Air enters closed systems not only via plastic tubes

Energy Technology Data Exchange (ETDEWEB)

1980-03-01

Some trouble within the branch was caused by passages of the SHT-interview with Hans Viessmann on the question of guarantee in case of corrosion in heating systems with plastic tubes. The plastic tubes are made responsible for bringing air into the heating system by diffusion, thus causing a danger of corrosion. In this connection, the Thermoval Systemheatings Company, Cologne, makes a statement.

Pressure tube type research reactor

International Nuclear Information System (INIS)

Ueda, Hiroshi.

1976-01-01

Object: To prevent excessive heat generation due to radiation of a pressure tube vessel. Structure: A pressure tube encasing therein a core comprises a dual construction comprising inner and outer tubes coaxially disposed. High speed cooling water is passed through the inner tube for cooling. In addition, in the outer periphery of said outer tube there is provided a forced cooling tube disposed coaxially thereto, into which cooling fluid, for example, such as moderator or reflector is forcibly passed. This forced cooling tube has its outer periphery surrounded by the vessel into which moderator or reflector is fed. By the provision of the dual construction of the pressure tube and the forced cooling tube, the vessel may be prevented from heat generation. (Ikeda, J.)

ACCEPT: a three-dimensional finite element program for large deformation elastic-plastic-creep analysis of pressurized tubes (LWBR/AWBA Development Program)

International Nuclear Information System (INIS)

Hutula, D.N.; Wiancko, B.E.

1980-03-01

ACCEPT is a three-dimensional finite element computer program for analysis of large-deformation elastic-plastic-creep response of Zircaloy tubes subjected to temperature, surface pressures, and axial force. A twenty-mode, tri-quadratic, isoparametric element is used along with a Zircaloy materials model. A linear time-incremental procedure with residual force correction is used to solve for the time-dependent response. The program features an algorithm which automatically chooses the time step sizes to control the accuracy and numerical stability of the solution. A contact-separation capability allows modeling of interaction of reactor fuel rod cladding with fuel pellets or external supports

High temperature deformation behavior of gradually pressurized zircaloy-4 tubes

International Nuclear Information System (INIS)

Suzuki, Motoye

1982-03-01

In order to obtain preliminary perspectives on fuel cladding deformation behavior under changing temperature and pressure conditions in a hypothetical loss-of-coolant accident of PWR, a Zircaloy-4 tube burst test was conducted in both air and 99.97% Ar atomospheres. The tubes were directly heated by AC-current and maintained at various temperatures, and pressurized gradually until rupture occurred. Rupture circumferential strains were generally larger in Ar gas than in air and attained a maximum around 1100 K in both atmospheres. Some tube tested in air produced axially-extended long balloons, which proved not to be explained by such properties or ideas as effect of cooling on strain rate, superplasticity, geometrical plastic instability and stresses generated by surface oxide layer. A cause of the long balloon may be obtained in the anisotropy of the material structure. But even a qualitative analysis based on this property can not be made due to insufficient data of the anisotropy. (author)

Method for shaping polyethylene tubing

Science.gov (United States)

Kramer, R. C.

1981-01-01

Method forms polyethylene plastic tubing into configurations previously only possible with metal tubing. By using polyethylene in place of copper or stain less steel tubing inlow pressure systems, fabrication costs are significantly reduced. Polyethylene tubing can be used whenever low pressure tubing is needed in oil operations, aircraft and space applications, powerplants, and testing laboratories.

Analysis of the ballooning deformation of an internally pressurized thin-wall tube during fast thermal transients

International Nuclear Information System (INIS)

Lin, E.I.H.

1977-01-01

A large-strain time-dependent thermoplastic analysis has been developed for the ballooning deformation of a thin-wall tube subjected to internal pressure, axial loading, and fast thermal transients. This deformation initiates with the onset of plastic instability in the material, the onset being determined by a plastic-instability criterion for strain-rate sensitive materials. The interaction among the local ballooning geometry, the state of stress, and the plastic flow process was considered, and integration of the flow equations yields the local curvature and the states of stress and strain in the vicinity of the maximum ballooning site. The effects of axial constraint and heating rate were also discussed. The analysis was applied to a LWR Zircaloy cladding subjected to a constant heating rate and a range of internal pressures. The results agree very well with experimental strain-time data obtained from tube-burst tests. In most cases, the time of rupture was accurately predicted despite the lack of complete material-property data

Method of repairing pressure tube type reactors

International Nuclear Information System (INIS)

Asada, Takashi.

