. A peak factor on J values was defined and evaluated as ratio of maximum J to average J across the thickness, crack opening area for each length was also evaluated. The knowledge of crack opening area is useful for leak before break studies. The failure assessment was also done using Central Electricity Generating Board (CEGB) R-6 method considering the ductile tearing. The reserve factors (or safety margins) for different crack lengths was evaluated using R-6 method. (author). 30 refs., 21 figs., 34 tabs

  11. Experimental study on the pressure and pulse wave propagation in viscoelastic vessel tubes-effects of liquid viscosity and tube stiffness.

    Science.gov (United States)

    Ikenaga, Yuki; Nishi, Shohei; Komagata, Yuka; Saito, Masashi; Lagrée, Pierre-Yves; Asada, Takaaki; Matsukawa, Mami

    2013-11-01

    A pulse wave is the displacement wave which arises because of ejection of blood from the heart and reflection at vascular bed and distal point. The investigation of pressure waves leads to understanding the propagation characteristics of a pulse wave. To investigate the pulse wave behavior, an experimental study was performed using an artificial polymer tube and viscous liquid. A polyurethane tube and glycerin solution were used to simulate a blood vessel and blood, respectively. In the case of the 40 wt% glycerin solution, which corresponds to the viscosity of ordinary blood, the attenuation coefficient of a pressure wave in the tube decreased from 4.3 to 1.6 dB/m because of the tube stiffness (Young's modulus: 60 to 200 kPa). When the viscosity of liquid increased from approximately 4 to 10 mPa·s (the range of human blood viscosity) in the stiff tube, the attenuation coefficient of the pressure wave changed from 1.6 to 3.2 dB/m. The hardening of the blood vessel caused by aging and the increase of blood viscosity caused by illness possibly have opposite effects on the intravascular pressure wave. The effect of the viscosity of a liquid on the amplitude of a pressure wave was then considered using a phantom simulating human blood vessels. As a result, in the typical range of blood viscosity, the amplitude ratio of the waves obtained by the experiments with water and glycerin solution became 1:0.83. In comparison with clinical data, this value is much smaller than that seen from blood vessel hardening. Thus, it can be concluded that the blood viscosity seldom affects the attenuation of a pulse wave.

  12. Mitigating aging in CANDU plants

    International Nuclear Information System (INIS)

    Aging degradation is a phenomenon we all experience throughout life, both on a personal basis and in business. Many industries have been successful in postponing the inevitable impact on their related systems and components through programs to maintain long-term reliability, maintainability and safety. However, this has not always been the case for nuclear power. While all power plants are experiencing the world trend of increasing operating costs with age, few (if any) have been able to fully define the parameters that solve the aging equation, particularly in relation to major components. Inspection and preventive maintenance have not been effective in predicting life-limiting degradation and failure. In CANDU nuclear plants, utilities are taking a comprehensive approach in dealing with the aging problem. Programs have been established to identify the current condition and degradation mechanisms of critical components, the failure of which would impact negatively on station competitiveness and safety. These include subcomponents under the general headings of reactor components, civil structures, piping (nuclear and conventional), steam generators, turbines and cables. In support of these efforts, R and D projects have been defined under the CANDU Owners Group to deal with generic issues on aging common to its members (e.g., investigation of degradation mechanisms, development of tools and techniques to mitigate the effects of aging, etc.). This paper describes recent developments of this cost-shared program with specific reference to concrete aging and crack repairs, flow-assisted corrosion in piping, elastomer service life, cable aging, degradation mechanisms in steam generators and lubricant breakdown. (author)

  13. Elastic-plastic fracture mechanics analyses of cracked steam generator tubes under internal pressure

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyeong Keun; Ahn, Min Yong; Moon, Seong In; Chang, Yoon Suk; Kim, Young Jin [Sungkyunkwan Univ., Suwon (Korea, Republic of); Hwang, Seong Sik; Kim, Joung Soo [KAERI, Taejon (Korea, Republic of)

    2005-07-01

    The structural and leakage integrity of steam generator tube should be maintained during operation even though a crack is existed on it. During the past three decades, several limit load solutions have been proposed to resolve the integrity issue. However, for exact load carrying capacity estimation of specific components under different conditions, these solutions have to be modified by using lots of experimental data. The purpose of this paper is to introduce a new burst pressure estimation scheme based on fracture mechanics analyses for steam generator tube with an axial or circumferential through-wall crack. To do this, closed-form engineering equations were derived to get relevant parameters from three dimensional elastic-plastic finite element analyses combined with reference stress method. Also, a series of structural integrity analyses were carried out using the calculated J-integral from engineering equations and fracture toughness data. Thereby, in comparison with the experimental data as well as corresponding estimation results from limit load solutions, it was proven that the proposed estimation scheme can be used as an efficient tool for integrity evaluation of cracked steam generator tubes.

  14. Use of pressurized eccentric tubes to study the effect of hydrostatic stress on swelling

    International Nuclear Information System (INIS)

    A technique for measuring the effect of hydrostatic stress on radiation-induced swelling is presented. This technique is based on the nonuniform hydrostatic stress that arises when an eccentric tube (a tube with inner and outer surfaces having dissimilar centers of revolution) is internally pressurized. The elastic analyses of the thin- and thick-walled eccentric tube are given. The elastic stress state is allowed to relax plastically, based on a constitutive law for deformation during neutron irradiation. In this case, the constitutive law contains a linearly stress-dependent deviatoric strain rate and a dilatation rate that is linearly dependent on hydrostatic stress. Emphasis is placed on the specimen design and experimental procedure for in-reactor experiments in which the coefficient relating hydrostatic stress and swelling is sought. It is shown that, for the 316L stainless steel specimens placed in EBR-II, we may expect that any appreciable effect of hydrostatic stress on swelling will be observable through changes in specimen curvature

  15. The failure of the pressure tube in fuel channel NO6 of Bruce NGS-A unit 2 in March 1986

    International Nuclear Information System (INIS)

    The events leading up to the failure of Bruce NGS-A Unit 2 pressure tube NO6 are described. Based on the subsequent examination of the tube, reactor operating history and pressure tube manufacturing information, the failure sequence is deduced. Remedial actions which have been adopted for avoiding similar occurrences are listed

  16. CANDU reactors, their regulation in Canada, and the identification of relevant NRC safety issues

    International Nuclear Information System (INIS)

    Atomic Energy of Canada, Limited (AECL) and its subsidiary in the US, are considering submitting the CANDU 3 design for standard design certification under 10 CFR Part 52. CANDU reactors are pressurized heavy water power reactors. They have some substantially different safety responses and safety systems than the LWRs that the commercial power reactor licensing regulations of the US Nuclear Regulatory Commission (NRC) have been developed to deal with. In this report, the authors discuss the basic design characteristics of CANDU reactors, specifically of the CANDU 3 where possible, and some safety-related consequences of these characteristics. The authors also discuss the Canadian regulatory provisions, and the CANDU safety systems that have evolved to satisfy the Canadian regulatory requirements as of December 1992. Finally, the authors identify NRC regulations, mainly in 10 CFR Parts 50 and 100, with issues for CANDU 3 reactor designs. In all, eleven such regulatory issues are identified. They are: (1) the ATWS rule (section 50.62); (2) station blackout (section 50.63); (3) conformance with Standard Review Plan (SRP); (4) appropriateness of the source term (section 50.34(f) and section 100.11); (5) applicability of reactor coolant pressure boundary (RCPB) requirements (section 50.55a, etc); (6) ECCS acceptance criteria (section 50.46)(b); (7) combustible gas control (section 50.44, etc); (8) power coefficient of reactivity (GDC 11); (9) seismic design (Part 100); (10) environmental impacts of the fuel cycle (section 51.51); and (11) (standards section 50.55a)

  17. The pressure field in the liquid column in the tube-arrest method

    Institute of Scientific and Technical Information of China (English)

    Ying Chong-Fu; Li Chao; Xu De-Long; Deng Jing-Jun

    2008-01-01

    We have been using the method of tube-arrest as a means of producing transient single cavitation bubble. In the present paper we seek to comprehend the mechanism of production and inquire into the structure of the ab initio pressure field in the arrested liquid column. The generated pressure wave is shown by combining the theoretical analysis with the experimental observation to be a slightly varied version of water hammer. With relatively clean liquid, the magnitude of the tension peak generating the TSB is likely to reach of several millions Pa. It is also shown that the so generated cavitation bubble originating from the gas-containing bulk liquid is in 'violent' motion.

  18. Characteristics of gas-liquid two-phase flow in a vertical small diameter tube at a medium pressure

    International Nuclear Information System (INIS)

    Most of correlations for calculating two-phase flow parameters, such as flow pattern transitions, void fraction and pressure drop, have been developed based on the experimental data on tubes greater than 10 mm in diameter at near atmospheric pressures. For that, the applicability of such correlations is doubtful to the flow in small diameter tubes at a medium pressure as seen in compact heat exchangers like residential room conditioners. In this connection, the purpose of this study is to provide experimental data for gas-liquid two-phase vertical flows in a small diameter tube at medium pressures since the published data for such flows is limited to examine existing correlations and/or develop a new one. Experiments have been conducted on air-water two-phase flows in a vertical circular tube of 9.48 mm internal diameter. In the experiment, system pressure in the channel has been systematically changed from 0.2 to 0.7 MPa (absolute) to study the effect of the pressure on two-phase flow parameters, i.e., two-phase flow pattern transitions, bubble size in bubble flow, void fraction, interfacial shear force, frictional pressure drop and static pressure fluctuations. Furthermore, the respective data obtained have been compared with existing correlations. (author)

  19. Assessment of RELAP5/CANDU+ code for regulatory auditing analysis of CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Young Seok; Kim, Hho Jung; Yang, Chae Yong

    2001-12-15

    The objectives of this study are to undertake the verification and validation of RELAP5/CANDU+ code, which is developed in this project, by simulating the B8711 test of RD-14 facility, and to examine the properties of this code by doing the sensitivity analysis for experimental prediction modes about thermal-hydraulics phenomena in CANDU reactor systems added to this code. The B8711 test was an experiment of a 45% ROH break for simulating large LOCA. Also, in this study, the methods for making input cards related to CANDU options are described, so that some users can use the RELAP5/CANDU+ code with easy. RELAP/CANDU+ code can choose the options of Henry-Fauske mode, Ransom-Trapp model, and Moody model for prediction of the critical mass flow. It is examined that Henry-Fauske model and Ransom-Trapp model are considered properly, but Moody model is still required to be improved. Heat transfer correlations available in RELAP5/CANDU+ code for CANDU-type reactors are a horizontal stratified model, a fuel heat-up model and D2O/H2O CHF correlations, and these models take an important role to improve the predictability of the experimental procedures. It is concluded that RELAP5/CANDU+ code is useful for the auditing of the accident analysis of CANDU reactors, and the results of the sensitivity analysis for thermal-hydraulic models examined in this study are valuable for the actual auditing of real CANDU-type power plants.

  20. Feasibility study of CANDU-9 fuel handling system

    International Nuclear Information System (INIS)

    CANDU's combination of natural uranium, heavy water and on-power refuelling is unique in its design and remarkable for reliable power production. In order to offer more output, better site utilization, shorter construction time, improved station layout, safety enhancements and better control panel layout, CANDU-9 is now under development with design improvement added to all proven CANDU advantages or applicable technologies. One of its major improvement has been applied to fuel handling system. This system is very similar to that of CANDU-3, and some parts of the system are applied to those of the existing CANDU-6 or CANDU-9. Design concepts and design requirements of fuel handling system for CANDU-9 have been identified to compare with those of the existing CANDU and the design feasibilities have been evaluated. (author). 4 tabs., 13 figs., 9 refs

  1. The development of emergency core cooling systems in the PWR, BWR, and HWR Candu type of nuclear power plants

    International Nuclear Information System (INIS)

    Emergency core cooling systems in the PWR, BWR, and HWR-Candu type of nuclear power plant are reviewed. In PWR and BWR the emergency cooling can be catagorized as active high pressure, active low pressure, and a passive one. The PWR uses components of the shutdown cooling system: whereas the BWR uses components of pressure suppression contaiment. HWR Candu also uses the shutdown cooling system similar to the PWR except some details coming out from moderator coolant separation and expensive cost of heavy water. (author)

  2. Pressure drop and stability of flow in Archimedean spiral tube with transverse corrugations

    Directory of Open Access Journals (Sweden)

    Đorđević Milan

    2016-01-01

    Full Text Available Isothermal pressure drop experiments were carried out for the steady Newtonian fluid flow in Archimedean spiral tube with transverse corrugations. Pressure drop correlations and stability criteria for distinguishing the flow regimes have been obtained in a continuous Reynolds number range from 150 to 15 000. The characterizing geometrical groups which take into account all the geometrical parameters of Archimedean spiral and corrugated pipe has been acquired. Before performing experiments over the Archimedean spiral, the corrugated straight pipe having high relative roughness e/d = 0.129 of approximately sinusoidal type was tested in order to obtain correlations for the Darcy friction factor. Insight into the magnitude of pressure loss in the proposed geometry of spiral solar receiver for different flow rates is important because of its effect upon the efficiency of the receiver. Although flow in spiral and corrugated geometries has the advantages of compactness and high heat transfer rates, the disadvantage of greater pressure drops makes hydrodynamic studies relevant. [Projekat Ministarstva nauke Republike Srbije, br. III 42006 i br. TR 33015

  3. CFD simulations of the single-phase and two-phase coolant flow of water inside the original and modified CANDU 37-element bundles

    International Nuclear Information System (INIS)

    Single-phase (inlet temperature of 180° C, outlet pressure of 9 MPa, total power of 2 MW and flow rate of 13.5 Kg/s), and two-phase (inlet temperature of 265° C, outlet pressure of 10 MPa, total power of 7.126 MW and flow rate of 19 Kg/s) water flows inside a CANDU thirty seven element fuel string are simulated using a Computational Fluid Dynamics (CFD) code with parallel processing and results are presented in this paper. The analyses have been performed for the original and modified (11.5 mm center element diameter) designs with skewed cosine axial heat flux distribution and 5.1% diametral creep of the pressure tube. The CFD results are in good agreement with the expected temperature and velocity cross-sectional distributions. The effect of the pressure tube creep on the flow bypass is detected, and the CHF improvement in the core region of the modified design is confirmed. The two-phase flow model reasonably predicted the void distribution and the role of interfacial drag on increasing the pressure drop. In all CFD models, the appendages were shown to enhance the production of cross flows and their corresponding flow mixing and asymmetry. (author)

  4. Overview of methods to increase dryout power in CANDU fuel bundles

    Energy Technology Data Exchange (ETDEWEB)

    Groeneveld, D.C., E-mail: degroeneveld@gmail.com [Chalk River Laboratories, AECL, Chalk River (Canada); University of Ottawa, Department of Mechanical Engineering, Ottawa (Canada); Leung, L.K.H. [Chalk River Laboratories, AECL, Chalk River (Canada); Park, J.H. [Korean Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-06-15

    Highlights: • Small changes in bundle geometry can have noticeable effects on the bundle CHF. • Rod spacing devices can results in increases of over 200% in CHF. • CHF enhancement decays exponentially downstream from spacers. • CHF-enhancing bundle appendages also increase the post-CHF heat transfer. - Abstract: In CANDU reactors some degradation in the CCP (critical channel power, or power corresponding to the first occurrence of CHF in any fuel channel) will occur with time because of ageing effects such as pressure-tube diametral creep, increase in reactor inlet-header temperature, increased hydraulic resistance of feeders. To compensate for the ageing effects, various options for recovering the loss in CCP are described in this paper. They include: (i) increasing the bundle heated perimeter, (ii) optimizing the bundle configuration, (iii) optimizing core flow and flux distribution, (iv) reducing the bundle hydraulic resistance, (v) use of CHF-enhancing bundle appendages, (vi) more precise experimentation, and (vii) redefining CHF. The increase in CHF power has been quantified based on experiments on full-scale bundles and subchannel code predictions. The application of several of these CHF enhancement principles has been used in the development of the 43-rod CANFLEX bundle.

  5. Examining the response pressure along a fluid-filled elastic tube to comprehend Frank's arterial resonance model.

    Science.gov (United States)

    Lin Wang, Yuh-Ying; Sze, Wah-Keung; Lin, Chin-Chih; Chen, Jiang-Ming; Houng, Chin-Chi; Chang, Chi-Wei; Wang, Wei-Kung

    2015-04-13

    Frank first proposed the arterial resonance in 1899. Arteries are blood-filled elastic vessels, but resonance phenomena for a fluid-filled elastic tube has not drawn much attention yet. In this study, we measured the pressure along long elastic tubes in response to either a single impulsive water ejection or a periodic water input. The experimental results showed the low damped pressure oscillation initiated by a single impulsive water input; and the natural frequencies of the tube, identified by the peaks of the response in the frequency domain, were inversely proportional to the length of the tube. We found that the response to the periodic input reached a steady distributed oscillation with the same period of the input after a short transient time; and the optimal pressure response, or resonance, occurred when the pumping frequency was near the fundamental natural frequency of the system. We pointed out that the distributed forced oscillation could also be a suitable approach to analyze the arterial pressure wave. Unlike Frank's resonance model in which the whole arterial system was lumped together to a simple 0-D oscillator and got only one natural frequency, a tube has more than one natural frequency because the pressure P(z,t) is a 1-D oscillatory function of the axial position z and the time t. The benefit of having more than one natural frequency was then discussed. PMID:25773589

  6. Optimization of process parameters for control of hydrogen in Zr-2.5Nb pressure tubes for PHWRs

    International Nuclear Information System (INIS)

    Hydrogen induced problems such as hydrogen embrittlement, blistering and DHC are some of the most critical life limiting factors for PHWR pressure tubes. The pressure tubes pick up on average 1 ppm hydrogen every year during reactor operation. Therefore to extend the life of the pressure tubes by countering the deleterious effects of hydrogen, the initial hydrogen content of the as manufactured pressure tube has to be brought down considerably. Earlier NFC was producing Zr-2.5Nb pressure tubes from double melted ingots with an H content of nearly 25 ppm. Owing to the above consideration, process optimization was carried out to reduce this H content from 25 ppm to 5 ppm. This paper describes the various steps adopted for reduction of the hydrogen content during a series of manufacturing operations and processes such as sponge production, melting, extrusion, pilgering, pickling, cleaning, heat treatment and final finishing operations. Intermediate product analysis and characterization has been carried out to monitor and minimize the hydrogen content at critical process steps. (author)

  7. Applicability of miniature specimen techniques for evaluating the mechanical properties of pressure tube and cladding material

    International Nuclear Information System (INIS)

    The Miniature Specimen Test Techniques (MSTT) namely Small Punch Test (SPT) and Ball Indentation Test (BIT) are commonly employed for evaluating the tensile properties of metallic materials. While discs of 3mm diameter with 0.25mm thickness are utilized for SPT method, the samples for BIT are of any shape with parallel polished top and bottom surfaces having thickness of at least 1mm. The SPT technique is based on driving a ball through clamped miniature disc specimens for deforming till it fractures whereas BIT involves multiple indentations, load and unload cycles at a single indentation point on a polished metallic surface by a spherical indenter. The specimens were fabricated from Zr-2.5%Nb pressure tube (PT) material that is used in pressurised heavy water reactor (PHWR). A suitable die-punch assembly was designed and developed in house to clamp the specimen to carry out the small punch test with the help of a screw-driven universal testing machine. In this work we have utilized miniature tensile specimens for evaluating and comparing the mechanical properties of PT material with that obtained from SPT and BIT. The tensile specimens were prepared using wire cut Electrical Discharge Machining (EDM) as per general guideline of ASTM standard E-8. Tests were carried out at ambient and higher temperatures. The tensile properties obtained from tension test and two MSTTs show that the tensile properties vary with orientation and temperature. In order to evaluate mechanical properties of cladding tube two techniques namely SPT and Ring Tension Test (RTT) have been used. The RTT is another technique, already established for estimation of the mechanical properties of cladding tube material in the transverse direction. Experimental results were generated at ambient and higher temperatures by preparing specimens from the same cladding tube in the form of 3mm discs and rings. The basic fixture that was used for carrying out ring tension test of the cladding tube consists

  8. Numerical Investigation of Air-Side Heat Transfer and Pressure Drop in Circular Finned-Tube Heat Exchangers

    OpenAIRE

    Mon, Mi Sandar

    2009-01-01

    A three-dimensional numerical study is performed to investigate the heat transfer and pressure drop performance on the air-side of circular finned tube bundles in cross flow. New heat transfer and pressure drop correlations for the air-cooled heat exchangers have been developed with the Reynolds number ranging from 5000 to 70000. The heat transfer and pressure drop results agree well with several existing experimental correlations. In addition, the influence of the geometric parameters on the...

  9. Improvement of high-voltage performance of acceleration tubes by cleaning the walls with a high-pressure water jet

    Energy Technology Data Exchange (ETDEWEB)

    Takeuchi, S. E-mail: takeuchi@tandem.tokai.jaeri.go.jp; Nakanoya, T.; Kabumoto, H.; Yoshida, T

    2003-11-11

    We cleaned electrostatic accelerator tubes by applying a high-pressure water jet and examined their high-voltage performances at 1 and 3 MV. The cleaning was very effective in reducing discharge activities at their rated voltages. We did some experimental investigations with the tubes and their ceramic insulators. We found that removal of microparticles loosely bound on the vacuum-side ceramic surfaces had an important effect in eliminating the discharge activities.

  10. THE EFFECTS OF SWIRL GENERATOR HAVING WINGS WITH HOLES ON HEAT TRANSFER AND PRESSURE DROP IN TUBE HEAT EXCHANGER

    Directory of Open Access Journals (Sweden)

    Zeki ARGUNHAN

    2006-02-01

    Full Text Available This paper examines the effect of turbulance creators on heat transfer and pressure drop used in concentric heat exchanger experimentaly. Heat exchanger has an inlet tube with 60 mm in diameter. The angle of swirl generators wings is 55º with each wing which has single, double, three and four holes. Swirl generators is designed to easily set to heat exchanger entrance. Air is passing through inner tube of heat exhanger as hot fluid and water is passing outer of inner tube as cool fluid.

  11. THE EFFECTS OF SWIRL GENERATOR HAVING WINGS WITH HOLES ON HEAT TRANSFER AND PRESSURE DROP IN TUBE HEAT EXCHANGER

    OpenAIRE

    ARGUNHAN, Zeki; Yildiz, Cengiz

    2006-01-01

    This paper examines the effect of turbulance creators on heat transfer and pressure drop used in concentric heat exchanger experimentaly. Heat exchanger has an inlet tube with 60 mm in diameter. The angle of swirl generators wings is 55º with each wing which has single, double, three and four holes. Swirl generators is designed to easily set to heat exchanger entrance. Air is passing through inner tube of heat exhanger as hot fluid and water is passing outer of inner tube as cool fluid.

  12. CANDU 9 safety enhancements and licensability

    International Nuclear Information System (INIS)

    The CANDU 9 design has followed the evolutionary product development approach that has characterized the CANDU family of nuclear power plants. In addition to utilizing proven equipment and systems from operating stations, the CANDU 9 design has looked ahead to incorporate design and safety enhancements necessary to meet evolving utility and regulatory requirements both in Canada and overseas. To demonstrate licensability in Canada, and to assure overseas customers that the design had independent regulatory review in the country of origin, the pre-project Basic Engineering Program included an extensive two year formal review by the Canadian regulatory authority, the Atomic Energy Control Board (AECB). Documentation submitted for this licensing review included the licensing basis, safety requirements and safety analysis necessary to demonstrate compliance with regulations as well as to assess system design and performance. The licensing review was successfully completed in 1997 January. In addition, to facilitate licensability in Korea, CANDU 9 incorporates feedback from the application of Korean licensing requirements to the CANDU 6 reactors at Wolsong site. (author)

  13. CANDU safety and licensing framework and process

    International Nuclear Information System (INIS)

    Nuclear Safety is a shared responsibility of the Industry, public and the Government. The International Atomic Energy Agency's (IAEA) safety fundamentals, basic objectives and safety guides lay down the principles from which requirements, recommendations and methodologies for safety design of Nuclear Power Plants (NPP) are derived. Within the framework of the international regulations and those of the Canadian Nuclear Safety Commission (CNSC), this paper will discuss the overall safety objectives, the defence in depth philosophy guiding CANDU safety, as well as the licensing process defined to meet all applicable CNSC regulations. The application of such philosophy to the ACR design and safety approach will also be discussed along with aspects of its implementation. The role of deterministic analysis, and Probabilistic Safety Analysis (PSA) in the design and licensing process of the Advanced CANDU Reactor will be discussed. Postulated initiating events and their combinations, acceptance criteria, CANDU margins and limits, supporting methodologies and computer codes used in safety analysis will be reviewed. The paper will also note intrinsic safety characteristics of CANDU, some of the ACR passive safety features built-in by design, CANDU distinctive features with respect to severe core damage, mechanisms of heat rejection in those extreme conditions, emergency coolant injection system features and other post accident mitigating systems. Update on the ACR Canadian and US licensing progress will also be provided. (authors)

  14. Evaluation of Pressure Stable Chip-to-Tube Fittings Enabling High-Speed Chip-HPLC with Mass Spectrometric Detection.

    Science.gov (United States)

    Lotter, Carsten; Heiland, Josef J; Stein, Volkmar; Klimkait, Michael; Queisser, Marco; Belder, Detlev

    2016-08-01

    Appropriate chip-to-tube interfacing is an enabling technology for high-pressure and high-speed liquid chromatography on chip. For this purpose, various approaches, to connect pressure resistant glass chips with HPLC pumps working at pressures of up to 500 bar, were examined. Three side-port and one top-port connection approach were evaluated with regard to pressure stability and extra column band broadening. A clamp-based top-port approach enabled chip-HPLC-MS analysis of herbicides at the highest pressure and speed. PMID:27397738

  15. Evaluation of Pressure Stable Chip-to-Tube Fittings Enabling High-Speed Chip-HPLC with Mass Spectrometric Detection.

    Science.gov (United States)

    Lotter, Carsten; Heiland, Josef J; Stein, Volkmar; Klimkait, Michael; Queisser, Marco; Belder, Detlev

    2016-08-01

    Appropriate chip-to-tube interfacing is an enabling technology for high-pressure and high-speed liquid chromatography on chip. For this purpose, various approaches, to connect pressure resistant glass chips with HPLC pumps working at pressures of up to 500 bar, were examined. Three side-port and one top-port connection approach were evaluated with regard to pressure stability and extra column band broadening. A clamp-based top-port approach enabled chip-HPLC-MS analysis of herbicides at the highest pressure and speed.

  16. Performance assessment of an inline horizontal swirl tube cyclone for gas-liquid separation at high pressure

    Institute of Scientific and Technical Information of China (English)

    Nurhayati Mellon; Azmi M. Shariff

    2011-01-01

    The application of swirl tube cyclone for gas-liquid separation is attractive due to its small size and weight.However,very scarce information on the performance of the swirl tube cyclone especially at high operating pressure emulating actual field condition was published in journals.Performance assessment was usually done at a low operating pressure using either air-water,air-fine particle mixtures or dense gas such as SF6.This paper fills the existing gaps and reports the initial findings on the performance assessment of a horizontal swirl tube cyclone for gas-liquid separation operating at a flow rate of 5 MMSCFD at 40-60 bar operating pressure.

  17. CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FOR THE DECOMMISSIONING OF THE HORIZONTAL FUEL CHANNELS IN THE CANDU 6 NUCLEAR REACTOR. PART 9 - CUTTING AND EXTRACTING DEVICE FUNCTIONING

    Directory of Open Access Journals (Sweden)

    Constantin POPESCU

    2015-05-01

    Full Text Available This paper presents a constructive solution proposed by the authors in order to achieve of a cutting and extracting device for the decommissioning of the horizontal fuel channels in the CANDU 6 nuclear reactor. The Cutting and Extraction Device (CED performs the dismantling, cutting and extraction of the fuel channel components. It's a flexible and modular device, which is designed to work inside the fuel channel and has the following functions: moving with variable speed, temperature monitoring and video surveillance inside the pipe, unblock and extract the channel closure plug (from End Fitting - EF, unblock and extract the channel shield plug (from Lattice Tube - LT, block and cut the middle of the pressure tube, block and cut the end of the pressure tube, block and extract the half of pressure tube. The Cutting and Extraction Device (CED consists of following modules: guiding-fixing module, traction modules, cutting module, guiding-extracting module and flexible elements for modules connecting. The guiding-fixing module is equipped with elastic guiding rollers and fixing claws in working position, the traction modules are provided with variable pitch rollers for allowing variable travel speed through the fuel channel. The cutting module is positioned in the middle of the device and it is equipped with three knife rolls for pressure tube cutting, using a system for cutting place video surveillance and pyrometers for monitoring cutting place temperature. The Cutting and Extraction Device (CED is fully automated, connected by wires to a Programmable Logic Controller (PLC and controlled from a Human Machine Interface (HMI. The design of the Cutting and Extraction Device (CED shall be achieved according to the particular features of the fuel channel components to be dismantled and to ensure radiation protection of workers.

  18. An analytic study on LBLOCA for CANDU type reactor using MARS-KS/CANDU

    International Nuclear Information System (INIS)

    This study provides the simulation results using MARS-KS/CANDU code for the Large Break LOCA of CANDU type reactor. The purpose of the study is to evaluate the capability of MARS-KS/CANDU for simulating the actual plants (Wolsong 2/3/4). The steady state and the transient analysis results were provided. After the sensitivity study depend on break size, the case that 35% of the inlet header known as the accident that has the most limiting effect on the temperature of the fuel sheath was calculated. In order to evaluate the results, the results were compared with those of CATHENA simulation. (author)

  19. Isothermal flow measurement using planar PIV in the 1/4 scaled model of CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Im, Sunghyuk; Sung, Hyung Jin [KAIST, Daejeon (Korea, Republic of); Seo, Han; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of); Kim, Hyoung Tae [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    The local temperature of the moderator is a key parameter in determining the available subcooling. To predict the flow field and local temperature distribution in the calandria, Korea Atomic Energy Research Institute (KAERI) started the experimental research on moderator circulation as one of a national R and D research programs from 2012. This research program includes the construction of the Moderator Circulation Test (MCT) facility, production of the validation data for self-reliant CFD tools, and development of optical measurement system using the Particle Image Velocimetry (PIV) and Laser Induced Fluorescence (LIF) techniques. Small-scale 1/40 and 1/8 small-scale model tests were performed prior to installation of the main MCT facility to identify the potential problems of the flow visualization and measurement expected in the 1/4 scale MCT facility. In the 1/40 scale test, a flow field was measured with a PIV measurement technique under an iso-thermal state, and the temperature field was visualized using a LIF technique. In this experiment, the key point was to illuminate the region of interest as uniformly as possible since the velocity and temperature fields in the shadow regions were distorted and unphysical. In the 1/8 scale test, the flow patterns from the inlet nozzles to the top region of the tank were investigated using PIV measurement at two different positions of the inlet nozzle. For each position of laser beam exposure the measurement sections were divided to 7 groups to overcome the limitation of the laser power to cover the relatively large test section. The MCT facility is the large-scale facility designed to reproduce the important characteristics of moderator circulation in a CANDU6 calandria under a range of operating conditions. It is reduced in a 1/4 scale and a moderator test vessel is built to the specifications of the CANDU6 reactor design, where a working fluid is sub-cooled water with atmospheric pressure. Previous studies were

  20. Experimental Evaluation of the Burst Pressure of Steam Generator Tube with Multiple Part-through-wall Cracks

    International Nuclear Information System (INIS)

    Since steam generator (SG) tube is a pressure boundary of pressurized water reactor (PWR), the maintaining integrity of SG tube is very important. However, various types of defect caused by a mechanical and chemical degradations have been observed in the SG tube. In particular, the outer diameter stress corrosion cracking (ODSCC) in the secondary side is a dominant type of defect, which can lead to leakage of primary coolant and burst of SG tube. Thus, the integrity evaluation of SG tubes with SCC is considered to be an important issue. A number of experimental and analytical studies have been conducted to evaluate burst pressure of SG tube with defects and proposed evaluation models. But, most of the models were developed based on single cracks, although SCC initiates and grows at multiple sites on the surface of SG tube. Therefore, this study carried out burst tests using SG tube specimens containing multiple part-trough-wall (PTW) flaws at room temperature (RT) and evaluated burst pressure of SG tubes with multiple PTW flaws. The burst tests were conducted on 56 specimens and burst pressures were obtained. Also, failure mode of SG tube with multiple flaws was investigated by examining the shape of crack and tearing from post-test specimens. The reduction was more pronounced for L=25.4mm than L=6.3mm and was more pronounced when three flaws were arranged than when two flaws were arranged. Burst pressure increased with increasing axial distance between flaws for collinear multiple flaws, whereas the pressure decreased and saturated with increasing circumferential distance between non-aligned multiple flaws. For SG tubes with parallel multiple flaws, the burst pressure was influenced by circumferential distance between flaws and length of flaws. When two flaws were parallel with circumferential distance of l1=1mm, the burst pressure was higher about 2% than that of single flaw for L=6.3mm, but it was lower about 7% than that of single flaw for L=25.4mm. Thus an

  1. Interchain tube pressure effect in extensional flows of oligomer diluted nearly monodisperse polystyrene melts

    DEFF Research Database (Denmark)

    Rasmussen, Henrik K.; Huang, Qian

    2014-01-01

    , if the solvent has less than two Kuhn steps, e.g. is not a chain. The constitutive equation is based on a phenomenological tube-based model within the methodology of the molecular stress function approach. The nonlinear dynamics have been explained as a consequence of a constant thermal interchain pressure......We have derived a constitutive equation to explain the extensional dynamics of oligomer-diluted monodisperse polymers, if the length of the diluent has at least two Kuhn steps. These polymer systems have a flow dynamics which distinguish from pure monodisperse melts and solutions thereof...... times and entanglements have been established based on published extensional experiments on nearly monodisperse polystyrene melts. The constitutive equation has shown agreement with the experimental startup of and steady extension data from Huang et al. (Macromolecules 46:5026–5035, 2013a) based on 285...