1983-01-01

Purpose: To enable to re-start the reactor operation in a short time, upon occurrence of failures in a pressure tube, as well as directly examine the cause for the failures in the pressure tube. Method: The pressure tube reactor main body comprises a calandria tank of a briquette form, pressure tubes, fuel assemblies and an iron-water shielding body. If failure is resulted to a pressure tube, the reactor operation is at first shutdown and nuclear fuel assemblies are extracted to withdraw from the pressure tube. Then, to an inlet pipe way and an outlet pipeway connected to the failed pressure tube, are attached plugs by means of welding or the like at the appropriate position where the radiation exposure dose is lower and the repairing work can be performed with ease. The pressure tube is disconnected to withdraw from the inlet pipeway and the outlet pipeway and, instead, radiation shielding plug tube is inserted and shield cooling device is actuated if required, wherein the reactor is actuated to re-start the operation. (Yoshino, Y.)

Contrastive Analysis and Research on Negative Pressure Beam Tube System and Positive Pressure Beam Tube System for Mine Use

Science.gov (United States)

Wang, Xinyi; Shen, Jialong; Liu, Xinbo

2018-01-01

Against the technical defects of universally applicable beam tube monitoring system at present, such as air suction in the beam tube, line clogging, long sampling time, etc., the paper analyzes the current situation of the spontaneous combustion fire disaster forecast of mine in our country and these defects one by one. On this basis, the paper proposes a research thought that improving the positive pressure beam tube so as to substitute the negative pressure beam tube. Then, the paper introduces the beam tube monitoring system based on positive pressure technology through theoretical analysis and experiment. In the comparison with negative pressure beam tube, the paper concludes the advantage of the new system and draws the conclusion that the positive pressure beam tube is superior to the negative pressure beam tube system both in test result and test time. At last, the paper proposes prospect of the beam tube monitoring system based on positive pressure technology.

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

International Nuclear Information System (INIS)

Krishnan, Suresh; Bhasin, Vivek; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.

1996-01-01

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

Annular gap measurement between pressure tube and calandria tube by eddy current technique

International Nuclear Information System (INIS)

Bhole, V.M.; Rastogi, P.K.; Kulkarni, P.G.

1992-01-01

In pressurised heavy water reactor (PHWR) major distinguishing feature is that there are number of identical fuel channels in the reactor core. Each channel consists of pressure tube of Zr-2.5 Nb or zircaloy-2 through which high temperature, high pressure primary coolant is passing. The pressure tube contains fuel. Surrounding the pressure tube there is low pressure, cool heavy water (moderator). The moderator is thermally separated from coolant by the tube which is nominally concentric with pressure tube called calandria tube. There are four garter springs in the annular gap between pressure tube and calandria tube. During the life of the reactor there are number of factors by which the pressure tube sags, most important factors are irradiation creep, thermal creep, fuel load etc. Because of the sag of pressure tube it can touch the calandria tube resulting in formation of cold spot. This leads to hydrogen concentration at that spot by which the material at that place becomes brittle and can lead to catastrophic failure of pressure tube. There is no useful access for measurement of annular gap either through the gas annular space or from exterior of calandria tube. So the annular gap was measured from inside surface of pressure tube which is accessible. Eddy current technique was used for finding the gap. The paper describe the details of split coil design of bobbin probe, selection of operating point on normalised impedance diagram by choosing frequency. Experimental results on full scale mock up, and actual gap measurement in reactor channel, are also given. (author). 7 figs

Dose distributions in electron irradiated plastic tubing

International Nuclear Information System (INIS)

Miller, A.; Pederson, W.B.

1981-01-01

Plastic tubes have been crosslinked by irradiation at a 10 MeV linear electron accelerator and at a 400 keV DC electron accelerator at different irradiation geometries. The diameter of the different tubes was 20, 33 and 110 millimeters. Dose distributions have been measured with thin radiochromic dye films, indicating that in all cases irradiation from two sides is a necessary and sufficient condition for obtaining a satisfactory dose distribution. (author)

Operating performance of CANDU pressure tubes

International Nuclear Information System (INIS)

Cheadle, B.A.; Price, E.G.

1989-04-01

The performance of Zircaloy-2 and Zr-2.5 Nb pressure tubes in CANDU reactors is reviewed. The accelerated hydriding of Zircaloy-2 in reducing water chemistries can lower the toughness of this material and it is essential that defect-initiating phenomena, such as hydride blister formation from pressure tube to calandria tube contact, be prevented. Zr-2.5 Nb pressure tubes are performing well with low rates of hydrogen pick-up and good retention of material properties

Plastic collapse of tubes submitted to a ring load by optimization

International Nuclear Information System (INIS)

Zouain, N.

1982-05-01

The limit analysis of a tube with finite lenght, made of a rigid - plastic material, is considered for the case of an internal load uniformely distributed in a cross section of the tube. The exact creep law is calculated for several qualitatively differents cases, namely different tube lenghts. The corresponding stress and collapse mechanisms are given so that they can be compared to the approximations developed here. The static and kinematic theorems on plastic collapse are used for establishing two numerical methods of resolution, specifically mathematical programming and finite element method. These mathematical methods are applied to collapse load for the considered tube. (E.G.) [pt

N Reactor pressure tube 1350 postirradiation examination

International Nuclear Information System (INIS)

Cook, D.J.

1977-01-01

The N Reactor pressure tubes were fabricated from Zircaloy-2 primarily due to the excellent corrosion resistance, low neutron absorption, and high strength properties of this alloy. Irradiation damage mechanisms increase the strength and decrease the ductility of the Zircaloy-2. Irradiation data available at the time the tubes were installed indicated that fast neutron irradiation damage mechanisms would not decrease the ductility to unacceptable levels over the estimated plant life of 25 to 30 years. However, because the tubes are a primary coolant system component and only limited data are available on irradiation effects at high fluences, a Postirradiation Examination (PIE) program was developed to assure that service factors do not compromise pressure tube integrity essential to reactor safety. The PIE program requires that a pressure tube be periodically removed from the reactor for destructive testing. The N Reactor Technical Specifications specify that the frequency of pressure tube removal and examination be based upon the previous PIE test results. Four pressure tubes were examined before tube 1350, and the test results were summarized in individual reports. PIE results on tube 1350 were summarized along with the test results on the previous four tubes in a previous report. The purpose of this report is to present in detail the results on PIE of pressure tube 1350, and, in particular, document the technique by which the fracture toughness of the pressure tube was determined

Adherence of volatile propofol to various types of plastic tubing.

Science.gov (United States)

Maurer, F; Lorenz, D J; Pielsticker, G; Volk, T; Sessler, D I; Baumbach, J I; Kreuer, S

2017-01-23

Propofol is an intravenous anesthetic. Currently, it is not possible to routinely measure blood concentration of the drug in real time. However, multi-capillary column ion-mobility spectrometry of exhaled gas can estimate blood propofol concentration. Unfortunately, adhesion of volatile propofol on plastic materials complicates measurements. Therefore, it is necessary to consider the extent to which volatile propofol adheres to various plastics used in sampling tubing. Perfluoralkoxy (PFA), polytetrafluorethylene (PTFE), polyurethane (PUR), silicone, and Tygon tubing were investigated in an experimental setting using a calibration gas generator (HovaCAL). Propofol gas was measured for one hour at 26 °C, 50 °C, and 90 °C tubing temperature. Test tubing segments were then flushed with N 2 to quantify desorption. PUR and Tygon sample tubing absorbed all volatile propofol. The silicone tubing reached the maximum propofol concentration after 119 min which was 29 min after propofol gas exposure stopped. The use of PFA or PTFE tubing produced comparable and reasonably accurate propofol measurements. The desaturation time for the PFA was 10 min shorter at 26 °C than for PTFE. PFA tubing thus seems most suitable for measurement of volatile propofol, with PTFE as an alternative.

Failure Pressure Estimates of Steam Generator Tubes Containing Wear-type Defects

International Nuclear Information System (INIS)

Yoon-Suk Chang; Jong-Min Kim; Nam-Su Huh; Young-Jin Kim; Seong Sik Hwang; Joung-Soo Kim

2006-01-01

It is commonly requested that steam generator tubes with defects exceeding 40% of wall thickness in depth should be plugged to sustain all postulated loads with appropriate margin. The critical defect dimensions have been determined based on the concept of plastic instability. This criterion, however, is known to be too conservative for some locations and types of defects. In this context, the accurate failure estimation for steam generator tubes with a defect draws increasing attention. Although several guidelines have been developed and are used for assessing the integrity of defected tubes, most of these guidelines are related to stress corrosion cracking or wall-thinning phenomena. As some of steam generator tubes are also failed due to fretting and so on, alternative failure estimation schemes for relevant defects are required. In this paper, three-dimensional finite element (FE) analyses are carried out under internal pressure condition to simulate the failure behavior of steam generator tubes with different defect configurations; elliptical wastage type, wear scar type and rectangular wastage type defects. Maximum pressures based on material strengths are obtained from more than a hundred FE results to predict the failure of the steam generator tube. After investigating the effect of key parameters such as wastage depth, wastage length and wrap angle, simplified failure estimation equations are proposed in relation to the equivalent stress at the deepest point in wastage region. Comparison of failure pressures predicted according to the proposed estimation scheme with some corresponding burst test data shows good agreement, which provides a confidence in the use of the proposed equations to assess the integrity of steam generator tubes with wear-type defects. (authors)

Tensile properties of quadruple melted Zr-2.5Nb pressure tubes evaluated from pressure tube offcuts

International Nuclear Information System (INIS)

Shah, Priti Kotak; Dubey, J.S.; Anantharaman, S.