  2. Microstructural change of a 9Cr steel longitudinal welded tube under internal pressure creep loading

    International Nuclear Information System (INIS)

    In this study, the microstructure of a base metal, and the heat-affected zone (HAZ) and weld metal of a 9Cr steel longitudinal welded tube with various internal pressure creep damage levels was studied by transmission electron microscopy and energy dispersive X-ray spectroscopy. Each portion of the steel weldment clearly changed with damage level. The present results indicate that the recovery of the microstructure is faster in HAZ than in other portions, and thus creep deformation preferentially occurs in HAZ. Moreover, it was revealed that stress accelerates the growth of M23C6 precipitates as well as the reduction of dislocation density, consequently promoting recovery. It was also confirmed that the so-called type IV failure is reasonably explained by precipitate strengthening.

  3. Detecting Nonlinearity in Pressure Data Inside the Draft Tube of a Real Francis Turbine

    CERN Document Server

    Sello, S

    1995-01-01

    A general method for testing nonlinearity in time series is described and applied to measurements of different pressure data inside the draft tube surge of a real Francis turbine. Comparing the current original time series to an ensemble of surrogates time series, suitably constructed to mimic the linear properties of the original one, we was able to distinguish a linear stochastic from a nonlinear deterministic behaviour and, moreover, to quantify the degree of nonlinearity present in the related dynamics. The problem of detecting nonlinear structure in real data is quite complicated by the influence of various contaminations, like broadband noise and/or long coherence times. These difficulties have been overcame using the combination of a suitable nonlinear filtering technique and a qualitative redundancy statistic analysis. The above investigations allow a quantitative characterization of different dynamical regimes of motion of gas cavities inside real turbines and, moreover, allow to support the reliabil...

  4. An analytical assessment of the longitudinal ridging of CANDU type fuel element

    International Nuclear Information System (INIS)

    There are 380 fuel channels in a CANDU-6 reactor, and twelve fuel bundles are loaded into each fuel channel. High-pressure, heavy water coolant passes through the fuel bundle string to remove heat generated from the fuel. Fuel sheath collapses down around the uranium dioxide pellet due to the coolant pressure when the fuel is loaded into the reactor. Longitudinal ridges may form in CANDU fuel element sheaths as a result of sheath collapse onto the pellets. A static analysis, finite-element (FE) model was developed to simulate the longitudinal ridging of the fuel element with use of the structural analysis computer code ABAQUS. Collapse pressures were calculated for the fifty-one cases for which test results of WCL in 1973 and 1975 are available. Calculation results under-predicted the critical collapse pressure but it showed significant relationship against test results

  5. Physics and Control Assessment of AN 850 Mw(e) - Leu-Candu Reactor.

    Science.gov (United States)

    Barbone, Michelangelo

    The physics and control assessment of an 850 MW(e) Low Enriched Uranium CANDU Pressurized Heavy Water (LEU -CANDU-PHW) reactor constitute the major objective of this thesis. The use of Low Enriched Uranium fuel in the present CANDU nuclear power generating stations is recognized as economically beneficial due to reduced fuelling costs. The LEU fuel cycle is also recognized as a stepping stone to transit from the present CANDU-PHW once-through natural Uranium cycle to advanced cycles such as those based on Plutonium recycle, once-through Th + U-235 cycle, Thorium with Uranium recycle and net U-235 feed, Thorium with Uranium recycle and Plutonium feed. However, although the use of Low Enriched Uranium in the present CANDU-PHW reactor has economic advantages, and it would act as a technical bridge between the present cycle and advanced cycles, technical problems in different areas of reactor physics and fuel management were anticipated. The present thesis research work adresses the areas of reactor physics, fuel management, and control (in particular, the spatial control of large CANDU-PHW reactors). The main conclusions that have been drawn following these studies are as follows: (1) The Low Enriched Uranium Cycle is feasible in a CANDU-PHW reactor of present design and provided that: (a) The enrichment is kept relatively low (that is, about 1% instead of 0.711%); (b) the number of bundles to be replaced at every refuelling operation is about one-half that of the natural Uranium fuel case; (c) The channels are refuelled in the same direction as the coolant. (2) The response of an LEU-CANDU-PHW reactor to reactivity perturbation such as single- and two-channel refuelling operation, shim transient, shutdown-start-up transient with enrichment levels of 0.9% and 1.2% is essentially very similar {provided that certain conditions in (1) are respected} to that of the natural uranium reactor core case without any reactor reoptimization. The general behaviour of the reactor

  6. Enhancing the seismic capability of the on-power refueling system of the CANDU reactor

    International Nuclear Information System (INIS)

    The CANDU reactor assembly includes several hundred horizontal fuel channels, each containing twelve fuel bundles, arranged in a square lattice, and supported by the reactor structures. CANDU operates on natural uranium or other low fissile content fuel, and is refueled on-power, with either four or eight fuel bundles in a channel being replaced during each refueling operation. The fueling machines clamp onto the opposite ends of the fuel channel to be refueled. The seismic capacity of this refueling system is evaluated in terms of its dynamic response during an earthquake. This paper describes the approach adopted to enhance the seismic capability of the fueling machine and calandria assembly for earthquakes of O.3g ground acceleration covering a broad range of soil conditions ranging from soft to hard. A detailed, 3-D finite element seismic model of the fueling machine and calandria assembly system is developed to calculate the seismic responses of the structure. Some relatively simple hardware design changes have been considered to increase the seismic capacity of the CANDU 6 reactor. These changes in the fueling machine and calandria assembly of the CANDU 6 reactor are briefly described. They have been incorporated into the finite element seismic model of the system. Most of these design changes have already been considered and implemented in other CANDU reactor projects. The current CANDU 6 reactor design fully meets the requirements of seismic qualification for sites with potential for O.2g ground acceleration where the seismic loads need to be combined with the other design loads for the support and pressure boundary components to demonstrate compliance with the applicable Code requirements. In the present study it is demonstrated that, with relatively simple hardware changes, the fueling machine and calandria assembly of the CANDU 6 reactor can withstand earthquakes of O.3g ground acceleration. Based on the current study and some preliminary analysis of the

  7. Assessment of In-Core Damage for Feeder Stagnation Break in CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Jong Yeob; Park, Joo Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-05-15

    A feeder break is a single channel accident while the other channels remain intact in the CANDU core. For some ranges of feeder break size, a flow in the channel can become stagnate due to a force balance between the upstream and the downstream ends. In the extreme, this can lead to a rapid fuel heat up and fuel damage, and the failure of a fuel channel. This break scenario is called a feeder stagnation break. Following the feeder stagnation break, the fuel and pressure tube in the affected channel heat up quickly. The channel fails due to overheating and the channel contents begin to discharge into the moderator. The discharge is composed of steam, some hydrogen produced by possible metal-water reaction, and solid fuel elements of fuel fragments with molten material. The severity of the transient is primarily determined by the amount of molten material discharged into the moderator, and by the interaction between the molten material and the moderator, which determines the rate of energy release. After a channel rupture (pressure tube and calandria tube) some SOR (Shut-Off Rod) guide tubes, which are located in the vicinity of the break in the core, may be damaged. If the damage to the guide tube is substantial, some SORs may not be able to descend into the moderator, and therefore, not contribute to the shut down of the reactor. The increase in system reactivity, due to factors such as poison dilution from discharging coolant and void formation, may challenge the reactivity worth of the available undamaged SORs. Therefore, an analysis of the reactivity worth of the partially impaired SDS 1 (Shut-Down System 1) is required to determine that it can compensate for the increase in reactivity and shut down the reactor. In this study, the hydrodynamic transient, due to the dispersed molten material and the discharged steam, was calculated following the feeder stagnation break. The timing of the channel failure and the mass of the molten material were provided from the

  8. ACR-1000TM - advanced Candu reactor design

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) has developed the Advanced CANDU ReactorTM- 1000 (ACR-1000TM) as an evolutionary advancement of the current CANDU 6TM reactor. This evolutionary advancement is based on AECL's in-depth knowledge of CANDU structures, systems, components and materials, gained during 50 years of continuous construction, engineering and commissioning, as well as on the experience and feedback received from operators of CANDU plants. The ACR design retains the proven strengths and features of CANDU reactors, while incorporating innovations and state-of-the-art technology. These innovations improve economics, inherent safety characteristics, and performance, while retaining the proven benefits of the CANDU family of nuclear power plants. The Canadian nuclear reactor design evolution that has reached today's stage represented by the ACR-1000, has a long history dating back to the early 1950's. In this regard, Canada is in a unique situation, shared only by a very few other countries, where original nuclear power technology has been invented and further developed. The ACR design has been reviewed by domestic and international regulatory bodies, and has been given a positive regulatory opinion about its licensability. The Canadian regulator, the Canadian Nuclear Safety Commission (CNSC) completed the Phase 1 and Phase 2 pre-project design reviews in December 2008 and August 2009, respectively, and concluded that there are no fundamental barriers to licensing the ACR-1000 design in Canada. The final stage of the ACR-1000 design is currently underway and will be completed by fall of 2011, along with the final elements of the safety analyses and probabilistic safety analyses supporting the finalized design. The generic Preliminary Safety Analysis Report (PSAR) for the ACR-1000 was completed in September 2009. The PSAR demonstrates ACR-1000 safety case and compliance with Canadian and international regulatory requirements and expectations. (authors)

  9. Neutronic parameters calculations of a CANDU reactor

    International Nuclear Information System (INIS)

    Neutronic calculations that reproduce in a simplified way some aspects of a CANDU reactor design were performed. Starting from some prefixed reactor parameters, cylindrical and uniform iron adjuster rods were designed. An appropriate refueling scheme was established, defininig in a 2 zones model their dimensions and exit burnups. The calculations have been done using the codes WIMS-D4 (cell), SNOD (reactivity device simulations) and PUMA (reactor). Comparing with similar calculations done with codes and models usually employed for CANDU design, it is concluded that the models and methods used are appropriate. (Author)

  10. Instrumented thick-walled tube method for measuring thermal pressure in fluids and isotropic stresses in thermosetting resins

    Science.gov (United States)

    Merzlyakov, Mikhail; Simon, Sindee L.; McKenna, Gregory B.

    2005-06-01

    We have developed a method for measuring the thermal pressure coefficient and cure-induced and thermally induced stresses based on an instrumented thick-walled tube vessel. The device has been demonstrated at pressures up to 330 MPa and temperatures to 300 °C. The method uses a sealed stainless steel thick-walled tube to impose three-dimensional isotropic constraints. The tube is instrumented with strain gauges in hoop and in axial directions and can be used in open or closed configurations. By making measurements of the isotropic stresses as a function of temperature, the method allows determination of the thermal pressure coefficient in both the glassy and rubbery (or liquid) states. The method also can be used to measure isotropic stress development in thermosetting resins during cure and subsequent thermal cycling. Experimental results are presented for sucrose benzoate, di-2-ethylhexylsebacate, and an epoxy resin. The current report shows that the method provides reliable estimates for the thermal pressure coefficient. The thermal pressure coefficient is determined with resolution on the order of 10kPa/K. Among advantages of the method is that the tubes are reusable, even when measurements are made for cure response of thermosetting resins.

  11. Estimation of the hot extrusion process pressure cycle of zircaloy tubes by torsion and compression tests

    International Nuclear Information System (INIS)

    In the production of Zircaloy-4 tubes for nuclear reactors, the first semi-processed tubular form is obtained using the extrusion process. Empirical equations are normally used, which can be applied to extrusion with axial symmetry, or analytical ones are used such as Seibel's equation to evaluate the extrusion process based on the material flow tension. When we use the flow tension corresponding to the mean value of the velocity of extrusion deformation, the extrusion pressure is significantly underestimated, with relation to the experimentally measured pressure. This is because of the flow tension's heavy dependence on the velocity of deformation, which is typical of commercial zirconium alloys. Therefore, the pressure was estimated by calculating the power dissipated during the deformation assuming a velocity field of homogenous deformation in each stage of deformation but without considering friction forces between the work and the extrusion matrix. The flow tension for the torsion tests performed are compared with the results obtained by compression as reported in the literature. These results are compared with four extrusion sequences carried out with different: reduction rates, temperatures, and deformation velocities. The flow tension from the compression test presents greater tension values than those estimated by the torsion test. The origin of these differences is discussed and the conclusion is that they can be attributed to the different crystallography textures generated in both tests. Once the correction is made for the texture variation, the flow tension values evaluated with both testing types in samples of Zircaloy-4 are the same. The peculiarities of each test in relation to the extrusion process are discussed. Despite the very simplified hypotheses that were assumed, the extrusion pressures calculated with the compression and torsion flow tension results, considering their dependence on the speed of deformation and temperature variation during

  12. The CANTEACH project: preserving CANDU technical knowledge

    International Nuclear Information System (INIS)

    Almost sixty years have passed since the nuclear energy venture began in Canada. Fifty years have passed since the founding of AECL. Tens of thousands of dedicated people have forged a new and successful primary energy supply. CANDU technology is well into its second century. This specialty within the world's fission technology community is quite unique, first because it was established as a separate effort very early in the history of world fission energy, and second because it grew in an isolated environment, with tight security requirements, in its early years. Commercial security rules later sustained a considerable degree of isolation. The pioneers of CANDU development have finished their work. Most of the second generation also has moved on. As yet, we cannot point to a consistent and complete record of this remarkable achievement. We, as a nuclear enterprise, have not captured the design legacy in a form that is readily accessible to the current and future generation of professionals involved with CANDU reactors, be they students, designers, operations staff, regulators, consultants or clients. This is a serious failure. Young people entering our field of study must make do with one or two textbooks and a huge collection of diverse technical papers augmented by limited-scope education and training materials. Those employed in the various parts of the nuclear industry rely mostly on a smaller set of CANDU- related documents available within their own organization; documents that sometimes are rather limited in scope. University professors often have even more limited access to in-depth and up to date information. In fact, they often depend on literature published in other countries when preparing lectures, enhanced by guest lecturers from various parts of the industry. Because CANDU was developed mostly inside Canada, few of these text materials contain useful data describing processes important to the CANDU system. For many years it has been recognized that

  13. Preliminary fracture analysis of the core pressure boundary tube for the Advanced Neutron Source Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Schulz, K.C. [Univ. of Turabo, Gurabo, Puerto (Puerto Rico). College of Engineering; Yahr, G.T. [Oak Ridge National Lab., TN (United States)

    1995-08-01

    The outer core pressure boundary tube (CPBT) of the Advanced neutron Source (ANS) reactor being designed at Oak Ridge National Laboratory is currently specified as being composed of 6061-T6 aluminum. ASME Boiler and Pressure Vessel Code fracture analysis rules for nuclear components are based on the use of ferritic steels; the expressions, tables, charts and equations were all developed from tests and analyses conducted for ferritic steels. Because of the nature of the Code, design with thin aluminum requires analytical approaches that do not directly follow the Code. The intent of this report is to present a methodology comparable to the ASME Code for ensuring the prevention of nonductile fracture of the CPBT in the ANS reactor. 6061-T6 aluminum is known to be a relatively brittle material; the linear elastic fracture mechanics (LEFM) approach is utilized to determine allowable flaw sizes for the CPBT. A J-analysis following the procedure developed by the Electric Power Research Institute was conducted as a check; the results matched those for the LEFM analysis for the cases analyzed. Since 6061-T6 is known to embrittle when irradiated, the reduction in K{sub Q} due to irradiation is considered in the analysis. In anticipation of probable requirements regarding maximum allowable flaw size, a survey of nondestructive inspection capabilities is also presented. A discussion of probabilistic fracture mechanics approaches, principally Monte Carlo techniques, is included in this report as an introduction to what quantifying the probability of nonductile failure of the CPBT may entail.

  14. Strategies for obtaining long constant-pressure test times in shock tubes

    Science.gov (United States)

    Campbell, M. F.; Parise, T.; Tulgestke, A. M.; Spearrin, R. M.; Davidson, D. F.; Hanson, R. K.

    2015-11-01

    Several techniques have been developed for obtaining long, constant-pressure test times in reflected shock wave experiments in a shock tube, including the use of driver inserts, driver gas tailoring, helium gas diaphragm interfaces, driver extensions, and staged driver gas filling. These techniques are detailed here, including discussion on the most recent strategy, staged driver gas filling. Experiments indicate that this staged filling strategy increases available test time by roughly 20 % relative to single-stage filling of tailored driver gas mixtures, while simultaneously reducing the helium required per shock by up to 85 %. This filling scheme involves firstly mixing a tailored helium-nitrogen mixture in the driver section as in conventional driver filling and, secondly, backfilling a low-speed-of-sound gas such as nitrogen or carbon dioxide from a port close to the end cap of the driver section. Using this staged driver gas filling, in addition to the other techniques listed above, post-reflected shock test times of up to 0.102 s (102 ms) at 524 K and 1.6 atm have been obtained. Spectroscopically based temperature measurements in non-reactive mixtures have confirmed that temperature and pressure conditions remain constant throughout the length of these long test duration trials. Finally, these strategies have been used to measure low-temperature n-heptane ignition delay times.

  15. Flow condensation pressure drop characteristics of R410A-oil mixture inside small diameter horizontal microfin tubes

    Energy Technology Data Exchange (ETDEWEB)

    Huang, Xiangchao; Ding, Guoliang; Hu, Haitao; Zhu, Yu. [Institute of Refrigeration and Cryogenics, Shanghai Jiaotong University, Shanghai 200240 (China); Gao, Yifeng [International Copper Association Shanghai Office, Shanghai 200020 (China); Deng, Bin [Institute of Heat Transfer Technology, Golden Dragon Precise Copper Tube Group Inc., Shanghai 200135 (China)

    2010-11-15

    Flow condensation pressure drop characteristics of R410A-oil mixture inside small diameter (5.0 mm and 4.0 mm O.D.) horizontal microfin tubes were investigated experimentally covering nominal oil concentrations from 0% to 5%. The research results indicate that, comparing with the frictional pressure drop of pure R410A, the frictional pressure drop of R410A-oil mixture may decrease by maximum of 18% when the vapor quality is lower than 0.6, and increase by maximum of 13% when the vapor quality is higher than 0.6. A new frictional pressure drop correlation for R410A-oil mixture flow condensation inside microfin tubes is developed based on the refrigerant-oil mixture properties, and can agree with 94% of the experimental data within a deviation of -30% to +30%. (author)

  16. Proper Orthogonal Decomposition of Pressure Fields in a Draft Tube Cone of the Francis (Tokke) Turbine Model

    International Nuclear Information System (INIS)

    The simulations of high head Francis turbine model (Tokke) are performed for three operating conditions - Part Load, Best Efficiency Point (BEP) and Full Load using software Ansys Fluent R15 and alternatively OpenFOAM 2.2.2. For both solvers the simulations employ Realizable k-e turbulence model. The unsteady pressure pulsations of pressure signal from two monitoring points situated in the draft tube cone and one behind the guide vanes are evaluated for all three operating conditions in order to compare frequencies and amplitudes with the experimental results. The computed velocity fields are compared with the experimental ones using LDA measurements in two locations situated in the draft tube cone. The proper orthogonal decomposition (POD) is applied on a longitudinal slice through the draft tube cone. The unsteady static pressure fields are decomposed and a spatio-temporal behavior of modes is correlated with amplitude-frequency results obtained from the pressure signal in monitoring points. The main application of POD is to describe which modes are related to an interaction between rotor (turbine runner) and stator (spiral casing and guide vanes) and cause dynamic flow behavior in the draft tube. The numerically computed efficiency is correlated with the experimental one in order to verify the simulation accuracy

  17. Proper Orthogonal Decomposition of Pressure Fields in a Draft Tube Cone of the Francis (Tokke) Turbine Model

    Science.gov (United States)

    Stefan, D.; Rudolf, P.

    2015-01-01

    The simulations of high head Francis turbine model (Tokke) are performed for three operating conditions - Part Load, Best Efficiency Point (BEP) and Full Load using software Ansys Fluent R15 and alternatively OpenFOAM 2.2.2. For both solvers the simulations employ Realizable k-e turbulence model. The unsteady pressure pulsations of pressure signal from two monitoring points situated in the draft tube cone and one behind the guide vanes are evaluated for all three operating conditions in order to compare frequencies and amplitudes with the experimental results. The computed velocity fields are compared with the experimental ones using LDA measurements in two locations situated in the draft tube cone. The proper orthogonal decomposition (POD) is applied on a longitudinal slice through the draft tube cone. The unsteady static pressure fields are decomposed and a spatio-temporal behavior of modes is correlated with amplitude-frequency results obtained from the pressure signal in monitoring points. The main application of POD is to describe which modes are related to an interaction between rotor (turbine runner) and stator (spiral casing and guide vanes) and cause dynamic flow behavior in the draft tube. The numerically computed efficiency is correlated with the experimental one in order to verify the simulation accuracy.

  18. Temperature and pressure measurements at cold exit of counter-flow vortex tube with flow visualization of reversed flow

    Science.gov (United States)

    Yusof, Mohd Hazwan bin; Katanoda, Hiroshi; Morita, Hiromitsu

    2015-02-01

    In order to clarify the structure of the cold flow discharged from the counter-flow vortex tube (VT), the temperature and pressure of the cold flow were measured, and the existence and behavior of the reversed flow at the cold exit was studied using a simple flow visualization technique consisting of a 0.75mm-diameter needle, and an oil paint droplet. It is observed through this experiment that the Pitot pressure at the cold exit center can either be lower or higher than atmospheric pressure, depending on the inlet pressure and the cold fraction, and that a reversed flow is observed when the Pitot pressure at the cold exit center is lower than atmospheric pressure. In addition, it is observed that when reducing the cold fraction from unity at any arbitrary inlet pressure, the region of reversed and colder flow in the central part of cold exit extends in the downstream direction.

  19. CONSIDERATIONS FOR THE DEVELOPMENT OF A DEVICE FOR THE DECOMMISSIONING OF THE HORIZONTAL FUEL CHANNELS IN THE CANDU 6 NUCLEAR REACTOR. PART 10 - PRESENTATION OF THE DECOMMISSIONING DEVICE OPERATING

    Directory of Open Access Journals (Sweden)

    Constantin D. STANESCU,

    2015-05-01

    Full Text Available This paper presents a solution proposed by the authors in order to achieve of a cutting and extracting device operating panel for the decommissioning of the horizontal fuel channels in the CANDU 6 nuclear reactor. The Cutting and Extraction Device (CED is fully automated, connected by wires to a Programmable Logic Controller (PLC and controlled from a Human Machine Interface (HMI. The Cutting and Extraction Device (CED performs the dismantling, cutting and extraction of the fuel channel components, moving with variable speed, temperature monitoring and video surveillance inside the pipe, unblock and extract the channel closure plug (from End Fitting - EF, unblock and extract the channel shield plug (from Lattice Tube - LT, block and cut the middle of the pressure tube, block and cut the end of the pressure tube, block and extract the half of pressure tube. All operations can be monitored and controlled from a operating panel. The PLC fully command the device in automatic or manually mode, to control the internal sensors, transducers, electrical motors, video surveillance and pyrometers for monitoring cutting place temperature. The device controller has direct access to the measured values with these sensors, interprets and processes them, preparing the next actionafter confirming the action in progress. The design of the Cutting and Extraction Device (CED shall be achieved according to the particular features of the fuel channel components to be dismantled and to ensure radiation protection of workers.

  20. Post Irradiation Examination of Experomental CANDU Fuel Elements Irradiated in TRIGA-SSR Reactor

    International Nuclear Information System (INIS)

    The object of this work is the behaviour of CANDU fuel elements under power cycling conditions. The tests were run in the 14 MW (th) TRIGA-SSR (Steady State Reactor) reactor from Institute for Nuclear Research (INR) Pitesti. Zircaloy-4 is the material used for CANDU fuel sheath. The importance of studying its behaviour results from the fact that the mechanical properties of the CANDU fuel sheath suffer modifications during normal and abnormal operation. In the nuclear reactor the fuel elements endure dimensional and structural changes as well as cladding oxidation, hydriding and corrosion. These changes can lead to defects and even to the loss of integrity of the cladding. This paper presents the results of examinations performed in the Post Irradiation Examination Laboratory (PIEL) from INR Pitesti, on samples from a fuel element irradiated in TRIGA-SSR reactor: (i) Dimensional and macrostructural characterization; (ii) Gamma scanning and tomography; (iii) Measurement of pressure, volume and isotopic composition of fission gas; (iv) Microstructural characterization by metallographic analyses; (v) Determination of mechanical properties; amd (vi) Fracture surface analysis by scanning electron microscopy (SEM). The obtained data could be used to evaluate the security, reliability and nuclear fuel performance, and for CANDU fuel improvement. (author)

  1. Exporting apocalypse: CANDU reactors and nuclear proliferation

    International Nuclear Information System (INIS)

    The author believes that the peaceful use of nuclear technology leads inevitably to the production of nuclear weapons, and that CANDU reactors are being bought by countries that are likely to build bombs. He states that exports of reactors and nuclear materials cannot be defended and must be stopped

  2. Development situation about the Canadian CANDU Nuclear Power Generating Stations

    International Nuclear Information System (INIS)

    The CANDU reactor is the most versatile commercial power reactor in the world. The acronym 'CANDU', a registered trademark of Atomic Energy of Canada Limited, stands for 'CANada Deuterium Uranium'. CANDU uses heavy water as moderator and uranium (originally, natural uranium) as fuel. All current power reactors in Canada are of the CANDU type. Canada exports CANDU type reactor in abroad. CANDU type is used as the nuclear power plants to produce electrical. Today, there are 41 CANDU reactors in use around the world, and the design has continuously evolved to maintain into unique technology and performance. The CANDU-6 power reactor offers a combination of proven, superior and state-of-the-art technology. CANDU-6 was designed specifically for electricity production, unlike other major reactor types. One of its characteristics is a very high operating and fuel efficiency. Canada Nuclear Power Generating Stations were succeeded in a commercial reactor of which the successful application of heavy water reactor, natural uranium method and that on-power fuelling could be achieved. It was achieved through the joint development of a major project by strong support of the federal government, public utilities and private enterprises. The potential for customization to any country's needs, with competitive development and within any level of domestic industrial infrastructure, gives CANDU technology strategic importance in the 21st century

  3. Candu reactors with thorium fuel cycles

    International Nuclear Information System (INIS)

    Over the last decade and a half AECL has established a strong record of delivering CANDU 6 nuclear power plants on time and at budget. Inherently flexible features of the CANDU type reactors, such as on-power fuelling, high neutron economy, fuel channel based heat transport system, simple fuel bundle configuration, two independent shut down systems, a cool moderator and a defence-in-depth based safety philosophy provides an evolutionary path to further improvements in design. The immediate milestone on this path is the Advanced CANDU ReactorTM** (ACRTM**), in the form of the ACR-1000TM**. This effort is being followed by the Super Critical Water Reactor (SCWR) design that will allow water-cooled reactors to attain high efficiencies by increasing the coolant temperature above 5500C. Adaptability of the CANDU design to different fuel cycles is another technology advantage that offers an additional avenue for design evolution. Thorium is one of the potential fuels for future reactors due to relative abundance, neutronics advantage as a fertile material in thermal reactors and proliferation resistance. The Thorium fuel cycle is also of interest to China, India, and Turkey due to local abundance that can ensure sustainable energy independence over the long term. AECL has performed an assessment of both CANDU 6 and ACR-1000 designs to identify systems, components, safety features and operational processes that may need to be modified to replace the NU or SEU fuel cycles with one based on Thorium. The paper reviews some of these requirements and the associated practical design solutions. These modifications can either be incorporated into the design prior to construction or, for currently operational reactors, during a refurbishment outage. In parallel with reactor modifications, various Thorium fuel cycles, either based on mixed bundles (homogeneous) or mixed channels (heterogeneous) have been assessed for technical and economic viability. Potential applications of a

  4. Using Advanced Fuel Bundles in CANDU Reactors

    International Nuclear Information System (INIS)

    Improving the exit fuel burnup in CANDU reactors was a long-time challenge for both bundle designers and performance analysts. Therefore, the 43-element design together with several fuel compositions was studied, in the aim of assessing new reliable, economic and proliferation-resistant solutions. Recovered Uranium (RU) fuel is intended to be used in CANDU reactors, given the important amount of slightly enriched Uranium (~0.96% w/o U235) that might be provided by the spent LWR fuel recovery plants. Though this fuel has a far too small U235 enrichment to be used in LWR's, it can be still used to fuel CANDU reactors. Plutonium based mixtures are also considered, with both natural and depleted Uranium, either for peacefully using the military grade dispositioned Plutonium or for better using Plutonium from LWR reprocessing plants. The proposed Thorium-LEU mixtures are intended to reduce the Uranium consumption per produced MW. The positive void reactivity is a major concern of any CANDU safety assessment, therefore reducing it was also a task for the present analysis. Using the 43-element bundle with a certain amount of burnable poison (e.g. Dysprosium) dissolved in the 8 innermost elements may lead to significantly reducing the void reactivity. The expected outcomes of these design improvements are: higher exit burnup, smooth/uniform radial bundle power distribution and reduced void reactivity. Since the improved fuel bundles are intended to be loaded in existing CANDU reactors, we found interesting to estimate the local reactivity effects of a mechanical control absorber (MCA) on the surrounding fuel cells. Cell parameters and neutron flux distributions, as well as macroscopic cross-sections were estimated using the transport code DRAGON and a 172-group updated nuclear data library. (author)

  5. Thermochemical modelling of advanced CANDU reactor fuel

    Science.gov (United States)

    Corcoran, Emily Catherine

    2009-04-01

    With an aging fleet of nuclear generating facilities, the imperative to limit the use of non-renewal fossil fuels and the inevitable need for additional electricity to power Canada's economy, a renaissance in the use of nuclear technology in Canada is at hand. The experience and knowledge of over 40 years of CANDU research, development and operation in Ontario and elsewhere has been applied to a new generation of CANDU, the Advanced CANDU Reactor (ACR). Improved fuel design allows for an extended burnup, which is a significant improvement, enhancing the safety and the economies of the ACR. The use of a Burnable Neutron Absorber (BNA) material and Low Enriched Uranium (LEU) fuel has created a need to understand better these novel materials and fuel types. This thesis documents a work to advance the scientific and technological knowledge of the ACR fuel design with respect to thermodynamic phase stability and fuel oxidation modelling. For the BNA material, a new (BNA) model is created based on the fundamental first principles of Gibbs energy minimization applied to material phase stability. For LEU fuel, the methodology used for the BNA model is applied to the oxidation of irradiated fuel. The pertinent knowledge base for uranium, oxygen and the major fission products is reviewed, updated and integrated to create a model that is applicable to current and future CANDU fuel designs. As part of this thesis, X-Ray Diffraction (XRD) and Coulombic Titration (CT) experiments are compared to the BNA and LEU models, respectively. From the analysis of the CT results, a number of improvements are proposed to enhance the LEU model and provide confidence in its application to ACR fuel. A number of applications for the potential use of these models are proposed and discussed. Keywords: CANDU Fuel, Gibbs Energy Mimimization, Low Enriched Uranium (LEU) Fuel, Burnable Neutron Absorber (BNA) Material, Coulometric Titration, X-Ray Diffraction

  6. EXPERIMENTAL INVESTIGATION ON HEAT TRANSFER AND PRESSURE DROP CHARACTERISTICS OF AIR FLOW OVER A STAGGERED FLAT TUBE BANK IN CROSSFLOW

    Directory of Open Access Journals (Sweden)

    M. Ishak

    2013-06-01

    Full Text Available This paper presents an experimental investigation into the heat transfer and pressure drop characteristics of air flow in a staggered flat tube bank in crossflow with laminar-forced convection. Measurements were conducted for sixteen tubes in the direction of flow and four tubes in rows. The air velocity varies between 0.6–1.0 m/s and the Reynolds number varied from 373 to 623. The total heat flux supplied in all tubes are changed from 967.92 to 3629.70 W/m2. The results indicate that the average Nusselt number for all the flat tubes increased by 11.46–46.42%, with the Reynolds numbers varying from 373 to 623 at the fixed heat flux. The average Nusselt number increased by 21.39–84%, and the total heat flux varyied between 967.92–3629.70 W/m2 with a constant Reynolds number Re = 498. In addition, the pressure drop decreased with an increase in the Reynolds number. A new mean Nusselt number-Reynolds number correlation was found, and the correlation yielded good predictions for the measured data with a confidence interval of 98.9%.