2013-12-01

Rajasthan Atomic Power Station-2 (RAPS-2) is the first Pressurised Heavy Water Reactor (PHWR) in India having quadruple melted Zr-2.5Nb pressure tubes. Front-end and back-end off-cuts of sixteen pressure tubes were selected for studying the mechanical properties in axial and transverse directions of the tube. Tension tests were carried out at room temperature and at 300℃ using miniature tensile test specimens. The report presents the experimental details and discusses the base line tensile property data for the quadruple melted pressure tubes of RAPS-2. This data will be useful for the reactor life management. (author)

Dynamics of explosively imploded pressurized tubes

Science.gov (United States)

Szirti, Daniel; Loiseau, Jason; Higgins, Andrew; Tanguay, Vincent

2011-04-01

The detonation of an explosive layer surrounding a pressurized thin-walled tube causes the formation of a virtual piston that drives a precursor shock wave ahead of the detonation, generating very high temperatures and pressures in the gas contained within the tube. Such a device can be used as the driver for a high energy density shock tube or hypervelocity gas gun. The dynamics of the precursor shock wave were investigated for different tube sizes and initial fill pressures. Shock velocity and standoff distance were found to decrease with increasing fill pressure, mainly due to radial expansion of the tube. Adding a tamper can reduce this effect, but may increase jetting. A simple analytical model based on acoustic wave interactions was developed to calculate pump tube expansion and the resulting effect on the shock velocity and standoff distance. Results from this model agree quite well with experimental data.

Plastic Tubing and Maple Syrup Quality

Science.gov (United States)

Russell S. Walters; Harry W. Yawney

1978-01-01

Maple syrup made from sap collected using improperly or carelessly installed plastic pipelines varied more in color from day to day, and was more often darker in color, than sap collected from either the property installed pipeline or clean, frequently emptied galvanized buckets. Use of both properly installed tubing and buckets, following recommended procedures,...

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

Pressure loss characteristics of LSTF steam generator heat-transfer tubes. Pressure loss increase due to tube internal instruments

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro

1994-11-01

The steam generator of the Large-Scale Test Facility (LSTF) includes 141 heat-transfer U-tubes with different lengths. Six U-tubes among them are furnished with 15 or 17 probe-type instruments (conduction probe with a thermocouple; CPT) protuberant into the primary side of the U-tubes. Other 135 U-tubes are not instrumented. This results in different hydraulic conditions between the instrumented and non-instrumented U-tubes with the same length. A series of pressure loss characteristics tests was conducted at a test apparatus simulating both types of U-tube. The following pressure loss coefficient (K CPT ) was reduced as a function of Reynolds number (Re) from these tests under single-phase water flow conditions. K CPT =0.16 5600≤Re≤52820, K CPT =60.66xRe -0.688 2420≤Re≤5600, K CPT =2.664x10 6 Re -2.06 1371≤Re≤2420. The maximum uncertainty is 22%. By using these results, the total pressure loss coefficients of full length U-tubes were estimated. It is clarified that the total pressure loss of the shortest instrumented U-tube is equivalent to that of the middle-length non-instrumented U-tube and also that a middle-length instrumented U-tube is equivalent to the longest non-instrumented U-tube. Concludingly. it is important to take account of the CPT pressure loss mentioned above in estimation of fluid behavior at the non-instrumented U-tubes either by using the LSTF experiment data from the CPT-installed U-tubes or by using any analytical codes. (author)

Selected reading on introduction to pressure tube technology

International Nuclear Information System (INIS)

Causey, A.R.; Coleman, C.E.; Ells, C.E.