  7. Validation of the ASSERT subchannel code: Prediction of critical heat flux in standard and nonstandard CANDU bundle geometries

    International Nuclear Information System (INIS)

    The ASSERT code has been developed to address the three-dimensional computation of flow and phase distribution and fuel element surface temperatures within the horizontal subchannels of Canada uranium deuterium (CANDU) pressurized heavy water reactor fuel channels and to provide a detailed prediction of critical heat flux (CHF) distribution throughout the bundle. The ASSERT subchannel code has been validated extensively against a wide repertoire of experiments; its combination of three-dimensional prediction of local flow conditions with a comprehensive method of predicting CHF at these local conditions makes it a unique tool for predicting CHF for situations outside the existing experimental database. In particular, ASSERT is an appropriate tool to systematically investigate CHF under conditions of local geometric variations, such as pressure tube creep and fuel element strain. The numerical methodology used in ASSERT, the constitutive relationships incorporated, and the CHF assessment methodology are discussed. The evolutionary validation plan is also discussed and early validation exercises are summarized. More recent validation exercises in standard and nonstandard geometries are emphasized

  8. Evaluation of candidate Stirling engine heater tube alloys after 3500 hours exposure to high pressure doped hydrogen or helium

    Science.gov (United States)

    Misencik, J. A.; Titran, R. H.

    1984-01-01

    The heater head tubes of current prototype automotive Stirling engines are fabricated from alloy N-155, an alloy which contains 20 percent cobalt. Because the United States imports over 90 percent of the cobalt used in this country and resource supplies could not meet the demand imposed by automotive applications of cobalt in the heater head (tubes plus cylinders and regenerator housings), it is imperative that substitute alloys free of cobalt be identified. The research described herein focused on the heater head tubes. Sixteen alloys (15 potential substitutes plus the 20 percent Co N-155 alloy) were evaluated in the form of thin wall tubing in the NASA Lewis Research Center Stirling simulator materials diesel fuel fired test rigs. Tubes filled with either hydrogen doped with 1 percent CO2 or with helium at a gas pressure of 15 MPa and a temperature of 820 C were cyclic endurance tested for times up to 3500 hr. Results showed that two iron-nickel base superalloys, CG-27 and Pyromet 901 survived the 3500 hr endurance test. The remaining alloys failed by creep-rupture at times less than 3000 hr, however, several other alloys had superior lives to N-155. Results further showed that doping the hydrogen working fluid with 1 vol % CO2 is an effective means of reducing hydrogen permeability through all the alloy tubes investigated.

  9. Influence of the wetting state of a heated surface on heat transfer and pressure loss in an evaporator tube

    Energy Technology Data Exchange (ETDEWEB)

    Koehler, W; Hein, D

    1986-09-01

    The influence of the wetting state of a heated surface on heat transfer and pressure loss in an evaporator tube was investigated for a parameter range occurring in fossil-fired steam generators. Included in the analysis are quantities which determine the wetting state in steady and transient flow. The experimental work consists of the following: Occurrence of critical heat flux (CHF) and post-CHF heat transfer in a vertical upflow evaporator tube; influence of pressure and enthalpy transients on heat transfer in the unwetted region; influence of pipe orientation on heat transfer; and two phase flow pressure loss in wetted and unwetted region. Based on these experiments a method of predicting CHF for a vertical upflow evaporator tube was developed. The heat transfer in the unwetted region was newly formulated taking into account thermal nonequilibrium between the water and steam phases. Wall temperature excursions during pressure and enthalpy transients are interpreted with the help of the boiling curve and the Leidenfrost phenomenon. A method is developed by means of which it is possible to determine the influence of the pipe orientation on the location of the boiling crisis as well as on the heat transfer in the unwetted region. The influence of the wetting state of the heated surface on the two phase flow pressure loss is interpreted as ''Wall effect'' and is calculated using a simplified computer model. 68 refs., 83 figs.

  10. Evaluation of oxides formed at high temperatures in Zr-2.5Nb pressure tubing

    International Nuclear Information System (INIS)

    The oxidation behavior of Zr-2.5Nb pressure tube samples has been studied at four different temperatures, i.e., 400°, 600°, 800°, and 1000°C. The amount of tetragonal phase is found to decrease with increase of temperature. The oxide texture of (002)m and (111)m type increased with the temperature from 400°C to 600°C, however at temperatures above 600°C the texture strength seems to diminish and the oxide layer becomes structurally unstable. Further, the impedance response is found to be dependent on the microstructure of the oxide film. For the sample oxidized at 400°C, Electrochemical Impedance Spectroscopy (EIS) spectra exhibited a two-time constant behavior, showing the formation of two-layer oxide film on the Zr-2.5Nb alloy, which correspond to a porous outer oxide and a barrier inner oxide, respectively. In addition, the samples were oxidized at constant temperature of 600°C with varying oxidation time. The observation shows that the oxide is more protective in the early stage of oxide growth. However, further growth of oxide film has resulted in degeneration of its protective character. (author)

  11. Heavy ion irradiation effects in Zr excel alloy pressure tube material

    International Nuclear Information System (INIS)

    Zirconium Excel alloy (Zr-3.5wt.%Sn-0.8%Nb-0.8%Mo) is the candidate material for pressure tubes in the Generation-IV CANDU® Super Critical Water-cooled Reactor (SCWR) design. Changes in microstructure induced by neutron irradiation are known to have important consequences on the in-reactor deformation behavior. The in-situ ion irradiation technique has been employed to elucidate the irradiation damage in dual phase Zr-excel alloy (~60% hcp alpha and ~40% bcc beta). 1 MeV Kr ion irradiation experiments were conducted at different temperatures ranging from 100oC-400oC. Damage microstructures have been characterized by Transmission Electron Microscopy in both the alpha and beta phases at different temperatures after a maximum dose of 10 dpa. Several new observations including irradiation induced omega (ω) phase precipitation have been reported. The ω/β orientation relationship was determined by the detailed analysis of selected area diffraction patterns. In-situ irradiation provided an opportunity to observe the nucleation and growth of basal plane c-component loops. It has been shown that under Kr ion irradiation the c-loops start to nucleate and grow above a threshold dose, as has been observed for neutron irradiation. Furthermore, the role of temperature, material composition and pre-irradiation microstructure has been discussed in detail. (author)

  12. Characteristics of two-phase flow pattern transitions and pressure drop of five refrigerants in horizontal circular small tubes

    Energy Technology Data Exchange (ETDEWEB)

    Pamitran, A.S. [Department of Mechanical Engineering, University of Indonesia, Kampus Baru UI, Depok 16424 (Indonesia); Choi, Kwang-Il [Graduate School, Chonnam National University, San 96-1, Dunduk-Dong, Yeosu, Chonnam 550-749 (Korea); Oh, Jong-Taek [Department of Refrigeration and Air Conditioning Engineering, Chonnam National University, San 96-1, Dunduk-Dong, Yeosu, Chonnam 550-749 (Korea); Hrnjak, Pega [Department of Mechanical Science and Engineering, ACRC, University of Illinois at Urbana-Champaign, 1206 West Green Street, Urbana, IL 61801 (United States)

    2010-05-15

    An experimental investigation on the characteristics of two-phase flow pattern transitions and pressure drop of R-22, R-134a, R-410A, R-290 and R-744 in horizontal small stainless steel tubes of 0.5, 1.5 and 3.0 mm inner diameters is presented. Experimental data were obtained over a heat flux range of 5-40 kW/m{sup 2}, mass flux range of 50-600 kg/(m{sup 2} s), saturation temperature range of 0-15 C, and quality up to 1.0. Experimental data were evaluated with Wang et al. and Wojtan et al. [Wang, C.C., Chiang, C.S., Lu, D.C., 1997. Visual observation of two-phase flow pattern of R-22, R-134a, and R-407C in a 6.5-mm smooth tube. Exp. Therm. Fluid Sci. 15, 395-405; Wojtan, L., Ursenbacher, T., Thome, J.R., 2005. Investigation of flow boiling in horizontal tubes: part I - a new diabatic two-phase flow pattern map. Int. J. Heat Mass Transfer 48, 2955-2969.] flow pattern maps. The effects of mass flux, heat flux, saturation temperature and inner tube diameter on the pressure drop of the working refrigerants are reported. The experimental pressure drop was compared with the predictions from some existing correlations. A new two-phase pressure drop model that is based on a superposition model for two-phase flow boiling of refrigerants in small tubes is presented. (author)

  13. Creep collapse of thick-walled heat transfer tube subjected to external pressure at high temperature

    International Nuclear Information System (INIS)

    A series of creep collapse tests of thick-walled heat transfer tube were examined experimentally and analytically to confirm an analytical method for creep deformation behavior of a heat transfer tube of an intermediate heat exchanger (IHX) at a depressurization accident of secondary cooling system of HTTR (High Temperature Engineering Test Reactor). The tests were carried out using thick-walled heat transfer tubes made of Hastelloy XR at 950degC in helium gas environment. The predictions of creep collapse time obtained by a general purpose FEM-code ABAQUS were in good agreement with the experimental results. A lot of cracks were observed on the outer surface of the test tubes after the creep collapse. However, the cracks did not pass through the tube wall and, therefore, the leak tightness was maintained regardless of a collapse deformation for all tubes tested. (author)

  14. Technology spin-offs from a CANDU development program

    International Nuclear Information System (INIS)

    Both Enhanced CANDU 6 (EC6) and ACR-1000 design retain many essential features of the operating CANDU 6 plant design. As well as further-enhanced safety, the design also focuses on operability and maintainability, drawing on valuable customer input and OPEX. The engineering development of the ACR-1000 design has been accompanied by a research and confirmatory testing program. The ACR technology developed during the ACR-1000 Basic Engineering Program and the supporting development testing has extended the database of knowledge on the CANDU design. This paper provides a summary of technology arising from the ACR program that has been incorporated into new CANDU designs such as the Enhanced CANDU 6 (EC6), or can be applied for servicing operating CANDU reactors. (author)

  15. Analysis of flow induced density wave oscillations in the CANDU supercritical water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dutta, Goutam, E-mail: gd@iiitdmj.ac.in [Department of Electrical and Computer Engineering, University of Western Ontario, London, Ontario, Canada N6A 5B9 (Canada); Zhang, Chao, E-mail: czhang@eng.uwo.ca [Department of Mechanical and Materials Engineering, University of Western Ontario, London, Ontario, Canada N6A 5B9 (Canada); Jiang, Jin, E-mail: jjiang@eng.uwo.ca [Department of Electrical and Computer Engineering, University of Western Ontario, London, Ontario, Canada N6A 5B9 (Canada)

    2015-05-15

    In this paper, a 1-D thermal-hydraulic model, THRUST, is developed to simulate and analyze the CANDU supercritical water reactor (SCWR) from the thermodynamic point of view without considering the effect of neutronic coupling. THRUST, where a characteristic-based finite difference scheme is used, is validated against the available numerical results. The model is, then, used for the analysis of the CANDU SCWR with a primary focus to determine the conditions for potential density wave oscillations. Extensive numerical studies are performed to obtain the marginal stability boundary in the operating regime of the reactor. The effect of various parameters, such as mass flow rate, operating pressure, axial heat flux profile, local pressure drop coefficient, and friction factor, on the stability thresholds of the reactor have been investigated.

  16. Heat transfer and pressure drop characteristics of plain finned heat exchangers having 5.0 mm tubes

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Nae Hyun; Ham, Jung Ho; Oh, Wang Ku [Incheon Univ., Incheon (Korea, Republic of); Choi, Yong Hwa; Gaku, Hayase [Samsung Electric Company, Suwon (Korea, Republic of)

    2007-07-01

    In this study, pressure drop and heat transfer characteristics of plain finned heat exchangers having 5.0 diameter (fin collar 5.3 mm) tubes were investigated. Six samples having different fin pitches and tube rows were tested. The fin pitch had a negligible effect on j and f factors. Both j and f factors decreased as the number of tube row increased, although the difference was not significant for the f factor. When compared with the previous 7.3 mm diameter data, both the present j and f factors yielded lower values. However, the j/f ratio was larger at low Reynolds numbers. Possible reasoning is provided from the flow pattern consideration. Comparison with existing correlations were made.

  17. Heat transfer and pressure drop characteristics of plain finned heat exchangers having 5.0 mm tubes

    International Nuclear Information System (INIS)

    In this study, pressure drop and heat transfer characteristics of plain finned heat exchangers having 5.0 diameter (fin collar 5.3 mm) tubes were investigated. Six samples having different fin pitches and tube rows were tested. The fin pitch had a negligible effect on j and f factors. Both j and f factors decreased as the number of tube row increased, although the difference was not significant for the f factor. When compared with the previous 7.3 mm diameter data, both the present j and f factors yielded lower values. However, the j/f ratio was larger at low Reynolds numbers. Possible reasoning is provided from the flow pattern consideration. Comparison with existing correlations were made

  18. Validation of WIMS-CANDU using Pin-Cell Lattices

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Won Young; Min, Byung Joo; Park, Joo Hwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2006-07-01

    The WIMS-CANDU is a lattice code which has a depletion capability for the analysis of reactor physics problems related to a design and safety. The WIMS-CANDU code has been developed from the WIMSD5B, a version of the WIMS code released from the OECD/NEA data bank in 1998. The lattice code POWDERPUFS-V (PPV) has been used for the physics design and analysis of a natural uranium fuel for the CANDU reactor. However since the application of PPV is limited to a fresh fuel due to its empirical correlations, the WIMS-AECL code has been developed by AECL to substitute the PPV. Also, the WIMS-CANDU code is being developed to perform the physics analysis of the present operating CANDU reactors as a replacement of PPV. As one of the developing work of WIMS-CANDU, the U{sup 238} absorption cross-section in the nuclear data library of WIMS-CANDU was updated and WIMS-CANDU was validated using the benchmark problems for pin-cell lattices such as TRX-1, TRX-2, Bapl-1, Bapl-2 and Bapl-3. The results by the WIMS-CANDU and the WIMS-AECL were compared with the experimental data.

  19. Reactor physics and economic aspects of the CANDU reactor system

    International Nuclear Information System (INIS)

    A history of the development of the CANDU system is given along with a fairly detailed description of the 600 MW(e) CANDU reactor. Reactor physics calculation methods are described, as well as comparisons between calculated reactor physics parameters and those measured in research and power reactors. An examination of the economics of CANDU in the Ontario Hydro system and a comparison between fossil fuelled and light water reactors is presented. Some physics, economics and resources aspects are given for both low enriched uranium and thorium-fuelled CANDU reactors. Finally the RβD program in Advanced Fuel Cycles is briefly described

  20. Changes in endotracheal tube cuff pressure during laparoscopic surgery in head-up or head-down position

    Science.gov (United States)

    2014-01-01

    Background The abdominal insufflation and surgical positioning in the laparoscopic surgery have been reported to result in an increase of airway pressure. However, associated effects on changes of endotracheal tube cuff pressure are not well established. Methods 70 patients undergoing elective laparoscopic colorectal tumor resection (head-down position, n = 38) and laparoscopic cholecystecomy (head-up position, n = 32) were enrolled and were compared to 15 patients undergoing elective open abdominal surgery. Changes of cuff and airway pressures before and after abdominal insufflation in supine position and after head-down or head-up positioning were analysed and compared. Results There was no significant cuff and airway pressure changes during the first fifteen minutes in open abdominal surgery. After insufflation, the cuff pressure increased from 26 ± 3 to 32 ± 6 and 27 ± 3 to 33 ± 5 cmH2O in patients receiving laparoscopic cholecystecomy and laparoscopic colorectal tumor resection respectively (both p < 0.001). The head-down tilt further increased cuff pressure from 33 ± 5 to 35 ± 5 cmH2O (p < 0.001). There six patients undergoing colorectal tumor resection (18.8%) and eight patients undergoing cholecystecomy (21.1%) had a total increase of cuff pressure more than 10 cm H2O (18.8%). There was no significant correlation between increase of cuff pressure and either the patient's body mass index or the common range of intra-abdominal pressure (10-15 mmHg) used in laparoscopic surgery. Conclusions An increase of endotracheal tube cuff pressure may occur during laparoscopic surgery especially in the head-down position. PMID:25210501

  1. Simultaneous and long-lasting hydrophilization of inner and outer wall surfaces of polytetrafluoroethylene tubes by transferring atmospheric pressure plasmas

    Science.gov (United States)

    Chen, Faze; Song, Jinlong; Huang, Shuai; Xu, Sihao; Xia, Guangqing; Yang, Dezheng; Xu, Wenji; Sun, Jing; Liu, Xin

    2016-09-01

    Plasma hydrophilization is a general method to increase the surface free energy of materials. However, only a few works about plasma modification focus on the hydrophilization of tube inner and outer walls. In this paper, we realize simultaneous and long-lasting plasma hydrophilization on the inner and outer walls of polytetrafluoroethylene (PTFE) tubes by atmospheric pressure plasmas (APPs). Specifically, an Ar atmospheric pressure plasma jet (APPJ) is used to modify the PTFE tube’s outer wall and meanwhile to induce transferred He APP inside the PTFE tube to modify its inner wall surface. The optical emission spectrum (OES) shows that the plasmas contain many chemically active species, which are known as enablers for various applications. Water contact angle (WCA) measurements, x-ray photoelectron spectroscopy (XPS) and atomic force microscopy (AFM) are used to characterize the plasma hydrophilization. Results demonstrate that the wettability of the tube walls are well improved due to the replacement of the surface fluorine by oxygen and the change of surface roughness. The obtained hydrophilicity decreases slowly during more than 180 d aging, indicating a long-lasting hydrophilization. The results presented here clearly demonstrate the great potential of transferring APPs for surface modification of the tube’s inner and outer walls simultaneously.

  2. Dynamic neck development in a polymer tube under internal pressure loading

    DEFF Research Database (Denmark)

    Lindgreen, Britta; Tvergaard, Viggo; Needleman, Alan

    2008-01-01

    and a short wave length imperfection. After some thinning down at the necks, the mode of deformation switches to neck propagation along the circumference of the tube. A case is shown in which the necks have propagated along the entire tube wall, so that network locking in the polymer results in high stiffness...

  3. Experiments on ballooning in pressurized and transiently heated Zircaloy-4 tubes

    International Nuclear Information System (INIS)

    Single-rod burst tests were performed with Atucha I Zircaloy-4 cladding tubes in the REBEKA burst equipment of KfK. The objective was to investigate the ballooning and burst behavior of argentine cladding tubes obtained from NRG, Germany and CONVAR, Argentina. The burst data were compared with those of cladding tubes used in german PWR's. It was found that the burst data e.g. burst temperature, circumferential burst strain and its response to azimuthal temperature differences are identical for the Argentine and German tubing quality. The burst data are in good agreement with those of German PWR-Zircaloy tubes. Thus, the fuel rod behavior codes developed for German PWR's can also be used for the Argentine reactor Atucha I. (orig.)

  4. A review of CANDU plant lifetime management

    International Nuclear Information System (INIS)

    In recent years, plant lifetime management(PLIM) including life extension has become the focus of the nuclear industry worldwide due to a number of factors which have arisen over the past decade : new siting difficulties, imbalance of power supply and demand, and high construction costs. In order to solve the problems, the PLIM program is being developed for the purpose of life extension and improvement of plant availability and safety. This paper describes the current activities and prospects of AECL and CANDU utilities, the conceptional evaluation results for the degradation mechanisms, and PLIM regulatory aspects. In addition, this paper provides the applicability of CANDU PLIM to Wolsong Unit 1 which has been operated for 17 years

  5. CANDU 9 design for hydrogen in containment

    International Nuclear Information System (INIS)

    CANDU 9 is a single unit plant whose design is based on proven CANDU technology with some design improvements in plant performance and safety. One improvement is in the area of post-accident hydrogen control. The reactor building layout and hydrogen control system are designed to enhance atmospheric mixing and prevent unacceptably high local and global hydrogen concentrations. Preliminary safety analysis shows that a maximum of 300 kg of hydrogen is produced during the metal-water reaction phase of severe accident: a large LOCA with emergency core cooling unavailable. Even though the hydrogen production rate is conservatively overestimated, preliminary containment thermalhydraulic analysis predicts that the maximum hydrogen concentration in the accident vault peaks at 6.8% by volume without crediting hydrogen igniters and recombiners. After about one hour, the concentration throughout the reactor building is about 2.4%. (author)

  6. Measurement of High Temperature Anisotropic Elastic Constants of Zr-2.5Nb Pressure Tube Materials by Resonant Ultrasound Spectroscopy

    International Nuclear Information System (INIS)

    Anisotropic elastic constants of Zr-2.5Nb pressure tube materials were determined by a high temperature resonant ultrasound spectroscopy (RUS). The resonant frequencies were measured using alumina wave-guides and wide band ultrasonic transducers in a small furnace. The rectangular parallelepiped specimens were fabricated along with the axial, radial and circumferential direction of the pressure tube. A nine elastic stiffness tensor for orthotropic symmetry was determined in the range of room temperature ∼500 .deg. C. As the temperature increases, the elastic constant tensor, cij gradually decreases. Higher elastic constants along the transverse direction compared to those along the axial or radial direction are similar to the case of Young's modulus or shear modulus. A crossing of shear elastic constants along axial direction and radial direction was observed near 150 .deg. C. This fact corresponds to the crossing of c44 and c66 of single crystal zirconium

  7. Time-resolved detection of temperature, concentration, and pressure in a shock tube by intracavity absorption spectroscopy

    Science.gov (United States)

    Fjodorow, Peter; Fikri, Mustapha; Schulz, Christof; Hellmig, Ortwin; Baev, Valery M.

    2016-06-01

    In this paper, we demonstrate the first application of intracavity absorption spectroscopy (ICAS) for monitoring species concentration, total pressure, and temperature in shock-tube experiments. ICAS with a broadband Er3+-doped fiber laser is applied to time-resolved measurements of absorption spectra of shock-heated C2H2. The measurements are performed in a spectral range between 6512 and 6542 cm-1, including many absorption lines of C2H2, with a time resolution of 100 µs and an effective absorption path length of 15 m. Up to 18-times increase of the total pressure and a temperature rise of up to 1200 K have been monitored. Due to the ability of simultaneously recording many absorption lines in a broad spectral range, the presented technique can also be applied to multi-component analysis of transient single-shot processes in reactive gas mixtures in shock tubes, pulse detonation engines, or explosions.

  8. Analysis of stress-induced Burgers vector anisotropy in pressurized tube specimens of irradiated ferritic-martensitic steel: JLF-1

    Energy Technology Data Exchange (ETDEWEB)

    Gelles, D.S. [Pacific Northwest National Lab., Richland, WA (United States); Shibayama, T. [Univ. of Hokkaido, Oarai, Ibaraki (Japan). Inst. for Materials Research

    1998-09-01

    A procedure for determining the Burgers vector anisotropy in irradiated ferritic steels allowing identification of all a<100> and all a/2<111> dislocations in a region of interest is applied to a pressurized tube specimen of JLF-1 irradiated at 430 C to 14.3 {times} 10{sup 22} n/cm{sup 2} (E > 0.1 MeV) or 61 dpa. Analysis of micrographs indicates large anisotropy in Burgers vector populations develop during irradiation creep.

  9. Independence and diversity in CANDU shutdown systems

    International Nuclear Information System (INIS)

    Atomic Energy Control Board regulations state that Canadian CANDU reactors shall have two fully effective, independent and diverse shutdown systems. The Darlington Nuclear Generating Station is the first power plant operated by Ontario Hydro to make use of software-based computer control in its shutdown systems. By virtue of the reliance placed on these systems to prevent exposure of the public to harmful radioactivity in the event of an accident, the shutdown system software has been categorized as safety critical software. An important issue that was considered in the design of the Darlington shutdown systems was how the software should be designed and incorporated into the systems to comply with the independence and diversity requirement. This paper describes how the independence and diversity requirement was complied with in previous CANDU shutdown system designs utilizing hardware components. The difference between systems utilizing hardware alone, and those utilizing both hardware and software are discussed. The results of a literature search into the issue of software diversity, the behaviour of multi-version software systems, and experience in other industries utilizing safety critical software are referred to. This paper advocates a systems approach to designing independent shutdown systems utilizing software. Opportunities exist at the system level for design decisions that can enhance software diversity and can reduce the likelihood of common mode faults in the systems. In the light of recent experience in implementing diverse safety critical software, potential improvements to the design process for CANDU shutdown systems are identified. (Author) 19 refs

  10. CANDU spent fuel dry storage interim technique

    International Nuclear Information System (INIS)

    CANDU heavy water reactor is developed by Atomic Energy of Canada (AECL) it has 40 years of design life. During operation, the reactor can discharge a lot of spent fuels by using natural uranium. The spent fuel interim storage should be considered because the spent fuel bay storage capacity is limited with 6 years inventory. Spent fuel wet interim storage technique was adopted by AECL before 1970s, but it is diseconomy and produced extra radiation waste. So based on CANDU smaller fuel bundle dimension, lighter weight, lower burn-up and no-critical risk, AECL developed spent fuel dry interim storage technique which was applied in many CANDU reactors. Spent fuel dry interim storage facility should be designed base on critical accident prevention, decay heat removal, radiation protection and fissionable material containment. According to this introduction, analysis spent fuel dry interim storage facility and equipment design feature, it can be concluded that spent fuel dry interim storage could be met with the design requirement. (author)

  11. Coolability of severely degraded CANDU cores. Revised

    Energy Technology Data Exchange (ETDEWEB)

    Meneley, D.A.; Blahnik, C.; Rogers, J.T.; Snell, V.G.; Nijhawan, S

    1996-01-01

    Analytical and experimental studies have shown that the separately cooled moderator in a CANDU reactor provides an effective heat sink in the event of a loss-of-coolant accident (LOCA) accompanied by total failure of the emergency core cooling system (ECCS). The moderator heat sink prevents fuel melting and maintains the integrity of the fuel channels, therefore terminating this severe accident short of severe core damage. Nevertheless, there is a probability, however low, that the moderator heat sink could fail in such an accident. The pioneering work of Rogers (1984) for such a severe accident using simplified models showed that the fuel channels would fail and a bed of dry, solid debris would be formed at the bottom of the calandria which would heat up and eventually melt. However, the molten pool of core material would be retained in the calandria vessel, cooled by the independently cooled shield-tank water, and would eventually resolidify. Thus, the calandria vessel would act inherently as a 'core-catcher' as long as the shield tank integrity is maintained. The present paper reviews subsequent work on the damage to a CANDU core under severe accident conditions and describes an empirically based mechanistic model of this process. It is shown that, for such severe accident sequences in a CANDU reactor, the end state following core disassembly consists of a porous bed of dry solid, coarse debris, irrespective of the initiating event and the core disassembly process. (author)

  12. Operative behaviour of a condenser tube under ETA chemistry

    International Nuclear Information System (INIS)

    Among the various recommendations for the surveillance of the integrity of the materials of the Secondary Cycle (Balance of Plant) it is the periodic removal of a steam generator tube and a condenser tube and their analysis. It considers assessment of the water chemistry, corrosion and the reciprocal effect on or from other components of the cycle. Embalse N.P.P. is a CANDU 6 type, Pressurized Heavy Water Reactor, located in Cordoba Province, Argentina. Previous papers have shown results on tubes removed from the steam generators (Bordoni et al., NPC'08, September 15-18, 2008, Berlin, Germany; 6th Canadian Nuclear Society - Steam Generators Conference, November 8-11, 2009, Toronto, Canada). Considering that the Embalse BOP has mixed metallurgy, i.e., steam generator tubes made of A800, piping made of ferrous alloys and condenser tubes made of Admiralty Brass and also taking into account that the chemistry has been modified from Morpholine control to ETA control (Fernandez et. al, NPC'2010, October 3-7, Quebec City, Canada), it has been decided to remove and analyze a condenser tube that has been placed in operation coincidently with the establishment of the ETA chemical control. The extraction is dated along with the November 2011 Plant Programmed Outage. Objectives are assessing the operative behavior of the tube performing visual and optical microscope inspection, SEM analysis of the oxides and deposits in exposed surfaces and occluded locations like tube sheet and other tests as well. Results are compared to the same analysis performed on a new tube in storage and integrated with the chemical operative figures of the cycle during the period: chemical data and corrosion products transport. (authors)

  13. Convective Heat and Mass Transfer in Water at Super—Critical Pressures under Heating or Cooling Conditions in Vertical Tubes

    Institute of Scientific and Technical Information of China (English)

    Pei-XueJiang; Ze-PeiRen; 等

    1995-01-01

    Forced and mixed convection heat and mass transfer are studied numerically for water containing metallic corrosion products in a heated or cooled vertical tube with variable thermophysical properties at super-citical pressures.the fouling mechanisms and fouling models are presented.The influence of variable properties at super-critical pressures on forced or mixed convection has been analyzed.The differences between heat and mass transfer under heating and cooling conditions are discussed.It is found that variable properties,especially buoyancy,greatly influence the fluid flow and heat mass fransfer.

  14. Validation of the assert subchannel code: Prediction of CHF in standard and non-standard Candu bundle geometries

    International Nuclear Information System (INIS)

    The ASSERT code has been developed to address the three-dimensional computation of flow and phase distribution and fuel element surface temperatures within the horizontal subchannels of CANDU PHWR fuel channels, and to provide a detailed prediction of critical heat flux (CHF) distribution throughout the bundle. The ASSERT subchannel code has been validated extensively against a wide repertoire of experiments; its combination of three-dimensional prediction of local flow conditions with a comprehensive method of prediting CHF at these local conditions, makes it a unique tool for predicting CHF for situations outside the existing experimental data base. In particular, ASSERT is an appropriate tool to systematically investigate CHF under conditions of local geometric variations, such as pressure tube creep and fuel element strain. This paper discusses the numerical methodology used in ASSERT, the constitutive relationships incorporated, and the CHF assessment methodology. The evolutionary validation plan is discussed, and early validation exercises are summarized. The paper concentrates, however, on more recent validation exercises in standard and non-standard geometries

  15. Validation of the ASSERT subchannel code for prediction of CHF in standard and non-standard CANDU bundle geometries

    International Nuclear Information System (INIS)

    The ASSERT code has been developed to address the three-dimensional computation of flow and phase distribution and fuel element surface temperatures within the horizontal subchannels of CANDU PHWR fuel channels, and to provide a detailed prediction of critical heat flux distribution throughout the bundle. The ASSERT subchannel code has been validated extensively against a wide repertoire of experiments; its combination of three-dimensional prediction of local flow conditions with a comprehensive method of predicting critical heat flux (CHF) at these local conditions makes it a unique tool for predicting CHF for situations outside the existing experimental data base. In particular, ASSERT is the only tool available to systematically investigate CHF under conditions of local geometric variations, such as pressure tube creep and fuel element strain. This paper discusses the numerical methodology used in ASSERT, the constitutive relationships incorporated, and the CHF assessment methodology. The evolutionary validation plan is discussed, and early validation exercises are summarized. The paper concentrates, however, on more recent validation exercises in standard and non-standard geometries. 28 refs., 12 figs

  16. Prediction of extubation outcome: a randomised, controlled trial with automatic tube compensation vs. pressure support ventilation

    OpenAIRE

    Cohen, Jonathan; Shapiro, Maury; Grozovski, Elad; Fox, Ben; Lev, Shaul; Singer, Pierre

    2009-01-01

    Introduction Tolerance of a spontaneous breathing trial is an evidence-based strategy to predict successful weaning from mechanical ventilation. Some patients may not tolerate the trial because of the respiratory load imposed by the endotracheal tube, so varying levels of respiratory support are widely used during the trial. Automatic tube compensation (ATC), specifically developed to overcome the imposed work of breathing because of artificial airways, appears ideally suited for the weaning ...

  17. Recent experience related to neutronic transients in Ontario Hydro CANDU nuclear generating stations

    International Nuclear Information System (INIS)

    Ontario Hydro presently operates 18 CANDU reactors in the province of Ontario, Canada. All of these reactors are of the CANDU Pressurized Heavy Water design, although their design features differ somewhat reflecting the evolution that has taken place from 1971 when the first Pickering unit started operation to the present as the Darlington units are being placed in service. Over the last three years, two significant neutronic transients took place at the Pickering Nuclear Generating Station 'A' (NGS A) one of which resulted in a number of fuel failures. Both events provided valuable lessons in the areas of operational safety, fuel performance And accident analysis. The events and the lessons learned are discussed in this paper

  18. Applications of ASTEC integral code on a generic CANDU 6

    Energy Technology Data Exchange (ETDEWEB)

    Radu, Gabriela, E-mail: gabriela.radu@nuclear.ro [Institute for Nuclear Research, Campului 1, 115400 Mioveni, Arges (Romania); Prisecaru, Ilie [Power Engineering Department, University “Politehnica” of Bucharest, 313 Splaiul Independentei, Bucharest (Romania)

    2015-05-15

    Highlights: • Short overview of the models included in the ASTEC MCCI module. • MEDICIS/CPA coupled calculations for a generic CANDU6 reactor. • Two cases taking into account different pool/concrete interface models. - Abstract: In case of a hypothetical severe accident in a nuclear power plant, the corium consisting of the molten reactor core and internal structures may flow onto the concrete floor of containment building. This would cause an interaction between the molten corium and the concrete (MCCI), in which the heat transfer from the hot melt to the concrete would cause the decomposition and the ablation of the concrete. The potential hazard of this interaction is the loss of integrity of the containment building and the release of fission products into the environment due to the possibility of a concrete foundation melt-through or containment over-pressurization by the gases produced from the decomposition of the concrete or by the inflammation of combustible gases. In the safety assessment of nuclear power plants, it is necessary to know the consequences of such a phenomenon. The paper presents an example of application of the ASTECv2 code to a generic CANDU6 reactor. This concerns the thermal-hydraulic behaviour of the containment during molten core–concrete interaction in the reactor vault. The calculations were carried out with the help of the MEDICIS MCCI module and the CPA containment module of ASTEC code coupled through a specific prediction–correction method, which consists in describing the heat exchanges with the vault walls and partially absorbent gases. Moreover, the heat conduction inside the vault walls is described. Two cases are presented in this paper taking into account two different heat transfer models at the pool/concrete interface and siliceous concrete. The corium pool configuration corresponds to a homogeneous configuration with a detailed description of the upper crust.

  19. Draft tube discharge fluctuation during self-sustained pressure surge: fluorescent particle image velocimetry in two-phase flow

    Science.gov (United States)

    Müller, A.; Dreyer, M.; Andreini, N.; Avellan, F.

    2013-04-01

    Hydraulic machines play an increasingly important role in providing a secondary energy reserve for the integration of renewable energy sources in the existing power grid. This requires a significant extension of their usual operating range, involving the presence of cavitating flow regimes in the draft tube. At overload conditions, the self-sustained oscillation of a large cavity at the runner outlet, called vortex rope, generates violent periodic pressure pulsations. In an effort to better understand the nature of this unstable behavior and its interaction with the surrounding hydraulic and mechanical system, the flow leaving the runner is investigated by means of particle image velocimetry. The measurements are performed in the draft tube cone of a reduced scale model of a Francis turbine. A cost-effective method for the in-house production of fluorescent seeding material is developed and described, based on off-the-shelf polyamide particles and Rhodamine B dye. Velocity profiles are obtained at three streamwise positions in the draft tube cone, and the corresponding discharge variation in presence of the vortex rope is calculated. The results suggest that 5-10 % of the discharge in the draft tube cone is passing inside the vortex rope.

  20. Correlation between the critical heat flux and the fractal surface roughness of zirconium alloy tubes

    International Nuclear Information System (INIS)

    In CANDU fuel channels, Zircaloy calandria tubes isolate the hot pressure tubes from the cool heavy water moderator. The heavy-water moderator provides a backup heat sink during some postulated loss-of-coolant accidents. The decay heat from the fuel is transferred to the moderator to ensure fuel channel integrity during emergencies. Moderator temperature requirements are specified to ensure that the transfer of decay heat does not exceed the critical heat flux (CHF) on the outside surface of the calandria tube. An enhanced CHF provides increases in safety margin. Pool boiling experiments indicate the CHF is enhanced with glass-peening of the outside surface of the calandria tubes. The objective of this study was to evaluate the surface characteristics of glass-peened tubes and relate these characteristics to CHF. The micro-topologies of the tube surfaces were analysed using stereo-pair micrographs obtained from scanning electron microscopy (SEM) and photogrammetry techniques. A linear relationship correlated the CHF as a function of the 'fractal' surface roughness of the tubes. (author)

  1. Response of the water status of soybean to changes in soil water potentials controlled by the water pressure in microporous tubes

    Science.gov (United States)

    Steinberg, S. L.; Henninger, D. L.

    1997-01-01

    Water transport through a microporous tube-soil-plant system was investigated by measuring the response of soil and plant water status to step change reductions in the water pressure within the tubes. Soybeans were germinated and grown in a porous ceramic 'soil' at a porous tube water pressure of -0.5 kpa for 28 d. During this time, the soil matric potential was nearly in equilibrium with tube water pressure. Water pressure in the porous tubes was then reduced to either -1.0, -1.5 or -2.0 kPa. Sap flow rates, leaf conductance and soil, root and leaf water potentials were measured before and after this change. A reduction in porous tube water pressure from -0.5 to -1.0 or -1.5 kPa did not result in any significant change in soil or plant water status. A reduction in porous tube water pressure to -2.0 kPa resulted in significant reductions in sap flow, leaf conductance, and soil, root and leaf water potentials. Hydraulic conductance, calculated as the transpiration rate/delta psi between two points in the water transport pathway, was used to analyse water transport through the tube-soil-plant continuum. At porous tube water pressures of -0.5 to-1.5 kPa soil moisture was readily available and hydraulic conductance of the plant limited water transport. At -2.0 kPa, hydraulic conductance of the bulk soil was the dominant factor in water movement.

  2. Burnup calculations of light water-cooled pressure tube blanket for a fusion-fission hybrid reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zu, Tiejun, E-mail: tiejun@mail.xjtu.edu.cn; Wu, Hongchun; Zheng, Youqi; Cao, Liangzhi

    2014-06-15

    Highlights: • Detailed burnup calculations are performed on pressurized water cooled blankets with pressure tube assemblies. • The blanket is fueled with simple fuel, namely spent nuclear fuel discharged from light water reactors or natural uranium oxide. • The refueling strategies are proposed, and the uranium resource utilization rate can reach 5–6%. - Abstract: A fusion-fission hybrid reactor (FFHR) with pressure tube blanket has recently been proposed based on an ITER-type tokamak fusion neutron source and the well-developed pressurized water cooling technologies. In this paper, detailed burnup calculations are carried out on an updated blanket. Two different blankets respectively fueled with the spent nuclear fuel (SNF) discharged from light water reactors (LWRs) or natural uranium oxide is investigated. In the first case, a three-batch out-to-in refueling strategy is designed. In the second case, some SNF assemblies are loaded into the blanket to help achieve tritium self-sufficiency. And a three-batch in-to-out refueling strategies is adopted to realize direct use of natural uranium oxide fuel in the blanket. The results show that only about 80 tonnes of SNF or natural uranium are needed every 1500 EFPD (Equivalent Full Power Day) with a 3000 MWth output and tritium self-sufficiency (TBR > 1.15), while the required maximum fusion powers are lower than 500 MW for both the two cases. Based on the proposed refueling strategies, the uranium utilization rate can reach about 4.0%.

  3. Measurement of mechanical properties of a reactor operated Zr–2.5Nb pressure tube using an in situ cyclic ball indentation system

    Energy Technology Data Exchange (ETDEWEB)

    Chatterjee, S., E-mail: subrata@barc.gov.in; Panwar, Sanjay; Madhusoodanan, K.

    2015-07-15

    Highlights: • Measurement of mechanical properties of pressure tube is required for its fitness assessment. • Pressure tube removal from the core consumes large amount of radiation for laboratory test. • A remotely operable In situProperty Measurement System has been designed in house. • The tool head is capable to carry out in situ ball indentation trials inside pressure tube. • The paper describes the theory and results of the trials conducted on irradiated pressure tube. - Abstract: Periodic measurement of mechanical properties of pressure tubes of Indian Pressurised Heavy Water Reactors is required for assessment of their fitness for continued operation. Removal of pressure tube from the core for preparation of specimens to test for mechanical properties in laboratories consumes large amounts of radiation and hence is to be avoided as far as possible. In the field of in situ estimation of properties of materials, cyclic ball indentation is an emerging technique. Presently, commercial systems are available for doing indentation test either on outside surface of a component at site or on a test piece in a laboratory. However, these systems cannot be used inside a pressure tube for carrying out ball indentation trials under in situ condition. Considering this, a remotely operable hydraulic In situProperty Measurement System (IProMS) based on cyclic ball indentation technique has been designed and developed in house. The tool head of IProMS can be located inside a pressure tube at any axial location under in situ condition and the properties can be estimated from an analysis of the data on load and depth of indentation, recorded during the test. In order to qualify the system, a number of experimental trials have been conducted on spool pieces and specimens prepared from Zr–2.5Nb pressure tube having different mechanical properties. Based on the encouraging results obtained from the qualification trials, IProMS has been used inside a reactor operated

  4. Measurement of mechanical properties of a reactor operated Zr–2.5Nb pressure tube using an in situ cyclic ball indentation system

    International Nuclear Information System (INIS)

    Highlights: • Measurement of mechanical properties of pressure tube is required for its fitness assessment. • Pressure tube removal from the core consumes large amount of radiation for laboratory test. • A remotely operable In situProperty Measurement System has been designed in house. • The tool head is capable to carry out in situ ball indentation trials inside pressure tube. • The paper describes the theory and results of the trials conducted on irradiated pressure tube. - Abstract: Periodic measurement of mechanical properties of pressure tubes of Indian Pressurised Heavy Water Reactors is required for assessment of their fitness for continued operation. Removal of pressure tube from the core for preparation of specimens to test for mechanical properties in laboratories consumes large amounts of radiation and hence is to be avoided as far as possible. In the field of in situ estimation of properties of materials, cyclic ball indentation is an emerging technique. Presently, commercial systems are available for doing indentation test either on outside surface of a component at site or on a test piece in a laboratory. However, these systems cannot be used inside a pressure tube for carrying out ball indentation trials under in situ condition. Considering this, a remotely operable hydraulic In situProperty Measurement System (IProMS) based on cyclic ball indentation technique has been designed and developed in house. The tool head of IProMS can be located inside a pressure tube at any axial location under in situ condition and the properties can be estimated from an analysis of the data on load and depth of indentation, recorded during the test. In order to qualify the system, a number of experimental trials have been conducted on spool pieces and specimens prepared from Zr–2.5Nb pressure tube having different mechanical properties. Based on the encouraging results obtained from the qualification trials, IProMS has been used inside a reactor operated

  5. Automatic ultrasonic inspection system for wear determination in calandria tubes of Embalse Nuclear Power Plant

    Science.gov (United States)

    Katchadjian, Pablo; Desimone, Carlos; Garcia, Alejandro; Antonaccio, Carlos; Schroeter, Fernando; Molina, Héctor

    2015-03-01

    Embalse Nuclear Power Plant (CNE) (CANDU design) is reaching its end of life and due to elapsed operating time the problem of deformation by accelerated creep occurs in the pressure tubes (PT), leading to a possible contact between calandria tubes (CT), concentric to the PT, and some Liquid Injection Shutdown System (LISS) nozzles that pass underneath them. With determination of CT wear, after the predicted contact occurs, the wear rate of the TC could be determined and thus take less conservative measures over the remaining life of the component. This paper presents the development of an ultrasonic technique for measuring wear in CT, with nominal thickness of 1.34 mm. Because the only access is through the interior of PT, to perform this measurement it is necessary to pass through three different interfaces.

  6. Automatic ultrasonic inspection system for wear determination in calandria tubes of Embalse Nuclear Power Plant

    International Nuclear Information System (INIS)

    Embalse Nuclear Power Plant (CNE) (CANDU design) is reaching its end of life and due to elapsed operating time the problem of deformation by accelerated creep occurs in the pressure tubes (PT), leading to a possible contact between calandria tubes (CT), concentric to the PT, and some Liquid Injection Shutdown System (LISS) nozzles that pass underneath them. With determination of CT wear, after the predicted contact occurs, the wear rate of the TC could be determined and thus take less conservative measures over the remaining life of the component. This paper presents the development of an ultrasonic technique for measuring wear in CT, with nominal thickness of 1.34 mm. Because the only access is through the interior of PT, to perform this measurement it is necessary to pass through three different interfaces

  7. Automatic ultrasonic inspection system for wear determination in calandria tubes of Embalse Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Katchadjian, Pablo, E-mail: katcha@cnea.gov.ar; Desimone, Carlos, E-mail: katcha@cnea.gov.ar; Garcia, Alejandro, E-mail: katcha@cnea.gov.ar [Comisión Nacional de Energía Atómica, Depto. ENDE - INEND, Av. Gral. Paz 1499, Buenos Aires (Argentina); Antonaccio, Carlos; Schroeter, Fernando; Molina, Héctor [Nucleoeléctrica Argentina-SA, Arribeños 3619, Buenos Aires (Argentina)

    2015-03-31

    Embalse Nuclear Power Plant (CNE) (CANDU design) is reaching its end of life and due to elapsed operating time the problem of deformation by accelerated creep occurs in the pressure tubes (PT), leading to a possible contact between calandria tubes (CT), concentric to the PT, and some Liquid Injection Shutdown System (LISS) nozzles that pass underneath them. With determination of CT wear, after the predicted contact occurs, the wear rate of the TC could be determined and thus take less conservative measures over the remaining life of the component. This paper presents the development of an ultrasonic technique for measuring wear in CT, with nominal thickness of 1.34 mm. Because the only access is through the interior of PT, to perform this measurement it is necessary to pass through three different interfaces.

  8. Effect of humidity, temperature, and pressure on corona discharge characteristics and heat transfer enhancements in a tube

    International Nuclear Information System (INIS)

    This paper reports on the effect of working fluid humidity, temperature, and pressure on corona discharge characteristics and heat transfer enhancements for air flow in a tube with a co-axial wire electrode that were studied experimentally. It was found that higher working fluid temperatures result in higher corona currents and higher corresponding heat transfer enhancements. A higher humidity yielded lower corona discharge currents. However, its effect on heat transfer enhancement was not as definitive and depended on the flow Reynolds number. With respect to the effect of pressure, an increase was associated with a rapid decrease in corona current and increase in threshold voltage for both positive and negative polarities. At low pressures discharge consisted of a stable corona at lower currents and an unstable corona at higher currents

  9. Effect of geometrical defects and cracks on the collapse of straigth or curved tubes submitted to external pressure

    International Nuclear Information System (INIS)

    The study has been initiated and guided by the need to establish clear and accurate design criteria for the collapse of tubular structures under external pressure for heat exchanger applications. Extensive numerical studies have been conducted to clarify the buckling of thick cylindrical tubes submitted to lateral pressure or hydrostatic pressure. Straight configuration or curved pipes are considered. Effects of geometrical initial imperfection (initial ovalization) are considered in both configurations for different amplitudes. The obtained design curve is then compared to the RCC-MR code or the German vessel code. Different strain hardenings are considered to gauge the effect of material law on the collapse behaviour. Considering this large parametric numerical study conducted with different FE codes (CAST3M, ABAQUS, ASTER), the methodology of the design and recommendations are also proposed

  10. Multi-step approach to Code-coupling for progression induced severe accidents in CANDU NPPs (MACPISA-CANDU)

    Energy Technology Data Exchange (ETDEWEB)

    Pohl, D.J.; Luxat, J.C. [McMaster Univ., Dept. of Engineering Physics, Hamilton, Ontario (Canada); Giannotti, W.; D' Auria, F. [Univ. of Pisa, Dept. of Mechanical, Nuclear and Production Engineering, Pisa (Italy)

    2009-07-01

    This paper reviews the progression of severe accidents, describes computer codes currently employed for analysis of severe accidents and outlines a new methodology to modelling the progression of severe accidents in CANDU nuclear power plants (NPPs) called the Multi-step Approach to Code-coupling for Progression Induced Severe Accidents in CANDU NPPs (MACPISA-CANDU). The MACPISA-CANDU methodology was used to couple the U.S. NRC codes SCDAP/RELAP5 (RELAP/SCDAPSIM Mod 3.4) and MELCOR (1.8.5) in order to model a small break loss of coolant accident with loss of emergency coolant injection (SBLOCA-LOECI) under natural circulation in a CANDU 6 NPP. Using this model it was shown that the sheath temperature did not exceed the zirconium melting temperature of 2098 K and hence the progression of the severe accident was terminated as expected. (author)

  11. Modeling and experiments with low-frequency pressure wave propagation in liquid-filled, flexible tubes

    DEFF Research Database (Denmark)

    Bjelland, C; Bjarnø, Leif

    1992-01-01

    A model for wave propagation in a liquid-filled viscoelastic tube with arrays of receivers inside, is being used to analyze the influence of noise generated by in-line vibrational noise sources. In this model, distensibility is of greater importance than compressibility of the liquid...... relations and frequency-dependent attenuation. A 12-m-long, liquid-filled tube with interior stress members and connectors in each end is hanging vertically from an upper fixture. The lower end connector is excited by a power vibrator to generate the relevant wave modes. Measurements with reference...

  12. Analysis of the impact of coolant density variations in the high efficiency channel of a pressure tube super critical water reactor

    International Nuclear Information System (INIS)

    The Pressure Tube (PT) Supercritical Water Reactor (SCWR) is based on a light water coolant operating at pressures above the thermodynamic critical pressure; a separate low temperature and low pressure moderator. The coolant density changes by an order of magnitude depending on its local enthalpy in the porous ceramic insulator tube. This causes significant changes in the neutron transport characteristics, axially and radially, in the fuel channel. This work performs lattice physics calculations for a 78-element Pu-Th fuel at zero burnup and examines the effect of assumptions related to coolant density in the radial direction of a HEC, using the neutron transport code WIMS-AECL. (author)

  13. Experimental Research on Heat Transfer and Pressure Drop of Two Configurations of Pin Finned—Tubes in an In—line Array

    Institute of Scientific and Technical Information of China (English)

    ShouGuangYao; DeShuZhu

    1994-01-01

    In this paper,a local simulation method is employed to investigate the heat transfer and pressure drop characteristics of two configurations of pin finned tubes deployed in an in-line array,In this research,heat pipes are adopted as heating elements.Therefore,the experimental equipment becomes simple and has an advantage of sufficient reducibility.The air-side heat transfer and pressure drop correlations for each type of pin fin surface including the effect of the tube-row number are obtained in the Reynolds number range commonly encountered in engineering.These correlations may be used in the design of pin finned tube heat exchangers.

  14. Configuration management for CANDU feeder refurbishment

    International Nuclear Information System (INIS)

    The Canada Deuterium Uranium Reactor CANDU was originally designed to last twenty-five years. In 2005, Atomic Energy of Canada (AECL) made the decision to extend the life of the reactor by thirty years. One of the most critical elements of the life extension project was determining how to refurbish the Primary Heat Transport System. It was determined that the refurbishment required replacing the entire length of inlet and outlet feeders, from the end fittings to the header. The use of a robust Configuration Management program would have added significant value to the life extension project. (author)

  15. Development of the advanced CANDU technology

    Energy Technology Data Exchange (ETDEWEB)

    Suk, Soo Dong; Min, Byung Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Na, Y. H.; Lee, S. Y.; Choi, J. H.; Lee, B. C.; Kim, S. N.; Jo, C. H.; Paik, J. S.; On, M. R.; Park, H. S.; Kim, S. R. [Korea Electric Power Co., Taejon (Korea, Republic of)

    1997-07-01

    The purpose of this study is to develop the advanced design technology to improve safety, operability and economy and to develop and advanced safety evaluation system. More realistic and reasonable methodology and modeling was employed to improve safety margin in containment analysis. Various efforts have been made to verify the CATHENA code which is the major safety analysis code for CANDU PHWR system. Fully computerized prototype ECCS was developed. The feasibility study and conceptual design of the distributed digital control system have been performed as well. The core characteristics of advanced fuel cycle, fuel management and power upgrade have been studied to determine the advanced core. (author). 77 refs., 51 tabs., 108 figs.

  16. A generalized perturbation program for CANDU reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Do Heon; Kim, Jong Kyung [Hanyang University, Seoul (Korea, Republic of); Choi, Hang Bok; Roh, Gyu Hong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Yang, Won Sik [Chosun University, Kwangju (Korea, Republic of)

    1998-12-31

    A generalized perturbation program has been developed for the purpose of estimating zonal power variation of a CANDU reactor upon refueling operation. The forward and adjoint calculation modules of RFSP code were used to construct the generalized perturbation program. The numerical algorithm for the generalized adjoint flux calculation was verified by comparing the zone power estimates upon refueling with those of forward calculation. It was, however, noticed that the truncation error from the iteration process of the generalized adjoint flux is not negligible. 2 refs., 1 figs., 1 tab. (Author)

  17. Backup and Ultimate Heat Sinks in CANDU Reactors For Prolonged SBO Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Nitheanandan, T.; Brown, M. J. [Atomic Energy of Canada Limited, Ontario (Canada)

    2013-10-15

    In a pressurized heavy water reactor, following loss of the primary coolant, severe core damage would begin with the depletion of the liquid moderator, exposing the top row of internally-voided fuel channels to steam cooling conditions on the inside and outside. The uncovered fuel channels would heat up, deform and disassemble into core debris. Large inventories of water passively reduce the rate of progression of the accident, prolonging the time for complete loss of engineered heat sinks. The efficacy of available backup and ultimate heat sinks, available in a CANDU 6 reactor, in mitigating the consequences of a prolonged station blackout scenario was analysed using the MAAP4-CANDU code. The analysis indicated that the steam generator secondary side water inventory is the most effective heat sink during the accident. Additional heat sinks such as the primary coolant, moderator, calandria vault water and end shield water are also able to remove decay heat; however, a gradually increasing mismatch between heat generation and heat removal occurs over the course of the postulated event. This mismatch is equivalent to an additional water inventory estimated to be 350,000 kg at the time of calandria vessel failure. In the Enhanced CANDU 6 reactor ∼2,040,000 kg of water in the reserve water tank is available for prolonged emergencies requiring heat sinks.

  18. Optimization and implementation study of plutonium disposition using existing CANDU Reactors. Final report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-09-01

    Since early 1994, the Department of Energy has been sponsoring studies aimed at evaluating the merits of disposing of surplus US weapons plutonium as Mixed Oxide (MOX) fuel in existing commercial Canadian Pressurized Heavy Water reactors, known as CANDU`s. The first report, submitted to DOE in July, 1994 (the 1994 Executive Summary is attached), identified practical and safe options for the consumption of 50 to 100 tons of plutonium in 25 years in some of the existing CANDU reactors operating the Bruce A generating station, on Lake Huron, about 300 km north east of Detroit. By designing the fuel and nuclear performance to operate within existing experience and operating/performance envelope, and by utilizing existing fuel fabrication and transportation facilities and methods, a low cost, low risk method for long term plutonium disposition was developed. In December, 1995, in response to evolving Mission Requirements, the DOE requested a further study of the CANDU option with emphasis on more rapid disposition of the plutonium, and retaining the early start and low risk features of the earlier work. This report is the result of that additional work.

  19. Performance evaluation of two CANDU fuel elements tested in the TRIGA reactor

    International Nuclear Information System (INIS)

    Nuclear Research Institute at Pitesti has a set of facilities, which allow the testing, manipulation and examination of nuclear fuel and structure materials irradiated in CANDU reactors from Cernavoda NPP. These facilities consist of TRIGA materials testing reactor and Post-Irradiation Examination Laboratory (LEPI). The purpose of this work is to describe the post-irradiation examination, of two experimental CANDU fuel elements (EC1 and EC2). The fuel elements were mounted into a pattern port, one in extension of the other in a measuring test for the central temperature evolution. The results of post-irradiation examination are obtained from: Visual inspection and photography of the outer appearance of sheath; Profilometry (diameter, bending, ovalization) and length measuring; Determination of axial and radial distribution of the fission products activity by gamma scanning; Measurement of pressure, volume and isotopic composition of fission gas; Microstructural characterization by metallographic and ceramographic analyzes; Isotopic composition and burn-up determination. The post-irradiation examination results are used, on one hand, to confirm the security, reliability and performance of the irradiated fuel, and on the other hand, for further development of CANDU fuel. (authors)

  20. Standard compliance - NDE performance demonstration/inspection in the CANDU industry

    International Nuclear Information System (INIS)

    CANDU nuclear power plants are operated in 3 provinces in Canada for electric power generation. A table in the paper will show the built and operating plants in Ontario, Quebec, New Brunswick and overseas. The regulator for nuclear power in Canada is the Canadian Nuclear Safety Commission (CNSC). The CNSC holds the plant licensees accountable for compliance to CSA N285.4 for periodic inspections. The Standard basically specifies the 'what, when, where, how, how much and how frequently' NDE is to be done on pressure retaining systems and components in CANDU nuclear power plants. In inspection methods, the Standard specifies they must be non-destructive. The NDE methods were grouped into visual, dimensional, surface, volumetric and integrative. The Standard also specifies that the licensees are responsible for the performance demonstration (PD) of the adequacy of the procedures and the proficiency of the personnel. This paper describes the Standard's requirement in NDE qualification and presents a joint project participated by Canadian and overseas CANDU owners. The sub-project for NDE included providing evidence and technical justification on the adequacy of the procedures and the proficiency of the personnel. The paper describes the qualification methodology followed by the participants. This will be followed by how the participants produced Inspection Specification, tools and procedures, personnel training and qualification programs, test and qualification samples, independent peer reviews and Technical Justification. (author)

  1. Implementation of an on-line monitoring system for transmitters in a CANDU nuclear power plant

    Science.gov (United States)

    Labbe, A.; Abdul-Nour, G.; Vaillancourt, R.; Komljenovic, D.

    2012-05-01

    Many transmitters (pressure, level and flow) are used in a nuclear power plant. It is necessary to calibrate them periodically to ensure that their measurements are accurate. These calibration tasks are time consuming and often contribute to worker radiation exposure. Human errors can also sometimes degrade their performance since the calibration involves intrusive techniques. More importantly, experience has shown that the majority of current calibration efforts are not necessary. These facts motivated the nuclear industry to develop new technologies for identifying drifting instruments. These technologies, well known as on-line monitoring (OLM) techniques, are non-intrusive and allow focusing the maintenance efforts on the instruments that really need a calibration. Although few OLM systems have been implemented in some PWR and BWR plants, these technologies are not commonly used and have not been permanently implemented in a CANDU plant. This paper presents the results of a research project that has been performed in a CANDU plant in order to validate the implementation of an OLM system. An application project, based on the ICMP algorithm developed by EPRI, has been carried out in order to evaluate the performance of an OLM system. The results demonstrated that the OLM system was able to detect the drift of an instrument in the majority of the studied cases. A feasibility study has also been completed and has demonstrated that the implementation of an OLM system at a CANDU nuclear power plant could be advantageous under certain conditions.

  2. Thermalhydraulic analysis of small-scale tube rupture experiments

    International Nuclear Information System (INIS)

    Guillotine failure of a rupturing pressure tube is an accident situation currently being investigated in the safety analysis of CANDU reactors. One of the reasons for initiating the investigation was to determine the major factors controlling the onset of guillotine failure. As part of this program, small-scale rupture tests using fuel sheaths have been performed and numerically simulated. The fluid dynamic aspects of rupturing fuel sheaths simulated with a multi-dimensional prototype of the two-fluid thermalhydraulic code CATHENA are described in this paper. The results of the numerical simulations were examined by observing the behaviour of pressure transients of the fluid inside the tube during the rupture. A parametric study was first performed to determine optimum model conditions for two-dimensional simulations. Results from CATHENA simulations using these conditions were then compared with experimental data. Calculations were also extended to a three-dimensional thermalhydraulic analysis. This paper describes the results of the parametric and comparative studies. The effect of varying the simulation conditions on calculated pressure transients is also described. Although agreement between simulated results and experimental data was found to be good, some discrepancies were noted and are discussed. Advantages and disadvantages of the three-dimensional study are also presented. This investigation has been successful in demonstrating a method that can be used to enhance the understanding of the behaviour of pressure-tube rupture under accident conditions. Areas in which the numerical analysis could be advanced to further the understanding of rupturing pressure tubes are provided. 3 refs., 11 figs., 2 tabs

  3. The Candu suppliers: Growing beyond Canada

    International Nuclear Information System (INIS)

    Canada has demonstrated its ability to successfully manage Candu projects. It has been shown domestically by Ontario Hydro over a period of many years. It has been demonstrated abroad by private sector service companies, under Nuclear Construction Managers, on the 600 MW Wolsung project in South Korea, where NPM companies carried out virtually all project and construction management, the plant was in service only 60 months after the first concrete was poured for the reactor building. In the past, however, the project management team for Candu nuclear plants overseas, and for some at home, was often assembled ad hoc. Though this approach can be successful, it does not guarantee success in project and construction management. Because of the size, diversity and flexibility of the private owners of NPM, they have been able to keep their experienced nuclear power teams on staff even though the nuclear power industry has been going through slow times, assigning these people to large projects in other industries that call for the same high level of project, construction and commissioning management skills. Yet they are in a position to reassemble this team at very short notice, and have them ready to go to work on a major project within weeks of the green light

  4. The creep life of superheater and reheater tubes under varying pressure conditions in operational boilers

    International Nuclear Information System (INIS)

    The first of each manufacturer's 500 MW boilers supplied to the CEGB (Central Electricity Generating Board) have been subjected to an extensive programme of tests for performance optimization and safe operation. Around 250 thermocouples on superheater and reheater tubes have in each case been monitored as part of the exercise. The readings are corrected and used to compute creep rupture damage based on internationally agreed stress rupture data and a simple cumulative damage concept. Comparison of the design creep rupture life and the cumulative life consumed has in several applications been invaluable in influencing operating procedures and arranging tube modifications or replacements, so that loss of generation by creep rupture failure is minimized. (author)

  5. Amorphous carbon film deposition on inner surface of tubes using atmospheric pressure pulsed filamentary plasma source

    CERN Document Server

    Pothiraja, Ramasamy; Awakowicz, Peter

    2011-01-01

    Uniform amorphous carbon film is deposited on the inner surface of quartz tube having the inner diameter of 6 mm and the outer diameter of 8 mm. A pulsed filamentary plasma source is used for the deposition. Long plasma filaments (~ 140 mm) as a positive discharge are generated inside the tube in argon with methane admixture. FTIR-ATR, XRD, SEM, LSM and XPS analyses give the conclusion that deposited film is amorphous composed of non-hydrogenated sp2 carbon and hydrogenated sp3 carbon. Plasma is characterized using optical emission spectroscopy, voltage-current measurement, microphotography and numerical simulation. On the basis of observed plasma parameters, the kinetics of the film deposition process is discussed.