1981-10-01

Four lectures on pressure tube technology were presented at Sheridan Park, Ontario, on 1981 June 1. The titles were 'Pressure Tubes and Their Operational Environment', 'Fabrication, Inspection and Properties of Current Production Pressure Tubes', 'In-Reactor Deformation of Fuel Channels', and 'Potential Failure Modes in Pressure Tubes'. This report lists the references used in preparing the lectures. It is intended to provide a starting point in reading for people who need to become familiar with pressure tube technology but have little prior knowledge of the topic

Plasticity of pressure-sensitive materials

CERN Document Server

Ochsner, Andreas

2014-01-01

Classical plasticity theory of metals is independent of the hydrostatic pressure. However, if the metal contains voids or pores or if the structure is composed of cells, this classical assumption is no more valid and the influence of the hydrostatic pressure must be incorporated in the constitutive description. Looking at the microlevel, metal plasticity is connected with the uniform planes of atoms organized with long-range order. Planes may slip past each other along their close-packed directions. The result is a permanent change of shape within the crystal and plastic deformation. The presence of dislocations increases the likelihood of planes slipping. Nowadays, the theory of pressure sensitive plasticity is successfully applied to many other important classes of materials (polymers, concrete, bones etc.) even if the phenomena on the micro-level are different to classical plasticity of metals. The theoretical background of this phenomenological approach based on observations on the macro-level is describe...

N Reactor pressure tube 2566 postirradiation examination

International Nuclear Information System (INIS)

Scott, K.V.

1978-01-01

Pressure tube 2566 was removed from N Reactor in July, 1977 to initiate the postirradiation examination program required by the Technical Specifications. Destructive examination of the pressure tube, after a maximum accumulated fluence of 4.6 x 10 21 n/cm 2 (E > 1 MeV), was conducted at the Hanford Engineering Development Laboratory to determine the effects of reactor service on the mechanical properties and hydrogen absorption and corrosion characteristics of the pressure tube. Tube 2566 is the sixth tube removed for destructive examination since the initial reactor startup. Evaluation of test results reveal that no significant detrimental changes have occurred in the parameters studied, since the last tube was removed in 1974

Conceptual design for muon detectors using resistive plastic tubes. Final technical report

International Nuclear Information System (INIS)

Border, P.; Courant, H.; Heller, K.; Jones, A.; Lin, J.; Maxam, D.; Ruddick, K.

1998-01-01

Reliable low cost detectors which can be built in quantity require a simple design consisting of as few separate pieces as possible using inexpensive materials. For example, ordinary insulating plastics with good structural strength, such as polyethylene or polystyrene, have about 1/3 the cost of aluminum per unit weight. Since plastic is also about 1/3 the density of aluminum, the material cost for a drift tube would be reduced by an order of magnitude. This substitution of plastic for aluminum alone would save the muon system for the SDC more than $2M. Additional savings of greater magnitude can be expected since an entire drift tube, including a field shaping electrode structure, can be manufactured as a single piece by the technique of co-extrusion. A symmetric design with all walls far from the wire will also eliminate critical tolerances in the relative position of the electrodes with respect to the wire. Furthermore, module assembly and mounting costs will surely be reduced if the muon detectors were light weight and, as far as possible, had the same shape and size. With the 8 cm diameter plastic tube of the design, the electric drift field is nearly uniform as shown. This field is determined by a simple symmetric electrode structure, so that the necessary drift/position relationship can be achieved without precisely controlling the position of the electrode structure with respect to the wire. If the positioning of the electrode structure relative to the wire is not a critical dimension, the structural support for the tube need not be maintained to a high tolerance reducing the cost of the structure. Using a resistive plastic to shape the potential gives a simple electrode structure that will require a minimum number of electronic connections. The basic element of this design is the cylindrical plastic drift tube constructed from co-extruded plastics of different conductivity

Modelling of pressure tube Quench using PDETWO

International Nuclear Information System (INIS)

Parlatan, Y.; Lei, Q.M.; Kwee, M.

2004-01-01

Transient two-dimensional heat conduction calculations have been carried out to determine the time-dependent temperature distribution in an overheated pressure tube during quenching with water. The purpose of the calculations is to provide input for evaluation of thermal (secondary) stresses in the pressure tube due to quench. The quench phenomenon in pressure tubes could occur in several hypothetical accident scenarios, including incidents involving intermittent buoyancy-induced flow during outages. In these scenarios, there will be two (radial and axial) or three dimensional temperature gradients, resulting in thermal stresses in the pressure tube, as the water front reaches and starts to cool down the hot pressure tube. The transient, two-dimensional heat conduction equation in the pressure tube during quench is solved using a FORTRAN package called PDETWO, available in the open literature for solving time-dependent coupled systems of non-linear partial differential equations over a two-dimensional rectangular region. This routine is based on finite difference solution of coupled, non-linear partial differential equations. Temperature gradient in the circumferential gradient is neglected for conservatism and convenience. The advancing water front is not modelled explicitly, and assumed to be at a uniform temperature and moving at a constant velocity inferred from experimental data. For outer surface and both ends of the pressure tube in the axial direction, a zero-heat flux boundary condition is assumed, while for the inner surface a moving water-quench front is assumed by appropriately varying the fluid temperature and the heat transfer coefficient. The pressure tube is assumed to be at a uniform temperature of 400 o C initially, to represent conditions expected during an intermittent buoyancy-influenced flow scenario. The results confirm the expectations that axial temperature gradients and associated heat fluxes are small in comparison with those in the