  6. Experimental study of vapor local characteristics in upward low pressure boiling tube

    Institute of Scientific and Technical Information of China (English)

    SUN Qi; ZHAO Hua; XI Zhao; YANG Rui-Chang

    2003-01-01

    Radial distribution of vapor local parameters, including local void fraction, interfacial velocity, bubblesize, bubble frequency and interfacial area concentration, are investigated through the measurement in an upwardboiling tube using dual-sensor optical probe. In addition, a new local parameter -"local bubble number concentra-tion" is developed on the basis of bubble frequency. The analysis shows that this parameter can reflect bubble numberdensity in space, and has clear physical meaning.

  7. Pressure drop measurements in the transition region for a circular tube with a square-edged entrance

    Science.gov (United States)

    Ghajar, Afshin J.; Augustine, Jody R.

    1990-06-01

    Pressure drop measurements were made in a horizontal circular straight tube with a square-edged entrance under isothermal flow conditions. The experiments covered a Reynolds number range from 512 to 14,970. A total of thirty-three sets of experimental data for the twenty pressure tap locations along the 20 ft length of the test section were gathered. For the square-edged entrance the range of Reynolds number for which transition flow exists was determined to be between 2070 to 2840. A correlation for prediction of fully developed skin friction coefficient in this region is recommended. In the entrance region the length required for the friction factor to become fully developed in both the laminar and turbulent regions was found to be inversely proportional to the Reynolds number, with the turbulent data showing a stronger dependency. A correlation for prediction of entrance length in the turbulent region is offered.

  8. The evaluation of validity of the RELAP5/Mod3 flow regime map for horizontal small diameter tubes at low pressure

    Energy Technology Data Exchange (ETDEWEB)

    Agafonova, N. [St. Petersburg State Technical Univ. (Russian Federation); Banati, J. [Lappeenranta Univ. of Technology (Finland)

    1997-12-31

    RELAP5/MOD3 code was developed for Western type power water reactors with vertical steam generators. Thus, this code should be validated also for WWER design with horizontal steam generators. In application for horizontal steam generators the situation with two-phase flow inside small diameter tubes is possible when the first circuit pressure drops in accident below the pressure level in the boiling water. It is known that computer codes have not always modelled correctly the two-phase flow inside horizontal tubes at low pressures (less than 4-6 MPa). It may be the result of erroneous prediction of the flow regime. Correct prediction of the flow regime is especially important for the fully or partly stratified flow in horizontal tubes. The aim of this study is the attempt of verification of the flow regime map, which is used in the RELAP5/MOD3 computer code for two-phase flow in horizontal small diameter tubes. `Small diameter tube` means according RELAP5/MOD3 that the inner diameter of the tube is less (or equal) than 0.018 m. The inner tube diameter in horizontal steam generators is equal 0.013 m. (orig.). 19 refs.

  9. GERDA test facility for pressurized water reactors with straight tube steam generators

    International Nuclear Information System (INIS)

    A number of large-scale experimental facilities have been constructed and operate in order to experiment on the thermodynamic and thermohydraulic behaviour of nuclear facilities in case of LOCA. Most of them were designed for ''large leak'' accidents, but as ''small leak'' accidents became the focus of interest, such experiments were also carried out. Experiments carried out with this arrangement for PWR-type reactors with straight-tube steam generators are only partially evaluable. BBR and B and W therefore cooperated in the construction of the test facility GERDA, designed for testing reactors of BBR design. It supplied relevant experimental results for the nuclear power plant at Muelheim-Kaerlich. (orig.)

  10. General overview of CANDU advanced fuel cycles program

    International Nuclear Information System (INIS)

    The R and D program for CANDU advanced fuel cycles may be roughly divided into two components which have a near-and long-term focus, respectively. The near-term focus is on the technology to implement improved once-through cycles and mixed oxide (plutonium-uranium oxides) recycle in CANDU and on technologies to separate zirconium isotopes. Included is work on those technologies which would allow a CANDU-LWR strategy to be developed in a growing nuclear power system. For the longer-term, activities are focused on those technologies and fuel cycles which would be appropriate in a period when nuclear fuel demand significantly exceeds mined uranium supplies. Fuel cycles and systems under study are thorium recycle, CANDU fast breeder systems and electro-nuclear fissile breeders. The paper will discuss the rationale underlying these activities, together with a brief description of activities currently under way in each of the fuel cycle technology areas

  11. Proceedings of the Canadian Nuclear Society CANDU maintenance conference

    International Nuclear Information System (INIS)

    The conference proceedings comprise 51 papers on the following aspects of maintenance of CANDU reactors: Major maintenance projects, maintenance planning and preparation, maintenance effectiveness, future maintenance issues, safety and radiation protection. The individual papers have been abstracted separately

  12. Korea signs for 2nd CANDU at Wolsong

    International Nuclear Information System (INIS)

    The sale of a second CANDU 6 reactor to Korea for the Wolsong site is discussed in relation to nuclear power in Korea, the Korean economy generally, Canadian trade with Korea, and cooperation between AECL and KAERI

  13. Key thrusts in next generation CANDU. Annex 10

    International Nuclear Information System (INIS)

    Current electricity markets and the competitiveness of other generation options such as CCGT have influenced the directions of future nuclear generation. The next generation CANDU has used its key characteristics as the basis to leap frog into a new design featuring improved economics, enhanced passive safety, enhanced operability and demonstrated fuel cycle flexibility. Many enabling technologies spinning of current CANDU design features are used in the next generation design. Some of these technologies have been developed in support of existing plants and near term designs while others will need to be developed and tested. This paper will discuss the key principles driving the next generation CANDU design and the fuel cycle flexibility of the CANDU system which provide synergism with the PWR fuel cycle. (author)

  14. Reliable experimental setup to test the pressure modulation of Baerveldt Implant tubes for reducing post-operative hypotony

    Science.gov (United States)

    Ramani, Ajay

    Glaucoma encompasses a group of conditions that result in damage to the optic nerve and can cause loss of vision and blindness. The nerve is damaged due to an increase in the eye's internal (intraocular) pressure (IOP) above the nominal range of 15 -- 20 mm Hg. There are many treatments available for this group of diseases depending on the complexity and stage of nerve degradation. In extreme cases where drugs or laser surgery do not create better conditions for the patient, ophthalmologists use glaucoma drainage devices to help alleviate the IOP. Many drainage implants have been developed over the years and are in use; but two popular implants are the Baerveldt Glaucoma Implant and the Ahmed Glaucoma Valve Implant. Baerveldt Implants are non-valved and provide low initial resistance to outflow of fluid, resulting in post-operative complications such as hypotony, where the IOP drops below 5 mm of Hg. Ahmed Glaucoma Valve Implants are valved implants which initially restrict the amount of fluid flowing out of the eye. The long term success rates of Baerveldt Implants surpass those of Ahmed Valve Implants because of post-surgical issues; but Baerveldt Implants' initial effectiveness is poor without proper flow restriction. This drives the need to develop new ways to improve the initial effectiveness of Baerveldt Implants. A possible solution proposed by our research team is to place an insert in the Baerveldt Implant tube of inner diameter 305 microns. The insert must be designed to provide flow resistance for the early time frame [e.g., first 30 -- 60 post-operative days] until sufficient scar tissue has formed on the implant. After that initial stage with the insert, the scar tissue will provide the necessary flow resistance to maintain the IOP above 5 mm Hg. The main objective of this project was to develop and validate an experimental apparatus to measure pressure drop across a Baerveldt Implant tube, with and without inserts. This setup will be used in the

  15. Thermal-hydraulic interfacing code modules for CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Liu, W.S.; Gold, M.; Sills, H. [Ontario Hydro Nuclear, Toronto (Canada)] [and others

    1997-07-01

    The approach for CANDU reactor safety analysis in Ontario Hydro Nuclear (OHN) and Atomic Energy of Canada Limited (AECL) is presented. Reflecting the unique characteristics of CANDU reactors, the procedure of coupling the thermal-hydraulics, reactor physics and fuel channel/element codes in the safety analysis is described. The experience generated in the Canadian nuclear industry may be useful to other types of reactors in the areas of reactor safety analysis.

  16. Study of a high-temperature and high-pressure FBG sensor with Al2O3 thin-wall tube substrate

    Institute of Scientific and Technical Information of China (English)

    ZHOU Hong; QIAO Xue-guang; WANG Hong-liang; FENG De-quan; WANG Wei

    2008-01-01

    A fiber Bragg grating (FBG) high-temperature and high pressure sensor has been designed and fabricated by using the Al2O3 thin-wall tube as a substrate. The test results show that the sensor can withstand a pressure range of 0-45 MPa and a temperature range of-10-300℃, and has a pressure sensitivity of 0.0426 nm/MPa and a temperature sensitivity of 0.0112nm/℃

  17. Application of automatic inspection system to nondestructive test of heat transfer tubes of primary pressurized water cooler in the high temperature engineering test reactor. Joint research

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Takeshi; Furusawa, Takayuki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Miyamoto, Satoshi [Japan Atomic Power Company, Tokyo (Japan)

    2001-07-01

    Heat transfer tubes of a primary pressurized water cooled (PPWC) in the high temperature engineering test reactor (HTTR) form the reactor pressure boundary of the primary coolant, therefore are important from the viewpoint of safety. To establish inspection techniques for the heat transfer tubes of the PPWC, an automatic inspection system was developed. The system employs a bobbin coil probe, a rotating probe for eddy current testing (ECT) and a rotating probe for ultrasonic testing (UT). Nondestructive test of a half of the heat transfer tubes of the PPWC was carried out by the automatic inspection system during reactor shutdown period of the HTTR (about 55% in the maximum reactor power in this paper). The nondestructive test results showed that the maximum signal-to-noise ratio was 1.8 in ECT. Pattern and phase of Lissajous wave, which were obtained for the heat transfer tube of the PPWC, were different from those obtained for the artificially defected tube. In UT echo amplitude of the PPWC tubes inspected was lower than 20% of distance-amplitude calibration curve. Thus, it was confirmed that there was no defect in depth, which was more than the detecting standard of the probes, on the outer surface of the heat transfer tubes of the PPWC inspected. (author)

  18. Proceedings of the international conference on CANDU fuel

    International Nuclear Information System (INIS)

    These proceedings contain full texts of all paper presented at the first International Conference on CANDU Fuel. The Conference was organized and hosted by the Chalk River Branch of the Canadian Nuclear Society and utilized Atomic Energy of Canada Limited's facilities at Chalk River Nuclear Laboratories. Previously, informal Fuel Information Meetings were used in Canada to allow the exchange of information and technology associated with CANDU. The Chalk River conference was the first open international forum devoted solely to CANDU and included representatives of overseas countries with current or potential CANDU programs, as well as Canadian participants. The keynote presentation was given by Dr. J.B. Slater, who noted the correlation between past successes in CANDU fuel cycle technology and the co-operation between researchers, fabricators and reactor owner/operators in all phases of the fuel cycle, and outlined the challenges facing the industry today. In the banquet address, Dr. R.E. Green described the newly restructured AECL Research Company and its mission which blends traditional R and D with commercial initiatives. Since this forum for fuel technology has proven to be valuable, a second International CANDU Fuel Conference is planned for the fall of 1989, again sponsored by the Canadian Nuclear Society

  19. Conceptual Study on Dismantling of CANDU Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Woo-Tae; Lee, Sang-Guk [KHNP-CRI, Daejeon (Korea, Republic of)

    2014-10-15

    In this paper, we reviewed 3D design model of the CANDU type reactor and suggested feasible cutting scheme. The structure of CANDU nuclear reactor, the calandria assembly was reviewed using 3-D CAD model for future decommissioning. Through the schematic diagram of CANDU nuclear power plant, we identified the differences between PWR and CANDU reactor assembly. Method of dismantling the fuel channels from the calandria assembly was suggested. Custom made cutter is recommended to cut all the fuel channels. The calandria vessel is recommended to be cut by band saw or plasma torch. After removal of the fuel channels, it was assumed that radiation level near the calandria vessel is not very high. For cutting of the end shields, various methods such as band saw, plasma torch, CAMC could be used. The choice of a specific method is largely dependent on radiological environment. Finally, method of cutting the embedment rings is considered. As we assume that operators could cut the rings without much radiation exposure, various industrial cutting methods are suggested to be applied. From the above reviews, we could conclude that decommissioning of CANDU reactor is relatively easy compared to that of PWR reactor. Technologies developed from PWR reactor decommissioning could be applied to CANDU reactor dismantling.

  20. CANDU energy for steam assisted gravity drainage

    International Nuclear Information System (INIS)

    Traditional open-pit mining has been used by industry for many years to remove oil sands from shallow deposits. To increase production capacity, the industry is looking for new technology to exploit bitumen from deep deposits. Among them, SAGD (Steam-Assisted Gravity Drainage) appears to be the most promising approach. It uses steam to remove bitumen from underground reservoirs. Recently, the SAGD recovery process has been put into commercial operation by major oil companies.Atomic Energy Canada Limited has assessed the use of the ACR-1000 as a source of heat and electricity for oil sand extraction and processing. The ACR-1000 design is an evolutionary development of the familiar CANDU technology, adding innovations to enhance economics, operations, and safety margins. The net electrical output from a standard ACR-1000 will be close to 1100 MWe, depending on local cooling water temperature

  1. SLARette operations at CANDU 6 stations

    International Nuclear Information System (INIS)

    The Objective of the SLARette system as an ongoing maintenance device was to design a system which could be installed in reactor, commissioned and operational in less than 48 hours and removed in less than 24 hours, thus allowing a selected number of fuel channels to be SLARed during the normal yearly maintenance outage. This required a relatively simple design that does not require any changes to the Station Fuel Handling Systems and is portable, reliable and easily maintained. Over the last 10 years virtually every part of the SLARette system has been enhanced to further reduce man-hour and man-rem expenditure and it has now been used successfully on four CANDU 6 stations and many fuel channels. (author)

  2. Intercomparison of techniques for inspection and diagnostics of heavy water reactor pressure tubes: Flaw detection and characterization [Phase 1

    International Nuclear Information System (INIS)

    Nuclear power plants with heavy water reactors (HWRs) comprise nine percent of today's operating nuclear units, and more are under construction. Efficient and accurate inspection and diagnostic techniques for various reactor components and systems are an important factor in assuring reliable and safe plant operation. To foster international collaboration in the efficient and safe use of nuclear power, the IAEA conducted a Coordinated Research Programme (CRP) on Inter-comparison of Techniques for HWR Pressure Tube Inspection and Diagnostics. This CRP was carried out within the frame of the IAEA Department of Nuclear Energy's Technical Working Group on Advanced Technologies for HWRs (the TWG-HWR). The TWG-HWR is a group of experts nominated by their governments and designated by the IAEA to provide advice and to support implementation of the IAEA's project on advanced technologies for HWRs. The objective of the CRP was to inter-compare non-destructive inspection and diagnostic techniques, in use and being developed, for structural integrity assessment of HWR pressure tubes. During the first phase of this CRP, participants have investigated the capability of different techniques to detect and characterize flaws. During the second phase of this CRP, participants collaborated to detect and characterize hydride blisters and to determine the hydrogen concentration in Zirconium alloys. The intent was to identify the most effective pressure tube inspection and diagnostic methods, and to identify further development needs. The organizations that have participated in this CRP are: - The Comision Nacional de Energia Atomica (CNEA), Argentina; - Atomic Energy of Canada Ltd. (AECL); Chalk River Laboratories (CRL), Canada; - The Research Institute of Nuclear Power Operations (RINPO), China National Nuclear Corporation (CNNC), China; - Bhabha Atomic Research Centre (BARC), India; - The Korea Electric Power Research Institute (KEPRI), Republic of Korea; - The Korea Atomic Energy

  3. Consideration on evaluation of internal pressure creep rupture for tube with circumferential joint

    International Nuclear Information System (INIS)

    The behavior of internal pressure creep rupture of the thin-walled cylinders with circumferential joints is affected by the combination of creep characteristics of parent materials and weld metals. In particular, the compatibility of the creep strain rate of parent materials and weld metals becomes an important controlling factor. The behavior of internal pressure creep of the welded parts in circumferential joint cylinders can be evaluated simply with the uniaxial creep data of parent materials and weld metals, considering it by approximately substituting with the creep behavior of a uniaxial longitudinal joint. The method of evaluation is, first, to analyze the breaking behavior of uniaxial longitudinal joints using the uniaxial creep characteristic values of parent materials and weld metals, and next, by combining the equation for the relation between the rupture times of uniaxial creep and internal pressure creep with the analyzed breaking behavior of uniaxial joints, the internal pressure creep rupture behavior of the cylinders with circumferential joints can be evaluated. The internal pressure creep behavior of the thin-walled cylinders with circumferential joints, their rupture life and the uniaxial creep rupture life of longitudinal joints, and the examination of Hastelloy X cylinders are reported. (Kako, I.)

  4. Numerical Study on the Heat Transfer of Carbon Dioxide in Horizontal Straight Tubes under Supercritical Pressure.

    Directory of Open Access Journals (Sweden)

    Mei Yang

    Full Text Available Cooling heat transfer of supercritical CO2 in horizontal straight tubes with wall is numerically investigated by using FLUENT. The results show that almost all models are able to present the trend of heat transfer qualitatively, and the stand k-ε with enhanced wall treatment model shows the best agreement with the experimental data, followed by LB low Re turbulence model. Then further studies are discussed on velocity, temperature and turbulence distributions. The parameters which are defined as the criterion of buoyancy effect on convection heat transfer are introduced to judge the condition of the fluid. The relationships among the inlet temperature, outlet temperature, the mass flow rate, the heat flux and the diameter are discussed and the difference between the cooling and heating of CO2 are compared.

  5. Numerical Study on the Heat Transfer of Carbon Dioxide in Horizontal Straight Tubes under Supercritical Pressure.

    Science.gov (United States)

    Yang, Mei

    2016-01-01

    Cooling heat transfer of supercritical CO2 in horizontal straight tubes with wall is numerically investigated by using FLUENT. The results show that almost all models are able to present the trend of heat transfer qualitatively, and the stand k-ε with enhanced wall treatment model shows the best agreement with the experimental data, followed by LB low Re turbulence model. Then further studies are discussed on velocity, temperature and turbulence distributions. The parameters which are defined as the criterion of buoyancy effect on convection heat transfer are introduced to judge the condition of the fluid. The relationships among the inlet temperature, outlet temperature, the mass flow rate, the heat flux and the diameter are discussed and the difference between the cooling and heating of CO2 are compared. PMID:27458729

  6. Numerical Study on the Heat Transfer of Carbon Dioxide in Horizontal Straight Tubes under Supercritical Pressure.

    Science.gov (United States)

    Yang, Mei

    2016-01-01

    Cooling heat transfer of supercritical CO2 in horizontal straight tubes with wall is numerically investigated by using FLUENT. The results show that almost all models are able to present the trend of heat transfer qualitatively, and the stand k-ε with enhanced wall treatment model shows the best agreement with the experimental data, followed by LB low Re turbulence model. Then further studies are discussed on velocity, temperature and turbulence distributions. The parameters which are defined as the criterion of buoyancy effect on convection heat transfer are introduced to judge the condition of the fluid. The relationships among the inlet temperature, outlet temperature, the mass flow rate, the heat flux and the diameter are discussed and the difference between the cooling and heating of CO2 are compared.

  7. Feasibility study of modeling a CANDU fuel element using a multiphysics object-oriented simulation environment

    International Nuclear Information System (INIS)

    The first phase of the feasibility study of using a Multiphysics Object-Oriented Simulation Environment (MOOSE) for modeling a CANDU fuel element is presented. A two-dimensional model of a fuel pellet sheath was created to examine the contact algorithm within MOOSE. The results obtained show the expected behaviour of contact pressure and penetration in 2D. Preliminary results for a 3D model of a quarter fuel pellet and sheath are provided but at present contain anomalies currently being investigated. The next steps in the feasibility study are outlined. (author)

  8. Feasibility study of modeling a CANDU fuel element using a multiphysics object-oriented simulation environment

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, K., E-mail: Kyle.Gamble@rmc.ca [Royal Military College of Ontario, Kingston, Ontario (Canada); Williams, A. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Chan, P.K. [Royal Military College of Ontario, Kingston, Ontario (Canada)

    2013-07-01

    The first phase of the feasibility study of using a Multiphysics Object-Oriented Simulation Environment (MOOSE) for modeling a CANDU fuel element is presented. A two-dimensional model of a fuel pellet sheath was created to examine the contact algorithm within MOOSE. The results obtained show the expected behaviour of contact pressure and penetration in 2D. Preliminary results for a 3D model of a quarter fuel pellet and sheath are provided but at present contain anomalies currently being investigated. The next steps in the feasibility study are outlined. (author)

  9. CANDU-OCR power station options and costs

    International Nuclear Information System (INIS)

    This report updates and in some cases expands the technical and economic parameters presented originally in AECL-4441. 'Summary report on the design of a prototypical 500 MWe CANDU-OCR power station.' Updating is desirable owing to the increasing number of inquiries that have been received by Atomic Energy of Canada Ltd. from government agencies and the private sector. Each is exploring the available options in their continuing endeavour to provide sufficient and economical energy. The organic-cooled reactor (OCR) concept is particularly interesting to the oil industry because the high steam pressures it can develop allow it to be used for heavy oil extraction can also be used economically for other large thermal and electrical energy production requirements such as those encountered in district heating schemes, heavy water production and electricity production. The report describes a reference OCR-500 MWe reactor. It includes an overview of organic reactor experience and areas for further development based on 14 years of operating experience with the WR-1 reactor. A discussion of several variations on the reference design is given including estimates of costs for various reactor sizes, enrichments and operating functions. Costs are presented in a form which allow easy comparison with those of competing energy options. (auth)

  10. Experimental investigation of heat transfer and pressure drop of turbulent flow inside tube with inserted helical coils

    Science.gov (United States)

    Sharafeldeen, M. A.; Berbish, N. S.; Moawed, M. A.; Ali, R. K.

    2016-08-01

    The heat transfer and pressure drop were experimentally investigated in a coiled wire inserted tube in turbulent flow regime in the range of Reynolds number of 14,400 ≤ Re ≤ 42,900. The present work aims to extend the experimental data available on wire coil inserts to cover wire diameter ratio of 0.044 ≤ e/d ≤ 0.133 and coil pitch ratio of 1 ≤ p/d ≤ 5. Uniform heat flux was applied to the external surface of the tube and air was selected as fluid. The effects of Reynolds number and wire diameter and coil pitch ratios on the Nusselt number and friction factor were studied. The enhancement efficiency and performance criteria ranges are of (46.9-82.6 %) and (100.1-128 %) within the investigated range of the different parameters, respectively. Correlations are obtained for the average Nusselt number and friction factor utilizing the present measurements within the investigated range of geometrical parameters and Re. The maximum deviation between correlated and experimental values for Nusselt number and friction factor are ±5 and ±6 %, respectively.

  11. An experimental investigation of pressure drop of aqueous foam in laminar tube flow

    Science.gov (United States)

    Blackwell, B. F.; Sobolik, K. B.

    1987-04-01

    This report is the first of two detailing pressure-drop and heat-transfer measurements made at the Foam Flow Heat Transfer Loop. The work was motivated by a desire to extend the application of aqueous foam from petroleum drilling to geothermal drilling. Pressure-drop measurements are detailed in this report; a forthcoming report (SAND85-1922) will describe the heat-transfer measurements. The pressure change across a 2.4-m (8-ft) length of the 2.588-cm (1.019-in.) ID test section was measured for liquid volume fractions between 0.05 and 0.35 and average velocities between 0.12 and 0.80 m/s (0.4 and 2.6 ft/s). The resulting pressure-drop/flow-rate data were correlated to a theoretical model for a Bingham plastic. Simple expressions for the dynamic viscosity and the yield stress as a function of liquid volume fraction were estimated.

  12. Severe accident analysis of a station blackout accident using MAAP-CANDU for the Point Lepreau station refurbishment project level 2 PSA

    Energy Technology Data Exchange (ETDEWEB)

    Brown, M.J.; Petoukhov, S.M. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2012-07-01

    A Level 2 Probabilistic Safety Assessment was performed for the Point Lepreau Generating Station, using the MAAP-CANDU code to simulate the progression of severe core damage accidents and fission product releases. Five representative severe accidents were selected: Station Blackout, Small Loss-of-Coolant, Stagnation Feeder Break, Steam Generator Tube Rupture, and Shutdown State. Analysis results for the reference station blackout accident are discussed in this paper. (author)

  13. Method of experimental and theoretical modeling for multiple pressure tube rupture for RBMK reactor

    International Nuclear Information System (INIS)

    The rupture of single RBMK reactor channels has occurred at a number of stations with a variety of initiating events. It is assumed in RBMK Safety Cases that the force of the escaping fluid will not cause neighbouring channels to break. This assumption has not been justified. A chain reaction of tube breaks could over-pressurise the reactor cavity leading to catastrophic failure of the containment. To validate the claims of the RBMK Safety Cases the Electrogorsk Research and Engineering Centre, in participation with experts from the Institute of Mechanics of RAS, has developed the method of interacting multiscale physical and mathematical modelling for coupled thermophysical, hydrogasodynamic processes and deformation and break processes causing and (or) accompanying potential failures, design and beyond the design RBMK reactor accidents. To realise the method the set of rigs, physical and mathematical models and specialized computer codes are under creation. This article sets out an experimental philosophy and programme for achieving this objective to solve the problem of credibility or non-credibility for multiple fuel channel rupture in RBMK.(author)

  14. Pressure and Thrust Measurements of a High-Frequency Pulsed Detonation Tube

    Science.gov (United States)

    Nguyen, N.; Cutler, A. D.

    2008-01-01

    This paper describes measurements of a small-scale, high-frequency pulsed detonation tube. The device utilized a mixture of H2 fuel and air, which was injected into the device at frequencies of up to 1200 Hz. Pulsed detonations were demonstrated in an 8-inch long combustion volume, at about 600 Hz, for the quarter wave mode of resonance. The primary objective of this experiment was to measure the generated thrust. A mean value of thrust was measured up to 6.0 lb, corresponding to H2 flow based specific impulse of 2970 s. This value is comparable to measurements in H2-fueled pulsed detonation engines (PDEs). The injection and detonation frequency for this new experimental case was much higher than typical PDEs, where frequencies are usually less than 100 Hz. The compact size of the device and high frequency of detonation yields a thrust-per-unit-volume of approximately 2.0 pounds per cubic inch, and compares favorably with other experiments, which typically have thrust-per-unit-volume of order 0.01 pound per cubic inch. This much higher volumetric efficiency results in a potentially much more practical device than the typical PDE, for a wide range of potential applications, including high-speed boundary layer separation control, for example in hypersonic engine inlets, and propulsion for small aircraft and missiles.

  15. Power pulse tests on CANDU type fuel elements in TRIGA reactor of INR Pitesti

    International Nuclear Information System (INIS)

    Pulse irradiation tests on short fuel elements have been carried out in TRIGA Annular Core Pulse Reactor (TRIGA ACPR) of INR Pitesti to investigate aspects related to the thermal and mechanical behavior of CANDU type fuel elements under short duration and large amplitude power pulse conditions. Short test fuel elements were instrumented with thermocouples for cladding surface temperature measurements and pressure sensor for element internal pressure measurement. Transient histories of reactor power, cooling water pressure, fuel element internal pressure and cladding temperature were recorded during tests. The fuel elements were subjected to total energy deposition from 70 to 280cal g-1 UO2. Rapid fuel pellet expansion due to a power excursion caused radial and longitudinal deformation of the cladding. Cladding failure mechanism and the failure threshold have been established. This paper presents some recent results obtained from these power pulse tests performed in TRIGA ACPR of INR Pitesti. (author)

  16. Behavior of CANDU fuel under power pulse conditions at the TRIGA reactor of INR Pitesti

    International Nuclear Information System (INIS)

    Pulse irradiation tests on short fuel elements have been carried out in TRIGA Annular Core Pulse Reactor (TRIGA ACPR) of INR Pitesti to investigate aspects related to the thermal and mechanical behavior of CANDU type fuel elements under short duration and large amplitude power pulse conditions. Short test fuel elements were instrumented with thermocouples for cladding surface temperature measurements and pressure sensors for element internal pressure measurement. Transient histories of reactor power, cooling water pressure, fuel element internal pressure and cladding temperature were recorded during tests. The fuel elements were subjected to total energy deposition from 70 to 280 cal g-1 UO2. Rapid fuel pellet expansion due to a power excursion caused radial and longitudinal deformation of the cladding. Cladding failure mechanism and the failure threshold have been established. This paper presents some recent results obtained from these power pulse tests performed in TRIGA ACPR of INR Pitesti. (orig.)

  17. Evolution of the CANDU ICS-90+ control room design

    International Nuclear Information System (INIS)

    The design of the CANDU Control Room and the associated design process has evolved considerably over several generations of plants, from the first commercial scale demonstration CANDU at Douglas Point through to the large scale CANDUs at Darlington, and beyond, for the next generation of CANDU plant, ICS-90+, represented by new designs like CANDU 3. In the early plants, the control room configuration was based on designers' projections of control interface requirements. With succeeding generations, of designs, there has been an evolution towards: increasing attention to formal requirements definition, incorporation into the Human Machine Interface (HMI) of a larger base of operational experience, more systematic consideration of Human Factors (HF) aspects of the design and the application of a more powerful computer based HMI. For the newest plant, the CANDU 3, a Human Factors Engineering Program Plan (HFEPP) defines the overall HF engineering process, the associated requirements and HF engineering standards to be followed in each stage, and for all HMI aspects of the control room and plant design. The CANDU 3 control room also incorporates several new design innovations that will facilitate operating crew performance improvements. These are based on past experience with operating CANDU plants, incorporated with the use of formal design and validation methods plus results from Canadian research program to support control centre design and operation. For example, there are design improvements to facilitate: operator tracking of plant state, problem solving, alarm filtering, annunciation system interrogation, special safety system testing features, etc. The present paper will expand and elaborate on each of the above topics. (author). 7 refs, 2 figs, 1 tab

  18. Heat transfer, pressure drop and void fraction in two- phase, two-component flow in a vertical tube

    Science.gov (United States)

    Sujumnong, Manit

    1998-09-01

    There are very few data existing in two-phase, two- component flow where heat transfer, pressure drop and void fraction have all been measured under the same conditions. Such data are very valuable for two-phase heat-transfer model development and for testing existing heat-transfer models or correlations requiring frictional pressure drop (or wall shear stress) and/or void fraction. An experiment was performed which adds markedly to the available data of the type described in terms of the range of gas and liquid flow rates and liquid Prandtl number. Heat transfer and pressure drop measurements were taken in a vertical 11.68-mm i.d. tube for two-phase (gas-liquid) flows covering a wide range of conditions. Mean void fraction measurements were taken, using quick- closing valves, in a 12.7-mm i.d. tube matching very closely pressures, temperatures, gas-phase superficial velocities and liquid-phase superficial velocities to those used in the heat-transfer and pressure-drop experiments. The gas phase was air while water and two aqueous solutions of glycerine (59 and 82% by mass) were used as the liquid phase. In the two-phase experiments the liquid Prandtl number varied from 6 to 766, the superficial liquid velocity from 0.05 to 8.5 m/s, and the superficial gas velocity from 0.02 to 119 m/s. The measured two-phase heat-transfer coefficients varied by a factor of approximately 1000, the two-phase frictional pressure drop ranged from small negative values (in slug flow) to 93 kPa and the void fraction ranged from 0.01 to 0.99; the flow patterns observed included bubble, slug, churn, annular, froth, the various transitions and annular-mist. Existing heat-transfer models or correlations requiring frictional pressure drop (or wall shear stress) and/or void fraction were: tested against the present data for mean heat-transfer coefficients. It was found that the methods with more restrictions (in terms of the applicable range of void fraction, liquid Prandtl number or liquid

  19. Enhancing the aggressive intensity of hydrodynamic cavitation through a Venturi tube by increasing the pressure in the region where the bubbles collapse

    Science.gov (United States)

    Soyama, H.; Hoshino, J.

    2016-04-01

    In this paper, we used a Venturi tube for generating hydrodynamic cavitation, and in order to obtain the optimum conditions for this to be used in chemical processes, the relationship between the aggressive intensity of the cavitation and the downstream pressure where the cavitation bubbles collapse was investigated. The acoustic power and the luminescence induced by the bubbles collapsing were investigated under various cavitating conditions, and the relationships between these and the cavitation number, which depends on the upstream pressure, the downstream pressure at the throat of the tube and the vapor pressure of the test water, was found. It was shown that the optimum downstream pressure, i.e., the pressure in the region where the bubbles collapse, increased the aggressive intensity by a factor of about 100 compared to atmospheric pressure without the need to increase the input power. Although the optimum downstream pressure varied with the upstream pressure, the cavitation number giving the optimum conditions was constant for all upstream pressures.

  20. A Shock Tube Study of the CO + OH Reaction Near the Low-Pressure Limit

    KAUST Repository

    Nasir, Ehson Fawad

    2016-05-16

    Rate coefficients for the reaction between carbon monoxide and hydroxyl radical were measured behind reflected shock waves over 700 – 1230 K and 1.2 – 9.8 bar. The temperature/pressure conditions correspond to the predicted low-pressure limit of this reaction, where the channel leading to carbon dioxide formation is dominant. The reaction rate coefficients were inferred by measuring the formation of carbon dioxide using quantum cascade laser absorption near 4.2 µm. Experiments were performed under pseudo-first order conditions with tert-butyl hydroperoxide (TBHP) as the OH precursor. Using ultraviolet laser absorption by OH radicals, the TBHP decomposition rate was measured to quantify potential facility effects under extremely dilute conditions used here. The measured CO + OH rate coefficients are provided in Arrhenius form for three different pressure ranges: kCO+OH (1.2 – 1.6 bar) = 9.14 x 10-13 exp(-1265/T) cm3 molecule-1 s-1 kCO+OH (4.3 – 5.1 bar) = 8.70 x 10-13 exp(-1156/T) cm3 molecule-1 s-1 kCO+OH (9.6 – 9.8 bar) = 7.48 x 10-13 exp(-929/T) cm3 molecule-1 s-1 The measured rate coefficients are found to be lower than the master equation modeling results by Weston et al. [J. Phys. Chem. A, 117 (2013) 821] at 819 K and in closer agreement with the expression provided by Joshi and Wang [Int. J. Chem. Kinet., 38 (2006) 57].

  1. Experimental investigation of symmetric and asymmetric heating of pressure tube under accident conditions for Indian PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Yadav, Ashwini K., E-mail: ashwinikumaryadav@gmail.com [Department of Mechanical and Industrial Engineering, Indian Institute of Technology, Roorkee-247667 (India); Majumdar, P., E-mail: pmajum@barc.gov.in [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai-400085 (India); Kumar, Ravi, E-mail: ravikfme@iitr.ernet.in [Department of Mechanical and Industrial Engineering, Indian Institute of Technology, Roorkee-247667 (India); Chatterjee, B., E-mail: barun@barc.gov.in [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai-400085 (India); Gupta, Akhilesh, E-mail: akhilfme@iitr.ernet.in [Department of Mechanical and Industrial Engineering, Indian Institute of Technology, Roorkee-247667 (India); Lele, H.G., E-mail: hglele@barc.gov.in [Reactor Safety Division, Bhabha Atomic Research Centre, Mumbai-400085 (India)

    2013-01-15

    Highlights: Black-Right-Pointing-Pointer Circumferential temperature gradient for asymmetric heat-up was 400 Degree-Sign C. Black-Right-Pointing-Pointer At same pressure ballooning initiates at lower temperature in asymmetrical heat-up. Black-Right-Pointing-Pointer At 1 MPa ballooning initiated at 408 Degree-Sign C and with expansion rate of 0.005 mm/s. Black-Right-Pointing-Pointer At 2 MPa ballooning initiation at 330 Degree-Sign C and with expansion rate of 0.0056 mm/s. Black-Right-Pointing-Pointer For symmetrical heat-up strain rate was 10 times faster than asymmetric heat-up. - Abstract: In pressurized heavy water reactor (PHWR), under postulated scenario of small break Loss of Coolant Accident (LOCA) coincident with the failure of Emergency Core Cooling System (ECCS), a situation may arise under which reduction in mass flow rate of coolant through individual reactor channel can lead to stratified flow. Such stratified flow condition creates partial uncover of fuel bundle, which creates a circumferential temperature gradient over PT. The present investigation has been carried out to study thermo-mechanical behaviour of PT under asymmetric heating conditions for a 220 MWe PHWR. A 19-pin fuel simulator has been developed in which preferential heating of elements could be done by supplying power to the selected pins. The asymmetric heating of PT has been carried out at pressure 2 MPa and 1 MPa, respectively, by supplying power to upper region heating elements thus creating an half filled stratified flow conditions. The temperature difference up to 425 Degree-Sign C has been observed along top to bottom periphery of PT. A comparison is made between thermo-mechanical behaviour of PT under asymmetrical and symmetrical heat-up, expected from a large break LOCA condition. The radial expansion rate during symmetrical heating is found to be much faster as compared to that for asymmetric ballooning of PT at the same internal pressure. Integrity of PT is found to be

  2. Numerical simulation of flow and thermal field in supercritical pressure carbon dioxide flowing upward in a narrow tube

    International Nuclear Information System (INIS)

    A reliable heat transfer correlation valid at a supercritical pressure is indispensible for an accurate estimation of heat transfer in the sub-channel of a fuel assembly of Supercritical Water- Cooled Reactors (SCWR). Despite a number of supercritical heat transfer correlations having been proposed in the past several decades, a reliable one is still missing, since the predictions by the existing correlations show wide discrepancies from each other. In a mixed convection regime, no correlation is able to produce accurate predictions. Under the influence of strong buoyancy, the boundary layer structure is known to deform significantly, when the wall temperature is close to the pseudo-critical temperature; and, therefore, it is suspected to be one of the reasons for the enhancement or impairment of the heat transfer rate. However, a detailed analysis of the boundary-layer transformation process has never been successfully addressed, due partially to difficulty in experimenting at a condition of high pressure and temperature, and to an inadequacy of the numerical tools in dealing with substantial property variations. This paper provides results of the numerical analyses of flow and thermal field in CO2 flowing upward in a narrow tube. (author)

  3. Operational improvements in the CANDU 9 control centre

    International Nuclear Information System (INIS)

    AECL has adopted an evolutionary approach to the design of the CANDU 9 control centre. Several factors have contributed to this decision including the desire to build on the successes of the current generation of CANDU stations, the changing roles and responsibilities of operations staff, an improved understanding of human error in operational situations, the opportunity for improved plant performance through the introduction of new technologies, and evolving customer and regulatory requirements. Underlying this approach is a refined engineering design process that cost-effectively integrates operational feedback and human factors engineering to define the operating staff information and information presentation requirements. Based on this approach, the CANDU 9 control centre will provide utility operating staff with a layout and information organization that is better matched to operational tasks, thereby leading to reduced operations, maintenance and administration (OM and A) costs. Significant design features that contribute to the improved operational capabilities of the CANDU 9 control centre include: a control centre layout with improved functionality; a new Plant Display System that is separated from the digital control computer system; and an enhanced computerized reactor shutdown system. The paper will present a summary of the design process, a detailed description of the CANDU 9 control centre layout and features, a description of the plant control and display systems design, including findings from a regulatory review, and other improvements to enhance operability. (author)

  4. Status of R and D Activities related to Axial and Radial Creep of Pressure tubes of Heavy Water Reactors in India

    International Nuclear Information System (INIS)

    Scope of presentation: • R and D Strength of the organisation; • Brief introduction about PHWRs in India – Operating and under construction; • Issues associated with axial elongation and radial expansion due to neutron enhanced creep in pressure tube; • Status of work done till date; • Important Points related to present CRP

  5. Fluid-structure interaction in u-tube with surfce roughness and pressure drop

    International Nuclear Information System (INIS)

    In this research, the surface roughness affecting the pressure drop in a pipe used as the steam generator of a PWR was studied. Based on the CFD (Computational Fluid Dynamics) technique using a commercial code named ANSYS-FLUENT, a straight pipe was modeled to obtain the Darcy frictional coefficient, changed with a range of various surface roughness ratios as well as Reynolds numbers. The result is validated by the comparison with a Moody chart to set the appropriate size of grids at the wall for the correct consideration of surface roughness. The pressure drop in a full-scale U-shaped pipe is measured with the same code, correlated with the surface roughness ratio. In the next stage, we studied a reduced scale model of a U-shaped heat pipe with experiment and analysis of the investigation into fluid-structure interaction (FSI). The material of the pipe was cut from the real heat pipe of a material named Inconel 690 alloy, now used in steam generators. The accelerations at the fixed stations on the outer surface of the pipe model are measured in the series of time history, and Fourier transformed to the frequency domain. The natural frequency of three leading modes were traced from the FFT data, and compared with the result of a numerical analysis for unsteady, incompressible flow. The corresponding mode shapes and maximum displacement are obtained numerically from the FSI simulation with the coupling of the commercial codes, ANSYS-FLUENT and TRANSIENTSTRUCTURAL. The primary frequencies for the model system consist of three parts: structural vibration, BPF(blade pass frequency) of pump, and fluid-structure interaction.

  6. Bubble-assisted film evaporation correlation for saline water at sub-atmospheric pressures in horizontal-tube evaporator

    KAUST Repository

    Shahzad, Muhammad Wakil

    2013-01-01

    In falling film evaporators, the overall heat transfer coefficient is controlled by film thickness, velocity, liquid properties and the temperature differential across the film layer. This article presents the heat transfer behavior for evaporative film boiling on horizontal tubes, but working at low pressures of 0.93-3.60 kPa (corresponding solution saturation temperatures of 279-300 K) as well as seawater salinity of 15,000 to 90,000 mg/l or ppm. Owing to a dearth of literature on film-boiling at these conditions, the article is motivated by the importance of evaporative film boiling in the desalination processes such as the multi-effect distillation (MED) or multi-stage flashing (MSF): It is observed that in addition to the above-mentioned parameters, evaporative heat transfer of seawater is affected by the emergence of micro-bubbles within the thin film layer, particularly when the liquid saturation temperatures drop below 298 K (3.1 kPa). Such micro bubbles are generated near to the tube wall surfaces and they enhanced the heat transfer by two or more folds when compared with the predictions of conventional evaporative film boiling. The appearance of micro-bubbles is attributed to the rapid increase in the specific volume of vapor, i.e., dv/dT, at low saturation temperature conditions. A new correlation is thus proposed in this article and it shows good agreement to the measured data with an experimental uncertainty of 8% and regression RMSE of 3.5%. © 2012 Elsevier Ltd. All rights reserved.

  7. Study of the vortex-induced pressure excitation source in a Francis turbine draft tube by particle image velocimetry

    Science.gov (United States)

    Favrel, A.; Müller, A.; Landry, C.; Yamamoto, K.; Avellan, F.

    2015-12-01

    Francis turbines operating at part-load experience the development of a precessing cavitation vortex rope at the runner outlet, which acts as an excitation source for the hydraulic system. In case of resonance, the resulting pressure pulsations seriously compromise the stability of the machine and of the electrical grid to which it is connected. As such off-design conditions are increasingly required for the integration of unsteady renewable energy sources into the existing power system, an accurate assessment of the hydropower plant stability is crucial. However, the physical mechanisms driving this excitation source remain largely unclear. It is for instance essential to establish the link between the draft tube flow characteristics and the intensity of the excitation source. In this study, a two-component particle image velocimetry system is used to investigate the flow field at the runner outlet of a reduced-scale physical model of a Francis turbine. The discharge value is varied from 55 to 81 % of the value at the best efficiency point. A particular set-up is designed to guarantee a proper optical access across the complex geometry of the draft tube elbow. Based on phase-averaged velocity fields, the evolution of the vortex parameters with the discharge, such as the trajectory and the circulation, is determined for the first time. It is shown that the rise in the excitation source intensity is induced by an enlargement of the vortex trajectory and a simultaneous increase in the precession frequency, as well as the vortex circulation. Below a certain value of discharge, the structure of the vortex abruptly changes and loses its coherence, leading to a drastic reduction in the intensity of the induced excitation source.

  8. Twisted-tape-induced swirl flow heat transfer and pressure drop in a short circular tube under velocities controlled

    International Nuclear Information System (INIS)

    The twisted-tape-induced swirl flow heat transfer due to exponentially increasing heat inputs with various exponential periods (Q=Q0exp(t/τ), τ=7, 14 and 23 s) and the twisted-tape-induced pressure drop were systematically measured with mass velocities, G, ranging from 4022 to 15140 kg/m2s by an experimental water loop flow. Measurements were made on a 6 mm inner diameter, a 59.2 mm effective length and a 0.4 mm thickness of Platinum circular test tube which was spot-welded two potential taps on the outer surface of a 69.6 mm heated length. The twisted tapes with twist ratios, y [=H/d=(pitch of 180deg rotation)/d], of 2.39, 3.39 and 4.45 were used in this work. The relation between the swirl velocity and the pump input frequency and that between the fanning friction factor and Reynolds number (Red=2.04x104 to 9.96x104) were clarified. The twisted-tape-induced swirl flow heat transfers with y=2.39, 3.39 and 4.45 were compared with the values calculated by our correlation of the turbulent heat transfer for the empty tube and other worker's one for the circular tube with the twisted-tape-insert. The influence of y and Reynolds numbers based on swirl velocity, Resw, on the twisted-tape-induced swirl flow heat transfer was investigated into details and the widely and precisely predictable correlation of the twisted-tape-induced swirl flow heat transfer was derived based on the experimental data. The correlation can describe for the twisted-tape-induced swirl flow heat transfer for the wide ranges of twist ratios (y=2.39 to 4.45), mass velocities (G=4022 to 15140 kg/m2s) and Reynolds numbers based on swirl velocity (Resw=2.88x104 to 1.22x105) within 0 to +30% difference. (author)

  9. Twisted-tape-induced swirl flow heat transfer and pressure drop in a short circular tube under velocities controlled

    International Nuclear Information System (INIS)

    The twisted-tape-induced swirl flow heat transfer due to exponentially increasing heat inputs with various exponential periods (Q = Q0 exp(t/τ), τ = 7, 14 and 23 s) and the twisted-tape-induced pressure drop were systematically measured with mass velocities, G, ranging from 4022 to 15,140 kg/m2 s by an experimental water loop flow. Measurements were made on a 59.2 mm effective length which was spot-welded two potential taps on the outer surface of a 6 mm inner diameter, a 69.6 mm heated length and a 0.4 mm thickness of platinum circular test tube. The twisted tapes with twist ratios, y [=H/d = (pitch of 180° rotation)/d], of 2.39, 3.39 and 4.45 were used in this work. The relation between the swirl velocity and the pump input frequency and that between the fanning friction factor and Reynolds number (Red = 2.04 × 104 to 9.96 × 104) were clarified. The twisted-tape-induced swirl flow heat transfers with y = 2.39, 3.39 and 4.45 were compared with the values calculated by our correlation of the turbulent heat transfer for the empty tube and other worker's one for the circular tube with the twisted-tape insert. The influence of y and Reynolds numbers based on swirl velocity, Resw, on the twisted-tape-induced swirl flow heat transfer was investigated into details and the widely and precisely predictable correlation of the twisted-tape-induced swirl flow heat transfer was derived based on the experimental data. The correlation can describe for the twisted-tape-induced swirl flow heat transfer for the wide ranges of twist ratios (y = 2.39–4.45), mass velocities (G = 4022–15140 kg/m2 s) and Reynolds numbers based on swirl velocity (Resw = 2.88 × 104 to 1.22 × 105) within −10 to +30% difference.

  10. Verification of revised EOP using the CATHENA code for Wolsong NPP- Candu 6

    International Nuclear Information System (INIS)

    EOP (Emergency Operating Procedure) is the operating procedure document which describes the operator's action needed for the maintenance of nuclear power plant (NPP) safety when a certain accident takes place. EOPs of Wolsong NPP Units 2, 3, 4 (Candu 6) have been revised aiming at the optimization of procedure to recover the NPP from an accident and the minimization of human errors. The revised overall EOPs are composed of 2 EOPs describing the accident diagnosis and recovery of critical safety parameters and 13 EOPs describing the operator actions needed for the individual accidents which include LOCA and MSLB and so on. The purpose of the present study is to verify the revised EOPs if the procedures and operator actions described in each EOP are fully appropriate and effective for the safe shutdown of NPP leading to temperature low enough to start the long term cooling using the shutdown cooling system and for the maintenance of fuel and pressure boundary integrity. The verification was carried out with the deterministic safety analysis methodology using the CATHENA computer program which is a Thermalhydraulic numerical code used in the safety analysis of Candu 6 reactors. CATHENA can simulate the various thermalhydraulic behaviors which are expected to happen during each accident and by various operator actions. From the verification results it can be assured that the revised individual EOPs have the enough appropriateness with respect to the engineering aspects and effectiveness in view of fuel and pressure boundary integrity. (authors)

  11. Assessment of Welding System Modification of The Candu and PWR Fuel Element Types end Plug

    International Nuclear Information System (INIS)

    To anticipate future possibility of a nuclear fuel element industry in Indonesia, research on other types of nuclear fuel element beside Cirene type has to be done. It can be accomplished, one of them, by modifying the already available equipment. Based on the sheath material, the sheath dimension and the welding process parameters such as welding current and welding cycles, the available Magnetic Force Welding can be used for welding end plug of Candu nuclear fuel element by modifying some of its components (tube clamp, plug clamp, etc). The available Pellet drying and element filling furnace with its supporting system with includes helium gas filling, welding chamber, argon gas supply, vacuum system, sheath clamp and sheath driving system can be used for welding end plug with sheath of PWR nuclear fuel element by adding og Tungsten inert Gas (TIG) welding machine in the welding chamber and modifying a few components (seal clamp, sheath clamp)

  12. Hydrogen analysis using MELCOR at CANDU plant

    Energy Technology Data Exchange (ETDEWEB)

    Song, Sung Chu; Park, Jae Hong; Kim, Han Chul [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2011-10-15

    As a part of a submitted Wolsong Severe Accident Management Plan(SAMP) at the end of 2009, KINS needs to the review of it. In this the focus is on the hydrogen behavior during the station blackout(SBO) because the hydrogen explosion during severe accident in CANDU plants such as Wolsong units has been safety issue. A SBO occurs with failure of the emergency power system when loss of class IV electric power happens because the loss of class IV power is the most dominant internal event causing core damage for Wolsong units. And following a SBO event, most of the engineered safety features(ESFs), including hydrogen igniters, are inoperable except the passive systems such as Dousing systems. Hydrogen is generated due to Zr-steam reaction in fuel channels and core debris oxidation in the suspended debris beds, jet breakup of molten debris in the water pool of the reactor vault and molten core-concrete interaction (MCCI). A combustible gas control system consisting of Passive Autocatalytic Recombiner(PAR) is currently installed at Wolsong unit 1. The results of MELCOR analysis are presented

  13. Industrial process heat from CANDU reactors

    International Nuclear Information System (INIS)

    It has been demonstrated on a large scale that CANDU reactors can produce industrial process steam as well as electricity, reliably and economically. The advantages of cogeneration have led to the concept of an Industrial Energy Park adjacent to the Bruce Nuclear Power Development in the province of Ontario. For steam demands between 300,000 and 500,00 lb/h (38-63 kg/s) and an annual load factor of 80%, the estimated cost of nuclear steam at the Bruce site boundary is $3.21/MBtu ($3.04GJ), which is at least 30% cheaper than oil-fired steam at the same site. The most promising near term application of nuclear heat is likely to be found within the energy-intensive chemical industry. Nuclear energy can substitute for imported oil and coal in the eastern provinces if the price remains competitive, but low cost coal and gas in the western provinces may induce energy-intensive industries to locate near those sources of energy. In the long term it may be feasible to use nuclear heat for the mining and extraction of oil from the Alberta tar sands. (auth)

  14. Advanced CANDU reactor: an optimized energy source of oil sands application

    International Nuclear Information System (INIS)

    Atomic Energy of Canada Limited (AECL) is developing the ACR-700TM (Advanced CANDU Reactor-700TM) to meet customer needs for reduced capital cost, shorter construction schedule, high capacity factor while retaining the benefits of the CANDU experience base. The ACR-700 is based on the concept of CANDU horizontal fuel channels surrounded by heavy water moderator. The major innovation of this design is the use of slightly enriched uranium fuel in a CANFLEX bundle that is cooled by light water. This ensures: higher main steam pressures and temperatures providing higher thermal efficiency; a compact and simpler reactor design with reduced capital costs and shorter construction schedules; and reduced heavy water inventory compared to existing CANDU reactors. ACR-700 is not only a technically advanced and cost effective solution for electricity generating utilities, but also a low-cost, long-life and sustainable steam source for increasing Alberta's Oil Sand production rates. Currently practiced commercial surface mining and extraction of Oil Sand resources has been well established over the last three decades. But a majority of the available resources are somewhat deeper underground require in-situ extraction. Economic removal of such underground resources is now possible through the Steam Assisted Gravity Drainage (SAGD) process developed and proto-type tested in-site. SAGD requires the injection of large quantities of high-pressure steam into horizontal wells to form reduced viscosity bitumen and condensate mixture that is then collected at the surface. This paper describes joint AECL studies with CERI (Canadian Energy Research Institute) for the ACR, supplying both electricity and medium-pressure steam to an oil sands facility. The extensive oil sands deposits in northern Alberta are a very large energy resource. Currently, 30% of Canda's oil production is from the oil sands and this is expected to expand greatly over the coming decade. The bitumen deposits in the

  15. Ninth international conference on CANDU fuel, 'fuelling a clean future'

    International Nuclear Information System (INIS)

    The Canadian Nuclear Society's 9th International Conference on CANDU fuel took place in Belleville, Ontario on September 18-21, 2005. The theme for this year's conference was 'Fuelling a Clean Future' bringing together over 80 delegates ranging from: designers, engineers, manufacturers, researchers, modellers, safety specialists and managers to share the wealth of their knowledge and experience. This international event took place at an important turning point of the CANDU technology when new fuel design is being developed for commercial application, the Advanced CANDU Reactor is being considered for projects and nuclear power is enjoying a renaissance as the source energy for our future. Most of the conference was devoted to the presentation of technical papers in four parallel sessions. The topics of these sessions were: Design and Development; Fuel Safety; Fuel Modelling; Fuel Performance; Fuel Manufacturing; Fuel Management; Thermalhydraulics; and, Spent Fuel Management and Criticalty

  16. Evolution of the CANDU control centre retrofit and new stations

    International Nuclear Information System (INIS)

    Significant event data from operating nuclear plants in many countries consistently indicates human errors are the root cause for 40-60% of operating station significant events. Because so much information is already in digital form, opportunities exist to improve the CANDU control centre with retrofits that exploit this information. These opportunities are enhanced because of rapid technological development in computers and electronics, coupled with significant progress in the behavioural sciences that greatly increases our knowledge of the cognitive strengths and weaknesses of human beings. CANDU control rooms are undergoing retrofits and for future CANDU stations, a new concept of the control centre is emerging. The objective is to significantly reduce the incidence of human error, reduce operations and maintenance costs and improve both reliability and safety

  17. Distinctive safety aspects of the CANDU-PHW reactor design

    International Nuclear Information System (INIS)

    Two lectures are presented in this report. They were prepared in response to a request from IAEA to provide information on the 'Special characteristics of the safety analysis of heavy water reactors' to delegates from member states attending the Interregional Training Course on Safety Analysis Review, held at Karlsruhe, November 19 to December 20, 1979. The CANDU-PHW reactor is used as a model for discussion. The first lecture describes the distinctive features of the CANDU reactor and how they impact on reactor safety. In the second lecture the Canadian safety philosophy, the safety design objective, and other selected topics on reactor safety analysis are discussed. The material in this report was selected with a view to assisting those not familiar with the CANDU heavy water reactor design in evaluating the distinctive safety aspects of these reactors. (auth)

  18. Reactor Physics Analysis Models for a CANDU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Hang Bok

    2007-10-15

    Canada deuterium uranium (CANDU) reactor physics analysis is typically performed in three steps. At first, macroscopic cross-sections of the reference lattice is produced by modeling the reference fuel channel. Secondly macroscopic cross-sections of reactivity devices in the reactor are generated. The macroscopic cross-sections of a reactivity device are calculated as incremental cross-sections by subtracting macroscopic cross-sections of a three-dimensional lattice without reactivity device from those of a three-dimensional lattice with a reactivity device. Using the macroscopic cross-sections of the reference lattice and incremental cross-sections of the reactivity devices, reactor physics calculations are performed. This report summarizes input data of typical CANDU reactor physics codes, which can be utilized for the future CANDU reactor physics analysis.

  19. Investigation of weldability and property changes of high pressure heat-resistant cast stainless steel tubes used in pyrolysis furnaces after a five-year service

    International Nuclear Information System (INIS)

    Highlights: → To investigate the weldability and property changes of high pressure heat-resistant cast stainless steel (HP) tubes. → Welding was done by gas-tungsten arc welding (GTAW) process. → Composition of precipitates was characterized by means of SEM and EDS analyses. → The solution treatment was used to recover the properties of tubes. → To investigate mechanical strength of specimens, tensile tests were carried out at room temperature. -- Abstract: High pressure heat-resistant cast stainless steel (HP steel) tubes produced by centrifugal casting are used in petrochemical industries for pyrolysis furnaces. They have appropriate ductility and weldability in as-cast conditions. These steels lose their ductility and weldability after being used in service and, hence, require repair. In the present study, the effect of metallurgical changes on weldability and ductility was investigated. The life span of the studied tubes was 5 years. Using electrodes with a chemical composition close to the base metal analysis, welding was done by gas-tungsten arc welding (GTAW) process. Solution treatment was used to recover the properties of tubes which can be useful, depending on metallurgical changes.

  20. Flaw detecting method for welded portion between lower end plate and stab tube of reactor pressure vessel and liquid medium-filling device used for the method

    International Nuclear Information System (INIS)

    The present invention provides a method of reliably performing an ultrasonic flaw detecting test for a welded portion between a lower end plate and a stab tube of a reactor pressure vessel. Namely, a liquid medium is filled into a space formed between the outer circumference of a housing of a driving device, and an inner surface of a driving device-insertion hole and the stab tube. Ultrasonic waves are appropriately transferred to the welded portion by means of the liquid medium. Accordingly, the ultrasonic test can reliably be performed for the welded portion between the lower end plate and the stab tube. In addition, the housing of the driving device is coated at a portion where it is situated to the outer side of the main body of the pressure vessel. The liquid medium is continuously supplied from a medium supply device into the inside of the main body of the filling device. The liquid medium is filled into the space formed by the outer circumference of the housing of the driving device, and the inner surface of the insertion hole of the driving device and the stab tube. Accordingly, ultrasonic test can reliably performed for the welded portion between the lower end plate and the stab tube. (I.S.)

  1. Conference proceedings of the 4. international conference on CANDU fuel. V. 1,2

    International Nuclear Information System (INIS)

    These proceedings contain the full texts of all 65 papers presented at the 4th International Conference on CANDU fuel. As such, they represent an update on the state-of-the-art in such important CANDU fuel topics as International Development Programs and Operating Experience with CANDU fuel, Performance Assessments and Fuel Behavior Modeling, Fuel Properties, Licensing and Accident Analyses for CANDU fuel, Design, Testing and Manufacturing, and Advanced Fuel Designs. The large number of papers required the use of parallel sessions for the first time at a CANDU Fuel Conference

  2. Heavy water detritiation to support CANDU station maintenance activities

    International Nuclear Information System (INIS)

    Heavy water and tritium control are important aspects of CANDU operation and have a major impact on station maintenance activities. Station personnel are trained to understand the importance of heavy water management and the economics and environmental impact of tritiated heavy water losses. This paper discusses new GE technology that can now make a major improvement in CANDU maintenance activities through significant reductions in station tritium levels. Tritium is of particular concern in the CANDU industry given the nature of heavy water reactors to build up high levels of tritium over time. High tritium levels in the reactor vault significantly slow down maintenance activities in the reactor vault due to the requirement for personnel protective equipment, including breathing apparatus and cumbersome plastic air suits. The difficulties increase as reactors age and tritium levels increase. Building upon GE's extensive operational experience in tritium management in CANDU reactors and its own tritium handling facility, GE. has developed a new large-scale diffusion based isotope separation process as an alternative to conventional cryogenic distillation. Having a tritium inventory an order of magnitude lower than conventional cryogenic distillation, this process is very attractive for heavy water detritiation, and applicable to single and multi-unit CANDU stations. This new process can now provide a step change reduction in CANDU heavy water tritium levels resulting in reduced environmental emissions and lowering reactor vault tritium MPC(a) levels. Reactor vault tritium can be reduced sufficiently for maintenance activities to be done without plastic suits, leading to shorter outages, improved station capacity factors, and improved station economics. (author)

  3. Ear Tubes

    Science.gov (United States)

    ... of the ear drum or eustachian tube, Down Syndrome, cleft palate, and barotrauma (injury to the middle ear caused by a reduction of air pressure, ... specialist) may be warranted if you or your child has experienced repeated ... fluid in the middle ear, barotrauma, or have an anatomic abnormality that ...

  4. Evolution of the CANDU control centre design process

    International Nuclear Information System (INIS)

    The design of the CANDU NPP control centre and the associated control centre design process has evolved considerably over several generations of plants, from Douglas Point through Darlington, and beyond, to new designs like CANDU 3. In the early plants, the control centre configuration had to be based on designers' projections of control interface requirements. With succeeding generations of designs, along with the introduction of advancing computer control technology, a larger based of operational experience has been factored into the control interface design, and increasing attention has been given to more formal requirements definition, and more systematic consideration of human factors aspects of the design

  5. CANDU fuel quality and how it is achieved

    International Nuclear Information System (INIS)

    In this three part presentation CANDU fuel quality is reviewed from the point of view of a designer/operator and a fabricator. In Part 'A' fuel performance and quality considerations are discussed from the point of view of a designer-operator. In Parts 'B' and 'C' fuel quality is reviewed from the point of view of a fabricator. The presentation was divided in this way to convey the 'team effort' attitude which exists in the Canadian program; the team effort which is an essential part of the CANDU story. (auth)

  6. Fuel management simulation for CANFLEX-RU in CANDU 6

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Chang Joon; Suk, Ho Chun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    Fuel management simulations have been performed for CANFLEX-09% RU fuel in the CANDU 6 reactor. In this study, the bi-directional 4-bundle shift fuelling scheme was assumed. The lattice cell and time-average calculation were carried out. The refuelling simulation calculations were performed for 600 full power days. Time-averaged results show good axial power profile with the CANFLEX-RU fuel. During the simulation period, the maximum channel and bundle power were maintained below the licensing limit of CANDU 6 reactor. 7 refs., 4 figs. (Author)

  7. Pressure drop and heat transfer in the sodium to air heat exchanger tube banks on advanced sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    A numerical study was performed to investigate the thermal and hydraulic characteristics and build up design model of the AHX (sodium-to-air heat exchanger) unit of a sodium-cooled fast reactor. Helical-coiled tube banks in the AHX were modeled as porous media and simulated heat and momentum transfer. Two-dimensional flow characteristic appeared at the most region of AHX annulus. Pressure drop and heat transfer coefficient for rectangular, parallelogram and staggered tube banks as the main components of the AHX were evaluated and compared with Zhukauskas empirical correlations. (author)

  8. A comparison of the void reactivity effect between the CANDU standard and CANDU-6 SEU-43 cells

    International Nuclear Information System (INIS)

    In a CANDU type reactor the void coefficient of the reactivity is positive. The experimental data are available only for fresh fuel in cold conditions. On the other hand, taking into account the reactivity effects induced by changes of the coolant properties is often difficult. The safety analyses require an estimation of the calculation error. A comparison between models is an usual approach to obtain detailed information. In our paper a heterogeneous multi-stratified coolant model is used both for the CANDU standard fuel assembly cell and CANDU SEU-43 cell concept. The coolant is treated as a two phase (liquid and vapors) medium gravitationally separated. The results are inter-compared for different burnups in the partial or total void cases. (authors)

  9. An Effective Approach for Coupling Direct Analysis in Real Time with Atmospheric Pressure Drift Tube Ion Mobility Spectrometry

    Science.gov (United States)

    Keelor, Joel D.; Dwivedi, Prabha; Fernández, Facundo M.