Pressure vessels and methods of sealing leaky tubes disposed in pressure vessels

International Nuclear Information System (INIS)

Larson, G.C.

1980-01-01

This invention relates to pressure vessels and to methods of sealing leaky tubes in them and is especially applicable to pressure vessels in the form of sheet-and-tube type heat exchangers constructed with a large number of relatively small diameter tubes grouped in a bundle. To seal off a leaky tube in such a heat exchanger an explosive activated plug in the form of a hollow metal body is used, inserted at each end of the tube to be sealed. Using the arrangement of pressure vessel and associated tube sheets and the explosive activated plug method of sealing a leaky tube as described in this invention it is claimed that distortion of the adjacent tubes and the tube sheets is reduced when the explosive activated plugs are detonated. (U.K.)

Pressure tube rupture in a closed tank

International Nuclear Information System (INIS)

Khater, H.A.; Hadaller, G.I.; Stern, F.

1985-06-01

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

21 CFR 868.5860 - Pressure tubing and accessories.

Science.gov (United States)

2010-04-01

... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Pressure tubing and accessories. 868.5860 Section... (CONTINUED) MEDICAL DEVICES ANESTHESIOLOGY DEVICES Therapeutic Devices § 868.5860 Pressure tubing and accessories. (a) Identification. Pressure tubing and accessories are flexible or rigid devices intended to...

High-temperature transient creep properties of CANDU pressure tubes

International Nuclear Information System (INIS)

Fong, R.W.L.; Chow, C.K.

2002-06-01

During a hypothetical large break loss-of-coolant accident (LOCA), the coolant flow would be reduced in some fuel channels and would stagnate and cause the fuel temperature to rise and overheat the pressure tube. The overheated pressure tube could balloon (creep radially) into contact with its moderator-cooled calandria tube. Upon contact, the stored thermal energy in the pressure tube is transferred to the calandria tube and into the moderator, which acts as a heat sink. For safety analyses, the modelling of fuel channel deformation behaviour during a large LOCA requires a sound knowledge of the high-temperature creep properties of Zr-2.5Nb pressure tubes. To this extent, a ballooning model to predict pressure-tube deformation was developed by Shewfelt et al., based on creep equations derived using uniaxial tensile specimens. It has been recognized, however, that there is an inherent variability in the high-temperature creep properties of CANDU pressure tubes. The variability, can be due to different tube-manufacturing practices, variations in chemical compositions, and changes in microstructure induced by irradiation during service in the reactor. It is important to quantify the variability of high-temperature creep properties so that accurate predictions on pressure-tube creep behaviour can be made. This paper summarizes recent data obtained from high-temperature uniaxial creep tests performed on specimens taken from both unirradiated (offcut) and irradiated pressure tubes, suggesting that the variability is attributed mainly to the initial differences in microstructure (grain size, shape and preferred orientation) and also from tube-to-tube variations in chemical composition, rather than due to irradiation exposure. These data will provide safety analysts with the means to quantify the uncertainties in the prediction of pressure-tube contact temperatures during a postulated large break LOCA. (author)

Studies of the stability and sistematics of operation of the DELPHI plastic streamer tubes

International Nuclear Information System (INIS)

Alekseev, G.D.; Bilenky, M.S.; Bonyushkin, Yu.Ye.; Korrytov, A.V.

1989-01-01

The influeNce of gas pressure and temperature, slight changes in a gas mixture, self-sustaining discharge, aging, different deviations in the detector geometry on the operation of the DELPHI plastic streamer tubes is studied. The contribution of these factors to the DELPHI hadron calorimeter energy resolution is estimated. The major influence is found to be due to atmospheric pressure fluctuations (∼ ±1% streamer charge change per ±1 Torr). The question of the choice of a gas mixture is studied as well. Maximal streamer charge is shown to be independent of component concentrations in the Ar:CO 2 :i-C 4 H 10 ) or C 5 H 12 ) gas mixtures. 23 refs.; 21 figs.; 1 tab

Plastic Deformation of Pressured Metallic Glass

Directory of Open Access Journals (Sweden)

Yun Cheng

2017-11-01

Full Text Available Although pressured metallic glass (MG has been reported in the literature; there are few studies focusing on pressure effects on the structure; dynamics and its plastic deformation. In this paper; we report on and characterize; via molecular dynamics simulation, the structure and dynamics heterogeneity of pressured MGs, and explore a causal link between local structures and plastic deformation mechanism of pressured glass. The results exhibit that the dynamical heterogeneity of metallic liquid is more pronounced at high pressure, while the MGs were less fragile after the release of external pressure, reflected by the non-Gaussian parameter (NGP. High pressure glass shows better plastic deformation; and the local strain zone distributed more uniformly than of in normal glass. Further research indicates that although the number of icosahedrons in pressured glass was much larger than that in normal glass, while the interpenetrating connections of icosahedra (ICOI exhibited spatial correlations were rather poor; In addition, the number of ‘fast’ atoms indexed by the atoms’ moving distance is larger than that in normal glass; leading to the sharp decreasing in number of icosahedrons during deformation. An uniform distribution of ‘fast’ atoms also contributed to better plastic deformation ability in the pressured glass. These findings may suggest a link between the deformation and destruction of icosahedra with short-range order.

Pressure tube reactors

International Nuclear Information System (INIS)

Natori, Hisahide.

1981-01-01

Purpose: To improve the electrical power generation efficiency in a pressure tube reactor in which coolants and moderators are separated by feedwater heating with heat generated in heavy water and by decreasing the amount of steams to be extracted from the turbine. Constitution: A heat exchanger and a heavy water cooler are additionally provided to a conventional pressure tube reactor. The heat exchanger is disposed at the pre-stage of a low pressure feedwater heater series. High temperature heavy water heated in the core is passed through the primary side of the exchanger, while feedwater is passed through the secondary side. The cooler is disposed on the downstream of the heat exchanger in the flowing direction of the heavy water, in which heavy water from the heat exchanger is passed through the primary side and the auxiliary equipment cooling water is sent to the secondary side thereof. Accordingly, since extraction of heating steams is no more necessary, the steam can be used for the rotation of the turbine, and the electrical power generation efficiency can be improved. (Seki, T.)

Failure maps for internally pressurized Zr-2.5% Nb pressure tubes with circumferential temperature variations

International Nuclear Information System (INIS)

Shewfelt, R.S.W.

1986-01-01

During some postulated loss-of-coolant accidents, the pressure tube temperature may rise before the internal pressure drops, causing the pressure tube to balloon. The temperature around the pressure tube circumference would likely be nonuniform, producing localized deformation that could possibly cause failure. The computer program, GRAD, was used to determine the circumferential temperature distribution required to cause an internally pressurized Zr-2.5% Nb pressure tube to fail before coming into full contact with its calandria tube. These results were used to construct failure maps. 7 refs

Plastic instability criteria for necking of bars and ballooning of tubes

International Nuclear Information System (INIS)

Lin, E.I.H.

1977-01-01

Plastic instability criteria applicable to the necking of bars under tension and to the ballooning of thin-wall tubes under internal pressure were derived from basic geometrical considerations. In the case of bars under tension, plastic instability prevails if the percentage rate of decrease of the cross-sectional area in the (potential) necking region is greater than that in the bulk of the bar. When the loading characteristics and constitutive equation were taken into account, an instability criterion was deduced in terms of the stress, strain, strain rate, temperature and material properties. This criterion was shown to be reducible to the classical Considere condition for non-rate-sensitive materials. For rate-sensitive materials under isothermal conditions, a simple relationship among the strain, the strain-hardening and strain-rate-sensitivity parameters was also obtained. It was found that the uniform elongation decreases with increasing strainrate sensitivity, a conclusion which is in agreement with experimental measurements and some previous investigations. Finally, the relationship between the high strainrate sensitivity and the superplastic ductility of a material was explained without invoking any non-hardening arguments. (Auth.)

Development of heat treated Zr-2.5% Nb alloy tubes for pressure tubes

International Nuclear Information System (INIS)

Saibaba, N.; Jha, S.K.; Tonpe, S.