    2014-09-01

    Drift tube ion mobility spectrometry (DTIMS) has evolved as a robust analytical platform routinely used for screening small molecules across a broad suite of chemistries ranging from food and pharmaceuticals to explosives and environmental toxins. Most modern atmospheric pressure IM detectors employ corona discharge, photoionization, radioactive, or electrospray ion sources for efficient ion production. Coupling standalone DTIMS with ambient plasma-based techniques, however, has proven to be an exceptional challenge. Device sensitivity with near-ground ambient plasma sources is hindered by poor ion transmission at the source-instrument interface, where ion repulsion is caused by the strong electric field barrier of the high potential ion mobility spectrometry (IMS) inlet. To overcome this shortfall, we introduce a new ion source design incorporating a repeller point electrode used to shape the electric field profile and enable ion transmission from a direct analysis in real time (DART) plasma ion source. Parameter space characterization studies of the DART DTIMS setup were performed to ascertain the optimal configuration for the source assembly favoring ion transport. Preliminary system capabilities for the direct screening of solid pharmaceuticals are briefly demonstrated.

  10. Steam drum level dynamics in a multiple loop natural circulation system of a pressure-tube type BWR

    International Nuclear Information System (INIS)

    Advanced Heavy Water Reactor (AHWR) is a pressure tube type boiling water reactor employing natural circulation as the mode of heat removal under all the operating conditions. Main Heat Transport System (MHTS) of AHWR is essentially a multi-loop natural circulation system with all the loops connected to each other. Each loop of MHTS has a steam drum that provides for gravity based steam-water separation. Steam drum level is a very critical parameter especially in multi-loop natural circulation systems as large departures from the set point may lead to ineffective separation of steam-water or may affect the driving head. However, such a system is susceptible to steam drum level anomalies under postulated asymmetrical operating conditions among the different quadrants of the core like feedwater flow distribution anomaly among the steam drums or power anomaly among the core quadrants. Analyses were carried out to probe such scenarios and unravel the underlying dynamics of steam drum level using system code RELAP5/Mod3.2. In addition, a scheme to obviate such problem in a passive manner without dependence on level controller was examined. It was concluded that steam drums need to be connected in the liquid as well as steam space to make the system tolerant to asymmetrical operating conditions.

  11. Influence of Fe content on corrosion and hydrogen pick up behavior of Zr–2.5Nb pressure tube material

    Energy Technology Data Exchange (ETDEWEB)

    Choudhuri, Gargi, E-mail: gargi@barc.gov.in [Quality Assurance Division, BARC, Mumbai 400 085 (India); Jagannath [Theoretical Physics Division, BARC, Mumbai 400 085 (India); Kiran Kumar, M.; Kain, V.; Srivastava, D. [Material Science Division, BARC, Mumbai 400 085 (India); Basu, S. [Solid State Physics Division, BARC, Mumbai 400 085 (India); Shah, B.K. [Quality Assurance Division, BARC, Mumbai 400 085 (India); Saibaba, N. [Nuclear Fuel Complex, Hyderabad 500 062 (India); Dey, G.K. [Material Science Division, BARC, Mumbai 400 085 (India)

    2013-10-15

    The effects of Fe addition in the range of 300–1250 ppm in cold worked stress-relieved Zr–2.5Nb pressure tube on oxidation and hydrogen pick up behavior have been studied after 415 °C steam autoclaving. Microstructure and micro-chemistry of second phase and precipitates were characterized using electron microscope. Addition of 800 ppm Fe in Zr–2.5Nb alloy led to better oxidation resistance. With further addition of Fe no significant improvement of oxidation resistance was observed but hydrogen-pickup was found to increase. Zr–Nb–Fe bearing precipitates were observed in Zr–2.5Nb alloy containing 800 ppm Fe. Further addition of Fe led to formation of Zr–Fe intermetallic. The chemical state of oxide has been determined by X-ray photo electron spectroscopy. Grazing Incidence X-ray Diffraction revealed that oxide in alloys with higher Fe, contained a higher fraction of tetragonal-Zirconia which is indicative of a protective oxide film and hence better oxidation resistance of the alloy.

  12. The influence of endotracheal tube cuff pressures to PEEP values in mechanical ventilated patients%机械通气患者套囊压力的监测

    Institute of Scientific and Technical Information of China (English)

    仇成秀; 刘志梅; 钟琼

    2011-01-01

    目的 讨论机械通气患者气管插管导管、气管切开套管的套囊压力和注气量是否合适,提供正确给套囊注气的科学依据.方法 对35例气管插管、气管切开进行机械通气患者的气管套管套囊压力和注气量的实际值和理想值进行准确测量.结果 65%的患者气管套管套囊实际压力和注气量过高,大于理想值.其中套囊实际注气量大于理想注气量2~4ml,套囊压力超过理想压力2~26 cm H2O.结论 临床上大部分气管套管套囊压力和注气量偏高,因此,应对人工气道患者采用专用套囊测压仪指导套囊注气量及控制囊内压,最大限度地避免气道黏膜的损伤.%Objective In cases of tracheal intubation of patients with a mechanical ventilation tube,measure trachootomy tube cuff pressure and the injection of gas whether it was providing appropriate pressure under the conditions of the use of gas injection basis.Methods In 35 cases of tracheal intubation,measure the endotracheal tube cuff pressure of mechanical ventilation patients with tracheotomy and note the actual value and the ideal gas value of accurate measurement.Results About 65% of patients with tracheal tube cuff pressure and the actual injection gas were too high,greater than the ideal value.One cuff was greater than the actual injection gas ideal gas injection rate 2 ~4 ml,the pressure cuff pressure above the ideal of 2 ~ 26 cm H2 O.Conclusion Most of the clinical tracheal tube cuff pressure indicates too high gas injection,so doctors should use a dedicated artificial airway cuff pressure measuring instrument to guide and control cuff intracystic injection gas pressure in order to avoid possible damage of airway mucosa.

  13. Effect of tube size on electromagnetic tube bulging

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

    The commercial finite code ANSYS was employed for the simulation of the electromagnetic tube bulging process. The finite element model and boundary conditions were thoroughly discussed. ANSYS/EMAG was used to model the time varying electromagnetic field in order to obtain the radial and axial magnetic pressure acting on the tube. The magnetic pressure was then used as boundary conditions to model the high velocity deformation of various length tube with ANSYS/LSDYNA. The time space distribution of magnetic pressure on various length tubes was presented. Effect of tube size on the distribution of radial magnetic pressure and axial magnetic pressure and high velocity deformation were discussed. According to the radial magnetic pressure ratio of tube end to tube center and corresponding dimensionless length ratio of tube to coil, the free electromagnetic tube bulging was studied in classification. The calculated results show good agreements with practice.

  14. Candu plant life management - safeguarding the investment

    International Nuclear Information System (INIS)

    With a large number of the CANDU NPPs getting to the mid-point of their design life, a number of programs have been initiated by the Utilities and AECL to proactively manage plant aging, hence ensuring a safe and reliable operation for the remaining design life. An integrated aging management program is being formulated in response to the regulatory authority the Atomic Energy Control Board (AECB). This integrated program will take into account safety, performance and economic requirements. Elements of this program cover a variety of areas; Phase 1 covers life assessment studies of the critical systems, structures and components (SSCs) including a methodology for defining the critical SSCs. Other programs within Phase 2 address the preparation of a unified set of industry guidelines covering maintenance and inspection requirements during early, middle and late years of plant operation. A 'technology watch' program has also been initiated. The objective of this program is to identify far in advance potential aging phenomena and failure modes that could affect plant performance along with the inspection and maintenance required to monitor and mitigate such aging effects. The importance of maintaining the plant within a well-established 'licensing basis' envelope is also discussed. Plant licensing is carried out for an initial set of conditions and equipment status which may vary during plant life. Canadian utilities are required by the licensing authority, the AECB, to assess the impact of aging on the licensing case for each plant in operation. So far, the main focus of these assessments has been the impact of aging on key parameters such as fuel thermal margin and containment leak rate. Ongoing activities covering characterization of aging, maintenance, monitoring, and appropriate refurbishments are examined. Emphasis of R and D performed to characterize degradation mechanisms, support inspection and fitness for service guidelines are also briefly outlined. (author)

  15. [Variations in the internal pressure of the pneumatic cuffs of endotracheal tubes according to their contents and the anesthetic mixtures used. Experimental study].

    Science.gov (United States)

    de Santos, P; Castillo, J; Bogdanovich, A; Nalda, M A

    1989-01-01

    With the purpose of measuring pressure changes in the pneumatic cuffs of endotracheal tubes when the composition of the mixture of gases used for ventilation had to change for the same content, we designed a model of artificial respiration that consisted of a tube with a low pressure pneumatic cuff measuring 8.5 mm in inner diameter introduced in a replica of a human trachea, adjusted to two anesthetic bags. The cuff valve was connected to a pressure transducer by a three-ended stopcock and, after aspiration of its content, it was inflated with air, saline or nitrous oxide and oxygen at 60% up to a basal pressure of 20 mmHg. The tube was connected to a respirator adjusted to inflate 10 l/min at a rate of 15 insufflations/min of: oxygen 100% for 5 minutes, then nitrous oxide and oxygen at 60% for 30 minutes and oxygen 100% again for 15 minutes. When inflating the pneumatic cuff with air and ventilating with nitrous oxide and oxygen at 60%, its pressure reached a maximum mean value of 58 mmHg (190% with respect to base values). When insufflating with saline and ventilating in the same conditions, pressure reached a maximum mean value of 33 mmHg (65% with respect to base values). When the pneumatic cuff was inflated with nitrous oxide and oxygen at 60%, important changes in pressure were observed when the characteristics of the inspired gases were modified. We conclude that some method for monitoring pneumatic cuff pressure should be systematized.

  16. The results from the second high-pressure melt ejection test completed in the Molten Fuel Moderator Interaction Facility at Chalk River Laboratories

    Energy Technology Data Exchange (ETDEWEB)

    Nitheanandan, T.; Kyle, G.; O' Connor, R. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2007-09-15

    The Canadian nuclear power generation industry, represented by the CANDU Owners Group (COG), is funding an experimental program at Chalk River Laboratories to study the interaction between the molten material ejected from the fuel channel and the moderator. These experiments are designed to address one of the very low probability postulated accident events considered for CANDU Pressurized Heavy Water Reactors (PHWRs), where an array of fuel channels contain the nuclear fuel and high-temperature, high-pressure coolant. Under severely restricted flow blockage conditions postulated in a fuel channel, the temperature excursion could result in fuel melting, consequential failure of the fuel channel, and ejection of the molten fuel at high pressures into the heavy water moderator at near atmospheric pressure. The objective of the experimental program is to demonstrate that a highly energetic Molten Fuel Moderator Interaction (MFMI) and associated high-pressure pulse can be ruled out. The second high-pressure melt ejection test using 22 kg of prototypical corium was completed recently at Chalk River Laboratories. The second test consisted of heating a thermite mixture of U, U{sub 3}O{sub 8}, Zr, and CrO{sub 3}, simulating the molten material expected in a fuel channel, inside a 1 m length of insulated pressure tube. Once the molten material reached the desired temperature of {approx}2400{sup o}C, the molten material was ejected into the surrounding tank of 63{sup o}C water. At the time of melt ejection, the static pressure in the test section was 3.35 MPa. The confinement vessel pressure reached a peak volume of 201 kPa following the rupture of the test section. The peak dynamic pressure measured on the inner vessel walls ranged between 0.7 MPa and 1 MPa. The dynamic pressure history, debris size, and the effects of the material interacting with tubes representing neighbouring fuel channels were investigated. (author)

  17. The results from the second high-pressure melt ejection test completed in the Molten Fuel Moderator Interaction Facility at Chalk River Laboratories

    International Nuclear Information System (INIS)

    The Canadian nuclear power generation industry, represented by the CANDU Owners Group (COG), is funding an experimental program at Chalk River Laboratories to study the interaction between the molten material ejected from the fuel channel and the moderator. These experiments are designed to address one of the very low probability postulated accident events considered for CANDU Pressurized Heavy Water Reactors (PHWRs), where an array of fuel channels contain the nuclear fuel and high-temperature, high-pressure coolant. Under severely restricted flow blockage conditions postulated in a fuel channel, the temperature excursion could result in fuel melting, consequential failure of the fuel channel, and ejection of the molten fuel at high pressures into the heavy water moderator at near atmospheric pressure. The objective of the experimental program is to demonstrate that a highly energetic Molten Fuel Moderator Interaction (MFMI) and associated high-pressure pulse can be ruled out. The second high-pressure melt ejection test using 22 kg of prototypical corium was completed recently at Chalk River Laboratories. The second test consisted of heating a thermite mixture of U, U3O8, Zr, and CrO3, simulating the molten material expected in a fuel channel, inside a 1 m length of insulated pressure tube. Once the molten material reached the desired temperature of ∼2400oC, the molten material was ejected into the surrounding tank of 63oC water. At the time of melt ejection, the static pressure in the test section was 3.35 MPa. The confinement vessel pressure reached a peak volume of 201 kPa following the rupture of the test section. The peak dynamic pressure measured on the inner vessel walls ranged between 0.7 MPa and 1 MPa. The dynamic pressure history, debris size, and the effects of the material interacting with tubes representing neighbouring fuel channels were investigated. (author)

  18. Chemical aspects of hydrogen ingress in zirconium and zircaloy pressure tubes: ageing management of Indian PHWR coolant channels - determination of hydrogen and deuterium

    International Nuclear Information System (INIS)

    Pressurized heavy water reactors (PHWRs) use zirconium and zirconium based alloys as clad and coolant tubes since its beginning. The first ever zircaloy-2 pressure tube failure occurred in 1983 at Ontario Hydro's Pickering Unit 2 in Canada which necessitated a thorough examination of causes of such failure. The failure was attributed to massive hydriding at the failed spot of pressure tube. Continuous usage of zirconium alloys could result in their hydrogen and deuterium pick-up leading to hydrogen/ deuterium embrittlement. The life of the zircaloy coolant channels is dictated by hydrogen/deuterium content and hence ageing management of the pressure tubes is essential for ensuring their trouble-free usage. It is desirable to have a sound knowledge on the chemical aspects of zirconium and zirconium based alloys metallurgy, the mechanistic principles of hydrogen ingress into the pressure tubes during in reactor service, and identifying suitable analytical methodologies for precise and accurate determination of hydrogen in wafer thin sliver samples carved out from insides of pressure tubes without causing any structural damage so that it can continue to remain in service. This is desirable so that the ageing management does not result in cost-escalation. This report is divided in to three main parts. The first part deals with the chemical aspects of zirconium and zirconium based alloy metallurgy, the mechanism of hydrogen pick-up and hydride formation in zirconium matrix. The second part describes various methodologies and their limitations, available for hydrogen/deuterium determination. The third part deals in detail, about the extensive investigations carried out at Radioanalytical Chemistry Division (RACD) in Radiochemistry and Isotope Group for establishing an indigenously developed hot vacuum extraction system in combination with quadrupole mass spectrometry for precise determination of hydrogen and deuterium in wafer thin sliver sample of zircaloy. The

  19. Ludwig: A Training Simulator of the Safety Operation of a CANDU Reactor

    Directory of Open Access Journals (Sweden)

    Gustavo Boroni

    2011-01-01

    Full Text Available This paper presents the application Ludwig designed to train operators of a CANDU Nuclear Power Plant (NPP by means of a computer control panel that simulates the response of the evolution of the physical variables of the plant under normal transients. The model includes a close set of equations representing the principal components of a CANDU NPP plant, a nodalized primary circuit, core, pressurizer, and steam generators. The design of the application was performed using the object-oriented programming paradigm, incorporating an event-driven process to reflect the action of the human operators and the automatic control system. A comprehensive set of online graphical displays are provided giving an in-depth understanding of transient neutronic and thermal hydraulic response of the power plant. The model was validated against data from a real transient occurring in the Argentine NPP Embalse Río Tercero, showing good agreement. However, it should be stressed that the aim of the simulator is in the training of operators and engineering students.

  20. Overall heat transfer coefficient and pressure drop in a typical tubular exchanger employing alumina nano-fluid as the tube side hot fluid

    Science.gov (United States)

    Kabeel, A. E.; Abdelgaied, Mohamed

    2016-08-01

    Nano-fluids are used to improve the heat transfer rates in heat exchangers, especially; the shell-and-tube heat exchanger that is considered one of the most important types of heat exchangers. In the present study, an experimental loop is constructed to study the thermal characteristics of the shell-and-tube heat exchanger; at different concentrations of Al2O3 nonmetallic particles (0.0, 2, 4, and 6 %). This material concentrations is by volume concentrations in pure water as a base fluid. The effects of nano-fluid concentrations on the performance of shell and tube heat exchanger have been conducted based on the overall heat transfer coefficient, the friction factor, the pressure drop in tube side, and the entropy generation rate. The experimental results show that; the highest heat transfer coefficient is obtained at a nano-fluid concentration of 4 % of the shell side. In shell side the maximum percentage increase in the overall heat transfer coefficient has reached 29.8 % for a nano-fluid concentration of 4 %, relative to the case of the base fluid (water) at the same tube side Reynolds number. However; in the tube side the maximum relative increase in pressure drop has recorded the values of 12, 28 and 48 % for a nano-material concentration of 2, 4 and 6 %, respectively, relative to the case without nano-fluid, at an approximate value of 56,000 for Reynolds number. The entropy generation reduces with increasing the nonmetallic particle volume fraction of the same flow rates. For increase the nonmetallic particle volume fraction from 0.0 to 6 % the rate of entropy generation decrease by 10 %.

  1. Safety and reliability of the sealing cuff pressure of the Microcuff pediatric tracheal tube for prevention of post-extubation morbidity in children: A comparative study

    Directory of Open Access Journals (Sweden)

    Roshdi Roshdi Al-Metwalli

    2014-01-01

    Full Text Available Objectives: The objective of this study is to evaluate the efficacy and safety of sealing pressure as an inflation technique of the Microcuff pediatric tracheal cuffed tube. Materials and Methods: A total of 60 children were enrolled in this study. After induction of anesthesia and intubation with Microcuff pediatric tracheal tube, patients were randomly assigned, to one of the three groups. Control group (n = 20 the cuff was inflated to a cuff pressure of 20 cm H 2 O; sealing group (n = 20 the cuff was inflated to prevent the air leak at peak airway pressure of 20 cm H 2 O and the finger group (n = 20 the cuff was inflated to a suitable pressure using the finger estimation. Tracheal leak, incidence and severity of post-extubation cough, stridor, sore throat and hoarseness were recorded. Results: The cuff pressure as well as the volume of air to fill the cuff was significantly low in the sealing group when compared with the control group (P < 0.001; however, their values were significantly high in the finger group compared with both the control and the sealing group (P < 0.001. The incidence and severity of sore throat were significantly high in the finger group compared with both the control and the sealing group (P = 0.0009 and P = 0.0026. Three patients in the control group developed air leak around the endotracheal tube cuff. The incidence and severity of other complications were similar in the three groups. Conclusion: In pediatric N 2 O, free general anesthesia using Microcuff pediatric tracheal tub, sealing cuff pressure is safer than finger palpation technique regarding post-extubation morbidities and more reliable than recommended safe pressure in prevention of the air leak.

  2. Proceedings of the fourth international conference on CANDU maintenance

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-07-01

    These proceedings record the information presented at the 4th International Conference on CANDU Maintenance held November 16-18,1997 in Toronto, Canada. The papers for these proceedings were prepared on component maintenance, human performance, steam generator leak detection, fuel channel inspections, rotating equipment maintenance, surveillance programs, inspection techniques, valve maintenance, steam generator repairs and performance, reactor aging management and preventative maintenance.

  3. Proceedings of the fourth international conference on CANDU maintenance

    International Nuclear Information System (INIS)

    These proceedings record the information presented at the 4th International Conference on CANDU Maintenance held November 16-18,1997 in Toronto, Canada. The papers for these proceedings were prepared on component maintenance, human performance, steam generator leak detection, fuel channel inspections, rotating equipment maintenance, surveillance programs, inspection techniques, valve maintenance, steam generator repairs and performance, reactor aging management and preventative maintenance

  4. The Candu system - The way for nuclear autonomy

    International Nuclear Information System (INIS)

    The experience acquired by Canada during the development of Candu System is presented. Some basic foundations of technology transfer are defined and, the conditions of canadian nuclear industry to provide developing countries, technical assistence for acquisition of nuclear energy autonomy, are analysed. (M.C.K.)

  5. Plutonium Consumption Program, CANDU Reactor Project final report

    Energy Technology Data Exchange (ETDEWEB)

    1994-07-31

    DOE is investigating methods for long term dispositioning of weapons grade plutonium. One such method would be to utilize the plutonium in Mixed OXide (MOX) fuel assemblies in existing CANDU reactors. CANDU (Canadian Deuterium Uranium) reactors are designed, licensed, built, and supported by Atomic Energy of Canada Limited (AECL), and currently use natural uranium oxide as fuel. The MOX spent fuel assemblies removed from the reactor would be similar to the spent fuel currently produced using natural uranium fuel, thus rendering the plutonium as unattractive as that in the stockpiles of commercial spent fuel. This report presents the results of a study sponsored by the DOE for dispositioning the plutonium using CANDU technology. Ontario Hydro`s Bruce A was used as reference. The fuel design study defined the optimum parameters to disposition 50 tons of Pu in 25 years (or 100 tons). Two alternate fuel designs were studied. Safeguards, security, environment, safety, health, economics, etc. were considered. Options for complete destruction of the Pu were also studied briefly; CANDU has a superior ability for this. Alternative deployment options were explored and the potential impact on Pu dispositioning in the former Soviet Union was studied. An integrated system can be ready to begin Pu consumption in 4 years, with no changes required to the reactors other than for safe, secure storage of new fuel.

  6. A short history of the CANDU nuclear power system

    International Nuclear Information System (INIS)

    This paper provides a short historical summary of the evolution of the CANDU nuclear power system with emphasis on the roles played by Ontario Hydro and private sector companies in Ontario in collaboration with Atomic Energy of Canada Limited (AECL). (author). 1 fig., 61 refs

  7. A neutronic study of the cycle PWR-CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alberto da; Pereira, Claubia; Veloso, Maria Auxiliadora Fortini; Fortini, Angela; Pinheiro, Ricardo Brant [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear]. E-mail: albertomoc@terra.com.br; claubia@nuclear.ufmg.br; dora@nuclear.ufmg.br; fortini@nuclear.ufmg.br; rbp@nuclear.ufmg.br

    2007-07-01

    The cycle PWR-CANDU was simulated using the WIMSD-5B and ORIGEN2.1 codes. It was simulated a fuel burnup of 33,000 MWd/t for UO{sub 2} with enrichment of 3.2% and a fuel extended burnup of 45,000 MWd/t for UO{sub 2} with enrichments of 3.5%, 4.0% and 5.0% in a PWR reactor. The PWR discharged fuel was submitted to the simulation of deposition for five years. After that, it was submitted to AYROX reprocessing and used to produce a fuel to CANDU reactor. Then, it was simulated the burnup in the CANDU. Parameters such as infinite medium multiplication factor, k{sub inf}, fuel temperature coefficient of reactivity, {alpha}{sub TF}, moderator temperature coefficient of reactivity, {alpha}{sub TM}, the ratio rapid flux/total flux and the isotopic composition in the begin and the end of life were evaluated. The results showed that the fuels analyzed could be used on PWR and CANDU reactors without the need of change on the design of these reactors. (author)

  8. Plutonium Consumption Program, CANDU Reactor Project final report

    International Nuclear Information System (INIS)

    DOE is investigating methods for long term dispositioning of weapons grade plutonium. One such method would be to utilize the plutonium in Mixed OXide (MOX) fuel assemblies in existing CANDU reactors. CANDU (Canadian Deuterium Uranium) reactors are designed, licensed, built, and supported by Atomic Energy of Canada Limited (AECL), and currently use natural uranium oxide as fuel. The MOX spent fuel assemblies removed from the reactor would be similar to the spent fuel currently produced using natural uranium fuel, thus rendering the plutonium as unattractive as that in the stockpiles of commercial spent fuel. This report presents the results of a study sponsored by the DOE for dispositioning the plutonium using CANDU technology. Ontario Hydro's Bruce A was used as reference. The fuel design study defined the optimum parameters to disposition 50 tons of Pu in 25 years (or 100 tons). Two alternate fuel designs were studied. Safeguards, security, environment, safety, health, economics, etc. were considered. Options for complete destruction of the Pu were also studied briefly; CANDU has a superior ability for this. Alternative deployment options were explored and the potential impact on Pu dispositioning in the former Soviet Union was studied. An integrated system can be ready to begin Pu consumption in 4 years, with no changes required to the reactors other than for safe, secure storage of new fuel

  9. Assessment of LOCA with loss of class IV power for CANDU-6 reactors using RELAP-CANDU/SCAN coupled code system

    International Nuclear Information System (INIS)

    Recently, there is an effort to improve the accuracy and reality in the transient simulation of nuclear power plants. In the prediction of the system transient, the system code simulates the system transient using the power transient curve predicted from the reactor core physics code. However, the pre-calculated power curve could not adequately predict the behavior of power distribution during transient since the coolant density change has influence on the power shape due to the change of the void reactivity. Therefore, the consolidation between the reactor core physics code and the system thermal-hydraulic code takes into consideration to predict more accurate and realistic for the transient simulation. In this regard, there are two codes are developed to assess the safety of CANDU reactor. RELAP-CANDU is a thermal-hydraulic system code for CANDU reactors developed on the basis of RELAP5/MOD3 in such a way to modify inside model for simulating the thermal-hydraulic characteristics of horizontal type reactors. SCAN (SNU CANDU-PHWR Neutronics) is a three dimensional neutronics nodal code to simulate the core physics characteristics for CANDU reactors. To couple SCAN code with RELAP-CANDU code, SCAN code was improved as a spatial kinetics calculation module in such a way to generate a SCAN DLL (dynamic linked library version of SCAN). The coupled code system, RELAP-CANDU/SCAN, enables real-time feedback calculations between thermal-hydraulic variables of RELAP-CANDU and reactor powers of SCAN. To verify the reliability of RELAP-CANDU/SCAN coupled code system, an assessment of 40% reactor inlet header (RIH) break loss of coolant accident (LOCA) with loss of Class IV power (LOP) for Wolsong Unit 2 conducted using RELAP/CANDU-SCAN coupled system. The LOCA with LOP is one of GAI (Generic Action Items) for CANDU reactors issued by CNSC (Canadian Nuclear Safety Commission) and IAEA (International Atomic Energy Agency)

  10. LONGER: a computer program for longitudinal ridging and axial collapse assessment of CANDU fuel

    International Nuclear Information System (INIS)

    CANDU® fuel element sheath is designed to be thin and flexible for the benefit of enhanced heat transfer from the pellet to the coolant through the sheath. The flexibility of the sheath may allow the formation of longitudinal ridges on the sheath or collapse of the sheath into an axial gap under certain conditions. For both cases of deformations, the sheath may experience significant strains, and may result in sheath failure. To ensure the sheath mechanical integrity, the fuel element design needs to be assessed to preclude the conditions for longitudinal ridging and sheath collapse into the axial gap. The AECL developed LONGER computer program is used in fuel design analysis for such purpose. The LONGER code contains a number of models derived based on measurements (empirical models) and based on analytical equations, to predict the following parameters related to the deformations of CANDU nuclear fuel element sheaths. For longitudinal ridging: The critical diametral clearance for sheath longitudinal ridging, and The critical pressure for longitudinal ridging of the sheath. For axial collapse: The critical pressure for instantaneous sheath collapse into an axial gap. For circumferential collapse: The critical pressure for elastic collapse of the sheath, and The effective circumferential collapse pressure of the sheath by taking into account the axial and radial loads and the ovality of the sheath. The LONGER code has been qualified in accordance with the CSA standard N286.7-99 compliant AECL Software Quality Assurance (SQA) program. This paper describes the features and capabilities of the LONGER code that are used in CANDU fuel design analysis. (author)

  11. LONGER: a computer program for longitudinal ridging and axial collapse assessment of CANDU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Paul, U.K.; Xu, Z.; Xu, S.; Wang, X.; Chakraborty, K. [Atomic Energy of Canada Limited, Mississauga (Canada)

    2010-07-01

    CANDU® fuel element sheath is designed to be thin and flexible for the benefit of enhanced heat transfer from the pellet to the coolant through the sheath. The flexibility of the sheath may allow the formation of longitudinal ridges on the sheath or collapse of the sheath into an axial gap under certain conditions. For both cases of deformations, the sheath may experience significant strains, and may result in sheath failure. To ensure the sheath mechanical integrity, the fuel element design needs to be assessed to preclude the conditions for longitudinal ridging and sheath collapse into the axial gap. The AECL developed LONGER computer program is used in fuel design analysis for such purpose. The LONGER code contains a number of models derived based on measurements (empirical models) and based on analytical equations, to predict the following parameters related to the deformations of CANDU nuclear fuel element sheaths. For longitudinal ridging: The critical diametral clearance for sheath longitudinal ridging, and The critical pressure for longitudinal ridging of the sheath. For axial collapse: The critical pressure for instantaneous sheath collapse into an axial gap. For circumferential collapse: The critical pressure for elastic collapse of the sheath, and The effective circumferential collapse pressure of the sheath by taking into account the axial and radial loads and the ovality of the sheath. The LONGER code has been qualified in accordance with the CSA standard N286.7-99 compliant AECL Software Quality Assurance (SQA) program. This paper describes the features and capabilities of the LONGER code that are used in CANDU fuel design analysis. (author)

  12. Experimental investigation of syngas flame stability using a multi-tube fuel injector in a high pressure combustor

    Science.gov (United States)

    Maldonado, Sergio Elzar

    Over 92% of the coal consumed by power plants is used to generate electricity in the United States (U.S.). The U.S. has the world's largest recoverable reserves of coal, it is estimated that reserves of coal will last more than 200 years based in current production and demand levels. Integrated Gasification Combined Cycle (IGCC) power plants aim to reduce the amount of pollutants by gasifying coal and producing synthesis gas. Synthesis gas, also known as syngas, is a product of coal gasification and can be used in gas turbines for energy production. Syngas is primarily a mixture of hydrogen and carbon monoxide and is produced by gasifying a solid fuel feedstock such as coal or biomass. The objective of the thesis is to create a flame stability map by performing various experiments using high-content hydrogen fuels with varying compositions of hydrogen representing different coal feedstocks. The experiments shown in this thesis were performed using the High-Pressure Combustion facility in the Center for Space Exploration Technology Research (CSETR) at the University of Texas at El Paso (UTEP). The combustor was fitted with a novel Multi-Tube fuel Injector (MTI) designed to improve flame stability. This thesis presents the results of testing of syngas fuels with compositions of 20, 30, and 40% hydrogen concentrations in mixtures with carbon monoxide. Tests were completed for lean conditions ranging from equivalence ratios between 0.6 and 0.9. The experimental results showed that at an equivalence ratio of 0.6, a stable flame was not achieved for any of the fuel mixtures tested. It was also observed that the stability region of the syngas flame increased as equivalence ratio and the hydrogen concentration in syngas fuel increases with the 40% hydrogen-carbon monoxide mixture demonstrating the greatest stability region. Design improvements to the MTI are also discussed as part of the future work on this topic.

  13. A Numerical Model Prediction for Boiling Multi Channel Flow Rate Distribution and Application in 600MW Supercritical Variable-Pressure Once-Through Boiler with Vertical Tube Coils

    Institute of Scientific and Technical Information of China (English)

    1996-01-01

    Flow rate distribution is important in a multi channel system when the flow is heated non-uniformly,This paper describes a steady state approach for obtaining the flow distribution among various tubes of complex multi channel system,Based on the Present approach,a program has been developed which is directly applied in thermal hydraulic design and investigation of 600MW supercritical variable-pressure once through boiler.

  14. 管内高压智能封堵机器人%The In-tube Pressurized Intelligent Plugging Robot

    Institute of Scientific and Technical Information of China (English)

    刘华洁; 张策; 张仕民; 朱吉祥

    2013-01-01

    为满足国内管道快速维修的需要,开展了管道智能封堵技术研究.在介绍管内智能封堵机器人的封堵作业流程后,描述了封堵机器人的结构组成,包括双向清管式封堵单元、远程控制系统和地面控制中心,给出了主要技术参数.随后简要介绍和分析了封堵机器人的性能试验情况,包括通过性能试验、双向通信和压力试验以及解封试验.试验证明封堵机器人可在一段管道内实现多次封堵和解堵作业,大大缩短管道停输时间,且操作简单,封堵性能良好,能够封堵20MPa的高压,无渗漏.该智能封堵机器人的研制成功为国内管道维抢修技术提供了补充.%To meet the domestic needs of fast maintenance ot pipeline,research on the plugging technology ofpipeline was conducted.The paper first introduces the plugging process of in-tube intelligent plugging robot,describes the structural composition of the robot,including two-way pigging plugging unit,remote control system and ground control center,and offers the main technological parameters.Then,it briefly introduces and analyzes the performance test of the robot,including passage capacity test,two-way communication,pressure test and plug re-moval test.The tests have proved that the robot can achieve multiple plugging and plug removal operations in a sec-tion of pipeline.This remarkably shortens the pipeline shutdown time.The operation is simple and the plugging per-formance is desirable.The robot can plug as high as 20 MPa pressure with no leakage.The successful developmentof the robot serves as a supplementation for domestic pipeline maintenance technology.

  15. Endotracheal tube resistance and inertance in a model of mechanical ventilation of newborns and small infants—the impact of ventilator settings on tracheal pressure swings

    International Nuclear Information System (INIS)

    Resistive properties of endotracheal tubes (ETTs) are particularly relevant in newborns and small infants who are generally ventilated through ETTs with a small inner diameter. The ventilation rate is also high and the inspiratory time (ti) is short. These conditions effectuate high airway flows with excessive flow acceleration, so airway resistance and inertance play an important role. We carried out a model study to investigate the impact of varying ETT size, lung compliance and ventilator settings, such as peak inspiratory pressure (PIP), positive end expiratory pressure (PEEP) and inspiratory time (ti) on the pressure–flow characteristics with respect to the resistive and inertive properties of the ETT. Pressure at the Y piece was compared to direct measurement of intratracheal pressure (Ptrach) at the tip of the ETT, and pressure drop (ΔPETT) was calculated. Applying published tube coefficients (Rohrer's constants and inertance), Ptrach was calculated from ventilator readings and compared to measured Ptrach using the root-mean-square error. The most relevant for ΔPETT was the ETT size, followed by (in descending order) PIP, compliance, ti and PEEP, with gas flow velocity being the principle in common for all these parameters. Depending on the ventilator settings ΔPETT exceeded 8 mbar in the smallest 2.0 mm ETT. Consideration of inertance as an additional effect in this setting yielded a better agreement of calculated versus measured Ptrach than Rohrer's constants alone. We speculate that exact tracheal pressure tracings calculated from ventilator readings by applying Rohrer's equation and the inertance determination to small size ETTs would be helpful. As an integral part of ventilator software this would (1) allow an estimate of work of breathing and implementation of an automatic tube compensation, and (2) be important for gentle ventilation in respiratory care, especially of small infants, since it enables the physician to estimate

  16. Impact of suctioning on tracheal tube cuff pressure%吸痰对人工气道套囊内压力的影响

    Institute of Scientific and Technical Information of China (English)

    朱艳萍; 刘亚芳; 任璐璐; 潘红; 尹亚丽; 何静

    2011-01-01

    目的 通过观察吸痰时及吸疾后30min人工气道套囊压力的变化,探讨吸痰对人工气道套囊压力的影响.方法 吸痰前用测压表调整人工气道套囊压力为30CMH2O,持续监测吸痰时的套囊最高压力、患者有无咳嗽,以及吸痰后5min、10min、15min、30min的压力,并记录吸痰后套囊压力降至25cmH2O时所需时间.结果 在吸痰过程中81.25%(78/96)患者发生咳嗽.本组套囊内压力平均明显升高至(89.42±31.37)cmH20,咳嗽者套囊内平均压力为(96.00±25.99)cmH2O,高于无咳嗽者套囊内平均压力(60.89±37.14)cmH2O,差异有统计学意义(P<0.01).吸痰时套囊内压力升高者较保持者更易下降至正常低限(25cmH2O),差异有统计学意义(P<0.01).结论 患者吸痰过程中容易发生咳嗽,人工气道套囊内压力波动明显,建议临床上在吸痰后30min内调整套囊内压力,必要时应立即调整,避免囊内压力过低或过高对患者的伤害.%Objective To investigate the influence of suctioning on tracheal tube cuff pressure. Methods The tracheal tube cuff pressure was set to 30 cmH2O with pressure gauge before suctioning. Then,the maximum cuff pressure and patients'cough during suctioning and the cuff pressures at 5 min,10 min,15 min,30 min after sunctioning were recorded. Moreover,the duration until the cuff pressure dropped to 25 cmH2O was recorded. Results The incidence of cough during suctioning was 81.25 percent. The tracheal tube cuff pressure increased by (89.42±31.37) cmH2O during suctioning,and the cuff pressure in patients with cough was significanlty higher than that of patients without cough(P<0.01). The cuff pressure dropped to 25 cmH2O faster in patients whose cuff pressure was increased during suctioning than those whose cuff pressure maintained stable (P<0.01). Conclusion Cough is prevalent during suctioning and the change of tracheal tube cuff pressure is obvious. It is suggested to regulate the tracheal tube cuff pressure

  17. Experimental sizing and assessment of two-phase pressure drop correlations for a capillary tube with transcritical and subcritical carbon dioxide flow

    International Nuclear Information System (INIS)

    In the last years, CO2 was proposed as an alternative refrigerant for different refrigeration applications (automotive air conditioning, heat pumps, refrigerant plants, etc.) In the case of low power refrigeration applications, as a household refrigerator, the use of too expensive components is not economically sustainable; therefore, even if the use of CO2 as the refrigerant is desired, it is preferable to use conventional components as much as possible. For these reasons, the capillary tube is frequently proposed as expansion system. Then, it is necessary to characterize the capillary in terms of knowledge of the evolving mass flow rate and the associate pressure drop under all possible operative conditions. For this aim, an experimental campaign has been carried out on the ENEA test loop 'CADORE' to measure the performance of three capillary tubes having same inner diameter (0.55 mm) but different lengths (4, 6 and 8 meters). The test range of inlet pressure is between about 60 and 110 bar, whereas external temperatures are between about 20 to 42 °C. The two-phase pressure drop through the capillary tube is detected and experimental values are compared with the predictions obtained with the more widely used correlations available in the literature. Correlations have been tested over a wide range of variation of inlet flow conditions, as a function of different inlet parameters.

  18. Experimental sizing and assessment of two-phase pressure drop correlations for a capillary tube with transcritical and subcritical carbon dioxide flow

    Science.gov (United States)

    Trinchieri, R.; Boccardi, G.; Calabrese, N.; Celata, G. P.; Zummo, G.

    2014-04-01

    In the last years, CO2 was proposed as an alternative refrigerant for different refrigeration applications (automotive air conditioning, heat pumps, refrigerant plants, etc.) In the case of low power refrigeration applications, as a household refrigerator, the use of too expensive components is not economically sustainable; therefore, even if the use of CO2 as the refrigerant is desired, it is preferable to use conventional components as much as possible. For these reasons, the capillary tube is frequently proposed as expansion system. Then, it is necessary to characterize the capillary in terms of knowledge of the evolving mass flow rate and the associate pressure drop under all possible operative conditions. For this aim, an experimental campaign has been carried out on the ENEA test loop "CADORE" to measure the performance of three capillary tubes having same inner diameter (0.55 mm) but different lengths (4, 6 and 8 meters). The test range of inlet pressure is between about 60 and 110 bar, whereas external temperatures are between about 20 to 42 °C. The two-phase pressure drop through the capillary tube is detected and experimental values are compared with the predictions obtained with the more widely used correlations available in the literature. Correlations have been tested over a wide range of variation of inlet flow conditions, as a function of different inlet parameters.

  19. Analysis of the effect of tube arrangement and inclination on pressure drop in an intermediate heat exchanger of liquid metal reactor

    Energy Technology Data Exchange (ETDEWEB)

    ChoiI, Seok Ki; Choi, Il Kon; Nam, Ho Yun; Choi, Jong Hyeun [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2002-05-01

    An experimental study on the effect of tube arrangement and inclination on the pressure drop in the intermediate heat exchanger is performed. Measurements are made for pressure drop in the triangular and rotated triangular tue arrays whose inclined angles are 30, 45, 60, 75 and 90 degrees. The pitch to tube diameter ratio is 1.6 and the range of Reynolds number based on the free stream velocity and tube diameter is 870-64,000. The experimental results show that the magnitude of dimensionless pressure drop increases with the inclined angle and decreases significantly when the inclined angle is less than 45 degree. The previous correlations are evaluated using the experimental data. The ESDU correlation agrees well with the present data for the triangular arrays. But some discrepancies are observed for the rotated triangular arrays when the inclined angles are 45 and 30 degrees. The Idel'chik correlation generally agrees well with the measured data for the rotated triangular arrays except for inclined angle of 30 degree. The Idel'chik correlation needs modification for the triangular arrays. The modified Idel'chik correlation agrees well with the measure data within 10%. 32 refs., 59 figs., 11 tabs. (Author)

  20. High Pressure Pneumatic Forming of Ti-3Al-2.5V Titanium Tubes in a Square Cross-Sectional Die

    Directory of Open Access Journals (Sweden)

    Gang Liu

    2014-08-01

    Full Text Available A new high strain rate forming process for titanium alloys is presented and named High Pressure Pneumatic Forming (HPPF, which might be applicable to form certain tubular components with irregular cross sections with high efficiency, both with respect to energy cost and time consumption. HPPF experiments were performed on Ti-3Al-2.5V titanium alloy tubes using a square cross-sectional die with a small corner radius. The effects of forming of pressure and temperature on the corner filling were investigated and the thickness distributions after the HPPF processes at various pressure levels are discussed. At the same time, the stress state, strain and strain rate distribution during the HPPF process were numerically analyzed by the finite element method. Microstructure evolution of the formed tubes was also analyzed by using electron back scattering diffraction (EBSD. Because of different stress states, the strain and strain rate are very different at different areas of the tube during the corner filling process, and consequently the microstructure of the formed component is affected to some degree. The results verified that HPPF is a potential technology to form titanium tubular components with complicated geometrical features with high efficiency.

  1. Preliminary evaluation of licensing issues associated with U. S. -sited CANDU-PHW nuclear power plants

    Energy Technology Data Exchange (ETDEWEB)

    van Erp, J B

    1977-12-01

    The principal safety-related characteristics of current CANDU-PHW power plants are described, and a distinction between those characteristics which are intrinsic to the CANDU-PHW system and those that are not is presented. An outline is given of the main features of the Canadian safety and licensing approach. Differences between the U.S. and Canadian approach to safety and licensing are discussed. Some of the main results of the safety analyses, routinely performed for CANDU-PHW reactors, are presented. U.S.-NRC General Design Criteria are evaluated as regards their applicability to CANDU-PHW reactors; vice-versa the CANDU-PHW reactor is evaluated with respect to its conformance to the U.S.-NRC General Design Criteria. A number of design modifications are proposed to be incorporated into the CANDU-PHW reactor in order to facilitate its introduction into the U.S.

  2. Improved operability of the CANDU 9 control centre

    Energy Technology Data Exchange (ETDEWEB)

    Macbeth, M. J.; Webster, A. [Atomic Energy of Canada Limited, Saskatoon (Canada)

    1996-04-15

    The next generation CANDU nuclear power plant being designed by AECL is the 900 MWe class CANDU 9 station. It is based upon the Darlington CANDU station design which is among the world leaders in capacity factor with low Operation, Maintenance and Administration (OM and A) costs. This Control Centre design includes the proven functionality of existing CANDU control centres (including the Wolsong 2,3, and 4 control centre improvements, such as the Emergency Core Cooling panels), the characteristics identified by systematic design with human factors analysis of operations requirements and the advanced features needed to improve station operability which is made possible by the application of new technology. The CANDU 9 Control Centre provides plant staff with an improved operability capability due to the combination of proveness, systematic design with human factors engineering and enhanced operating features. Significant features which contribute to this improved operability include: {center_dot} Standard NSP, BOP and F/H panels with controls and indicators integrated by a standard display/presentation philosophy. {center_dot} Common plant parameter signal database for extensive monitoring, checking, display and annunciation. {center_dot} Powerful annunciation system allowing alarm filtering, prioritizing and interrogation to enhance staff recognition of events, plant state and required corrective procedural actions. {center_dot} The use of an overview display to present immediate and uncomplicated plant status information to facilitate operator awareness of unit status in a highly readable and recognizable format. {center_dot} Extensive cross checking of similar process parameters amongst themselves, with the counterpart safety system parameters and as well as with 'signature' values obtained from known steady state conditions. {center_dot} Powerful calculation capabilities, using the plant wide database, providing immediate recognizable and readable and

  3. The Feasibility of Multidimensional CFD Applied to Calandria System in the Moderator of CANDU-6 PHWR Using Commercial and Open-Source Codes

    OpenAIRE

    Kim, Hyoung Tae; Chang, Se-Myong; Shin, Jong-Hyeon; Kim, Yong Gwon

    2016-01-01

    The moderator system of CANDU, a prototype of PHWR (pressurized heavy-water reactor), has been modeled in multidimension for the computation based on CFD (computational fluid dynamics) technique. Three CFD codes are tested in modeled hydrothermal systems of heavy-water reactors. Commercial codes, COMSOL Multiphysics and ANSYS-CFX with OpenFOAM, an open-source code, are introduced for the various simplified and practical problems. All the implemented computational codes are tested for a benchm...

  4. Loss of efficiency of polymeric drag reducers induced by high Reynolds number flows in tubes with imposed pressure

    Science.gov (United States)

    Soares, Edson J.; Sandoval, Gustavo A. B.; Silveira, Lucas; Pereira, Anselmo S.; Trevelin, Renata; Thomaz, Fabricio

    2015-12-01

    This paper studies the loss of efficiency of polymeric drag reducers induced by high Reynolds number flows in tubes. The overall pressure was fixed and the apparatus was built so as to minimize the polymer degradation. We used three kinds of polymers: two flexible and one rigid. We conducted our tests to take into account the drag reduction (DR) for a wide range of concentrations of each polymer. The main results are displayed for the DR as a function of the number of passes through the apparatus. The mechanism of the loss of efficiency for the Xanthan Gum (XG) solutions (the rigid one) seems to be completely different from that observed for Poly (ethylene oxide) (PEO) and Polyacrylamide (PAM) (the flexible materials). While the PEO and PAM mechanically degrade by the action of the turbulent flow, the XG seems to remain intact, even after many passes through the pipe flow apparatus. From the practical point of view, it is worth noting that the PAM solutions are clearly more efficient than the PEO and XG. Another practical point that deserves attention is concerned with the asymptotic drag reduction found for XG. Although its maximum DR was significantly smaller than that found for PEO, the final value for both polymers were quite the same, which is obviously related to the intensified mechanical molecule scission in the PEO solutions. Our results for the relative drag reduction (the current value of DR divided by its maximum obtained at the first pass) was quite well fitted by the decay function proposed in our previous paper [A. S. Pereira and E. J. Soares, "Polymer degradation of dilute solutions in turbulent drag reducing flows in a cylindrical double gap rheometer device," J. Non-Newtonian Fluid Mech. 179, 9-22 (2012)], in which a rotating apparatus was used. This strongly suggests that the physical mechanism that governs the degradation phenomenon is independent of the geometry. We also used a degradation model for PEO proposed by Vonlanthen and Monkewitz

  5. Comparison Study on Thermal-Hydraulic Analysis Depending on Liquid Relief Valve Response for an Station Blackout in CANDU-6

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. M.; Kho, D. W. [KHNP-CRI, Daejeon (Korea, Republic of); Choi, S. H.; Moon, B. J.; Kim, S. R. [Nuclear Engineering Service and Solution Co., Daejeon (Korea, Republic of)

    2014-10-15

    The purpose of this analysis is to compare the results of thermal-hydraulic analysis depending on liquid relief valve response during a station black out (SBO) events in CANDU-6. The primary heat transport system (PHTS) behavior following the postulated SBO is analyzed using CATHENA code. In the paper, analysis was performed to evaluate the effect on coolant system where LRVs are assumed to be opened or opened according to normal open characteristics in the condition of SBO. The result showed that the primary pressure boundary is extended from LRV to DCT and the effects on primary system behavior were neglectable.

  6. Design principles for CANDU control centres in response to evolving utility business needs

    International Nuclear Information System (INIS)

    Nuclear generation operators are facing a challenging business environment at the beginning of the new millennium. Evolving changes in business context, competitive commercial pressures, and changes in technology have dictated recurring evaluation of operational practices and the adequacy of supporting tools, and the pursuit of opportunities for operational improvement. A key area of utility operations that has been impacted by these changes is the nuclear plant control centre. Changes to workspace layout, equipment provisions, staffing, and work organization are examples of some of the adjustments being introduced to improve operational and safety effectiveness. This paper discusses some of the key factors influencing these changes and identifies additional design principles for CANDU control centres that will enable new control centre designs and retrofits of existing control centres to remain relevant and responsive to utility needs. (author)

  7. Thermal stability of chloroform in the steam condensate cycle of CANDU-PHW nuclear power plant

    International Nuclear Information System (INIS)

    Analysis of samples taken at the Gentilly 2 (Quebec) CANDU-PHW (CANadian Deuterium Uranium - Pressurized Heavy Water) plant after chlorination and demineralization revealed the presence of all four trihalomethanes (THMs) (CHCl3, CHBrCl2, CHBr2Cl and CHBr3) and other unidentified halogenated volatile compounds. Among the THMs, chloroform was the major contaminant. A study of its thermal stability in water at different temperatures confirmed the degradation of the CHCl3 molecule according to the equation CHCl3 + H2O → CO + 3 HCl. The reaction follows first order kinetics and has an activation energy of 100 kJ/mol. The estimated half-life is six seconds at 260 deg C, the maximum temperature of the steam condensate cycle

  8. Used CANDU fuel waste consumed and eliminated: environmentally responsible, economically sound, energetically enormous

    International Nuclear Information System (INIS)

    The 43,800 tonnes of currently stored CANDU nuclear fuel waste can all be consumed in fast-neutron reactors (FNRs) to reduce its long-term radioactive burden 100,000 times while extracting about 130 times more nuclear energy than the prodigious amounts that have already been gained from the fuel in CANDU reactors. The cost of processing CANDU fuel for use in FNRs plus the cost of recycling the FNR fuel is about 2.5 times less on a per kWh energy basis than the currently projected cost of disposal of 3.6 million used CANDU fuel bundles in a deep geological repository. (author)

  9. A study to develop the domestic functional requirements of the specific safety systems of CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Man Woong; Lee, Jae Young; Bang, Kwang Hyun [Handong Global Univ., Pohang (Korea, Republic of)] (and others)

    2001-03-15

    The present research has been made to develop and review critically the functional requirements of the specific safety systems of CANDU such as SOS-1, SOS-2, ECCS and containment. Based on R documents for this, a systematic study was made to develop the domestic regulation statements. Also, the conventional laws are carefully reviewed to see the compatibility to CANDU. Also, the safety assessment method for CANDU was studied by reviewing C documents and recommendation of IAEA. Through the present works, the vague policy in the CANDU safety regulation is cleaning up in a systematic form and a new frame to measure the objective risk of nuclear power plants was developed.

  10. A study to develop the domestic functional requirements of the specific safety systems of CANDU

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Man Woong [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Lee, Jae Young; Park, Kun Chul [Handong Global Univ., Pohang (Korea, Republic of)] (and others)

    2003-03-15

    The present research has been made to develop and review critically the functional requirements of the specific safety systems of CANDU such as SDS-1, SDS2, ECCS, and containment. Based on R documents for this, a systematic study was made to develop the domestic regulation statements. Also, the conventional laws are carefully reviewed to see the compatibility to CANDU. Also, the safety assessment method for CANDU was studied by reviewing C documents and recommendation of IAEA. Through the present works, the vague policy in the CANDU safety regulation is cleaning up in a systematic form and a new frame to measure the objective risk of nuclear power plants was developed.

  11. Ultrasonic water level determination of the high-pressure boilers tubes; Determinacao do nivel d'agua em tubos verticais de caldeiras aquatubulares por ultra-som

    Energy Technology Data Exchange (ETDEWEB)

    Goettems, Felipe Samuel; Reolon, Amon Marques; Avancini, Flavio; Braga, Rubem Manoel de; Reguly, Afonso [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Lab. de Metalurgia Fisica], e-mail: fgoettems@demet.ufrgs.br

    2006-07-01

    Electric power is very important to our society and thermoelectric power plant. They are especially important mainly in the summer when there is a scarcity in water supply to hydroelectric power plants. Southern Brazilian thermoelectric power plants employ high-pressure boilers in order to generate water vapor which, in turn, moves turbines to produce electricity. These high-pressure boilers must work in a continuous way to avoid damages caused by emergency halts. To accomplish this, some actions must be taken. The water height inside of the tubes must be kept in a strict level to avoid thermal gradient in both water walls and super-heater header. In this water walls the water become in vapor. The best way to regulate the valves that command the water level is through the control of the water height and this is the main purpose of this work. The ultrasound is a nondestructive test which is able in doing this control without damaging the tube. This method allows determining the water level, improving the system performance and reducing the maintenance costs caused by tube collapse. (author)

  12. Development of the advanced CANDU technology -Development of CANDU advanced fuel fabrication technology-

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chang Bum; Park, Choon Hoh; Park, Chul Joo; Kwon, Woo Joo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-07-01

    This project is carrying out jointly with AECL to develop CANFLEX fuel which can enhance reactor safety, fuel economy and can be used with various fuel cycles (natural U, slightly enriched U, other advanced fuel). The final goal of this research is to load the CANFLEX fuel in commercial CANDU reactor for demonstration irradiation. The annual portion of research activities performed during this year are followings ; The detail design of CANFLEX-NU fuel was determined. Based on this design, various fabrication drawings and process specifications were revised. The seventeen CANFLEX-NU fuel bundles for reactivity test in ZED-2 and out-pile test, two CANFLEX-SEU fuel bundles for demo-irradiation in NRU were fabricated. Advanced tack welding machine was designed and sequence control software of automatic assembly welder was developed. The basic researches related to fabrication processes, such as weld evaluation by ECT, effect of additives in UO{sub 2}, thermal stabilities of Zr based metallic glasses, were curried out. 51 figs, 22 tabs, 42 refs. (Author).

  13. THE INSIDE PRESSURE OF STENT TUBE ON CHOLEDOCO-JEJUNOSTOMY SCAR: A STUDY ON SCAR TISSUE COLLAGEN

    Institute of Scientific and Technical Information of China (English)

    郭善禹; 周林斌; 姚德成; 孙建民

    2002-01-01

    Objective As the beneficial effect to the skin scar under external bandage compression, intra-choledocal stent must have the same effect on splanchnic scar formation. The experiment consists to work out the time optimum to yield a minimum scar formation. Methods By means of transmitting electronic microscope (TEM), computer assisted three-dimensional morphometry (CAM), and biochemical analysis to determine the extracellular collagen volume density (ECVD) and biochemical collagen content (BCC), to analyze the ultrastructure and components within scar tissues removed from the specimens in 3 groups of experimental animals were detailed. Results In the animals of simple choledoco-jejunostomy (CJ) group, active scar proliferation was seen in all specimens excised within one year after operation. In the stent group, decreasing collagen fibers arranged in orientation began to appear in the 6-month specimens and scar maturation existed in the 9- and 12-month specimens. In periodic tube withdrawal group, 3 months following tube ablation, scar proliferation recurred in the 6th month tube retaining animals, whereas scar maturation without recurrence happened in animals following 9 to 12 months tube retaining. Conclusion 9~12 months of tube stent is necessary for stable scar maturation.

  14. Economic analysis of alternative options in CANDU fuel cycle

    International Nuclear Information System (INIS)

    In this study, fuel cycle options for CANDU reactor were studied. Three main options in a CANDU fuel cycle involve use of : (1) natural uranium (0.711 weight percent U-235) fuel, (2) slightly enriched uranium (1.2 weight percent U-235) fuel, and (3) recovered uranium (0.83 weight percent U-235) fuel from light water reactor spent fuel. ORIGEN-2 computer code was used to identify composition of the spent fuel for each option , including the standard LWR fuel (3.3 weight percent U-235). Uranium and plutonium credit calculations were performed by using ORIGEN-2 output. WIMSD-5 computer code was used to determine maximum discharge burnup values for each case. Cost estimations were carried out using specially-developed computer programs. Comparison of levelized costs for the fuel cycle options and sensitivity analysis for the cost components are also presented

  15. CANDU type fuel behavior evaluation - a probabilistic approach

    International Nuclear Information System (INIS)

    In order to realistically assess the behavior of the fuel elements during in-reactor operation, probabilistic methods have recently been introduced in the analysis of fuel performance. The present paper summarizes the achievements in this field at the Institute for Nuclear Research (INR), pointing out some advantages of the utilized method in the evaluation of CANDU type fuel behavior in steady state conditions. The Response Surface Method (RSM) has been selected for the investigation of the effects of the variability in fuel element computer code inputs on the code outputs (fuel element performance parameters). A new developed version of the probabilistic code APMESRA based on RSM is briefly presented. The examples of application include the analysis of the results of an in-reactor fuel element experiment and the investigation of the calculated performance parameter distribution for a new CANDU type extended burnup fuel element design. (author)

  16. Effect of hydrogen isotope content on tensile flow behavior of Zr-2.5Nb pressure tube material between 25 and 300 °C

    Science.gov (United States)

    Bind, A. K.; Sunil, S.; Singh, R. N.

    2016-08-01

    Tensile properties of autoclaved Zr-2.5Nb pressure tube material containing hydrogen isotope between 5 and 200 wppm were evaluated between 25 and 300 °C using specimens with its axis oriented along longitudinal direction of the tube. Analysis of tensile test results showed that both YS and UTS of this alloy decreased linearly with increasing test temperature. The uniform and total plastic strain decreased marginally with increase in test temperature. At all test temperatures, before necking tensile properties were unaffected by hydrogen isotope concentration whereas hydrogen isotope had clear effect on post-necking tensile properties especially at 25 and 100 °C. Post-necking ductility showed a transition behavior at 25 and 100 °C and it was able to capture the effect of hydride embrittlement in this material.

  17. Proceedings of the third international conference on CANDU maintenance

    International Nuclear Information System (INIS)

    The third international conference on Candu maintenance included sessions on the following topics: predictive maintenance, reliability improvements, steam generator monitoring, tools and instrumentation, valve performance, fuel channel inspection and maintenance, steam generator maintenance, environmental qualification, predictive maintenance, instrumentation and control, steam generator cleaning, decontamination and radiation protection, inspection techniques, maintenance program strategies and valve packing experience, remote tooling/ robotics and fuel handling. The individual papers have been abstracted separately

  18. Scenarios for the transmutation of actinides in CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hyland, Bronwyn, E-mail: hylandb@aecl.ca [Atomic Energy of Canada Limited, Chalk River Laboratories, Chalk River, Ontario, K0J 1J0 (Canada); Gihm, Brian, E-mail: gihmb@aecl.ca [Atomic Energy of Canada Limited, 2251 Speakman Drive, Mississauga, Ontario, L5K 1B2 (Canada)

    2011-12-15

    With world stockpiles of used nuclear fuel increasing, the need to address the long-term utilization of this resource is being studied. Many of the transuranic (TRU) actinides in nuclear spent fuel produce decay heat for long durations, resulting in significant nuclear waste management challenges. These actinides can be transmuted to shorter-lived isotopes to reduce the decay heat period or consumed as fuel in a CANDU(R) reactor. Many of the design features of the CANDU reactor make it uniquely adaptable to actinide transmutation. The small, simple fuel bundle simplifies the fabrication and handling of active fuels. Online refuelling allows precise management of core reactivity and separate insertion of the actinides and fuel bundles into the core. The high neutron economy of the CANDU reactor results in high TRU destruction to fissile-loading ratio. This paper provides a summary of actinide transmutation schemes that have been studied in CANDU reactors at AECL, including the works performed in the past. The schemes studied include homogeneous scenarios in which actinides are uniformly distributed in all fuel bundles in the reactor, as well as heterogeneous scenarios in which dedicated channels in the reactor are loaded with actinide targets and the rest of the reactor is loaded with fuel. The transmutation schemes that are presented reflect several different partitioning schemes. Separation of americium, often with curium, from the other actinides enables targeted destruction of americium, which is a main contributor to the decay heat 100-1000 years after discharge from the reactor. Another scheme is group-extracted transuranic elements, in which all of the transuranic elements, plutonium (Pu), neptunium (Np), americium (Am), and curium (Cm) are extracted together and then transmuted. This paper also addresses ways of utilizing the recycled uranium, another stream from the separation of spent nuclear fuel, in order to drive the transmutation of other actinides.

  19. Experimental investigation and correlation of two-phase frictional pressure drop of R410A-oil mixture flow boiling in a 5 mm microfin tube

    Energy Technology Data Exchange (ETDEWEB)

    Ding, Guoliang; Hu, Haitao; Huang, Xiangchao [Institute of Refrigeration and Cryogenics, Shanghai Jiaotong University, Shanghai 200240 (China); Deng, Bin [Institute of Heat Transfer Technology, Golden Dragon Precise Copper Tube Group Inc., Shanghai 200135 (China); Gao, Yifeng [International Copper Association, Shanghai Office, Shanghai 200020 (China)

    2009-01-15

    This study presents experimental two-phase frictional data for R410A-oil mixture flow boiling in an internal spiral grooved microfin tube with outside diameter of 5 mm. Experimental parameters include the evaporation temperature of 5 C, the mass flux from 200 to 400 kg m{sup -2} s{sup -1}, the heat flux from 7.46 to 14.92 kW m{sup -2}, the inlet vapor quality from 0.1 to 0.8, and nominal oil concentration from 0 to 5%. The test results show that the frictional pressure drop of R410A initially increases with vapor quality and then decreases, presenting a local maximum in the vapor quality range between 0.7 and 0.8; the frictional pressure drop of R410A-oil mixture increases with the mass flux, the presence of oil enhances two-phase frictional pressure drop, and the effect of oil on frictional pressure drop is more evident at higher vapor qualities where the local oil concentrations are higher. The enhanced factor is always larger than unity and increases with nominal oil concentration at a given vapor quality. The range of the enhanced factor is about 1.0-2.2 at present test conditions. A new correlation to predict the local frictional pressure drop of R410A-oil mixture flow boiling inside the internal spiral grooved microfin tube is developed based on local properties of refrigerant-oil mixture, and the measured local frictional pressure drop is well correlated with the empirical equation proposed by the authors. (author)

  20. Safety benefits from CANDU reactor replacement. A case study

    International Nuclear Information System (INIS)

    Both total core replacement and core retubing have been used in the CANDU industry. For future plant refurbishments, based on experience both in new construction and in recent refurbishments, the concept of total core replacement has been revisited. This builds on practices for replacement of other large plant equipment like boilers. The Bruce CANDU reactors, with their local shield tanks built around the Calandria and containment closely located around that Calandria Shield Tank Assembly (CSTA), are believed to be good candidates for core replacement. A structured process was used to design a replacement CSTA suitable for Bruce A use. The work started with a study of opportunities for safety enhancements in the core. This progressed into design studies and related design assist safety analysis on the reactor. A key element of the work involved consideration of how verified features from later CANDU designs, and from our new reactor design work, could be tailored to fit this replacement core. The replacement reactor core brings in structural improvements in both calandria and end shield, and safety improvements like the natural circulation enhancing moderator cooling layout and further optimized reactivity layouts to improve shutdown system performance. Bruce Power are currently studying the business implications of this and retube techniques as part of preparation for future refurbishments. The work explained in this paper is in the context of the safety related changes and the work to choose and quantify them. (author)