2011-01-01

Zr-2.5% Nb alloy is the candidate material for pressure tubes of Pressurized Heavy Water Reactors (PHWR), and are manufactured in cold working condition while heat treated pressure tubes are used in RBMK and FUGEN type of reactors. The diametral creep of these tubes is the life limiting factor. This paper presents the extensive work carried out for the optimization of process parameters to manufacture heat treated Zr-2.5% Nb pressure tubes. Extensive dilactometry study was carried out to establish the transus temperature for the alloy and the effect of soaking temperature and cooling rate on the microstructure was characterized. On the basis of the study, water quenching (at 883 deg C) in the a b region with 20-25% primary a phase was selected, further cold worked, aged and finally autoclaved. Mechanical properties of the finished tubes were found to be comparable to the cold worked route. Large number of full sized tubes of about 700 - 800 mm long was produced to establish the repeatability. (author)

Stress analysis of pressurized water reactor steam generator tube denting phenomena. Interim report

International Nuclear Information System (INIS)

Thomas, J.M.; Cipolla, R.C.; Ranjan, G.V.; Derbalian, G.

1978-07-01

In some Pressurized Water Reactor (PWR) steam generators, a corrosion product has formed on the carbon steel support plate in the crevice between the tube and support plate. The corrosion product occupies more volume than the original metal; the tube-to-support plate crevice volume is thus consumed with corrosion product, and further corrosive action results in a radially inward force on the tube and a radially outward force on the corroding support plate. This has resulted in indentation (''denting'') of the tube, accompanied by occasional cracking. Large in-plane deformation and cracking of support plates has also been observed in the most severely affected plants along with some serious side effects, such as deformation and cracking of inner row tube U-bends caused by support plate movement. Mechanical aspects of the tube denting phenomena have been studied using analytical models. The models used ranged from closed form analytical solutions to state-of-the-art numerical elastic-plastic computer program for moderate strains. It was found that tube dents, such as those observed in operating steam generators, are associated with yielding of both the tubes and support plates. Also studied were the stresses in tube U-bends caused by support plate flow slot deformation

Probabilistic integrity assessment of pressure tubes in an operating pressurized heavy water reactor

Energy Technology Data Exchange (ETDEWEB)

Oh, Young-Jin; Park, Heung-Bae [KEPCO E and C, 188 Gumi-dong, Bundang-gu, Seongnam-si, Gyeonggi-do 463-870 (Korea, Republic of); Lee, Jung-Min; Kim, Young-Jin [School of Mechanical Engineering, Sungkyunkwan University, 300 Chunchun-dong, Jangan-gu, Suwon-si, Gyeonggi-do 440-746 (Korea, Republic of); Ko, Han-Ok [Korea Institute of Nuclear Safety, 34 Gwahak-ro, Yuseong-gu, Daejeon-si 305-338 (Korea, Republic of); Chang, Yoon-Suk, E-mail: yschang@khu.ac.kr [Department of Nuclear Engineering, Kyung Hee University, 1 Seocheon-dong, Giheung-gu, Yongin-si, Gyeonggi-do 446-701 (Korea, Republic of)

2012-02-15

Even though pressure tubes are major components of a pressurized heavy water reactor (PHWR), only small proportions of pressure tubes are sampled for inspection due to limited inspection time and costs. Since the inspection scope and integrity evaluation have been treated by using a deterministic approach in general, a set of conservative data was used instead of all known information related to in-service degradation mechanisms because of inherent uncertainties in the examination. Recently, in order that pressure tube degradations identified in a sample of inspected pressure tubes are taken into account to address the balance of the uninspected ones in the reactor core, a probabilistic approach has been introduced. In the present paper, probabilistic integrity assessments of PHWR pressure tubes were carried out based on accumulated operating experiences and enhanced technology. Parametric analyses on key variables were conducted, which were periodically measured by in-service inspection program, such as deuterium uptake rate, dimensional change rate of pressure tube and flaw size distribution. Subsequently, a methodology to decide optimum statistical distribution by using a robust method adopting a genetic algorithm was proposed and applied to the most influential variable to verify the reliability of the proposed method. Finally, pros and cons of the alternative distributions comparing with corresponding ones derived from the traditional method as well as technical findings from the statistical assessment were discussed to show applicability to the probabilistic assessment of pressure tubes.

Development of extruded resistive plastic tubes for proportional chamber cathodes

International Nuclear Information System (INIS)

Kondo, K.

1982-01-01

Carbon mixed plastic tubes with resistivity of 10 3 approx. 10 4 Ωcm have been molded with an extrusion method and used for the d.c. cathode of a proportional counter and a multi-wire proportional chamber. The signal by gas multiplication was picked up from a strip r.f. cathode set outside the tube. The characteristics of the counter in the proportional and limited streamer modes have been studied

Performance of pressure tubes in CANDU reactors

Energy Technology Data Exchange (ETDEWEB)

Rodgers, D.; Griffiths, M.; Bickel, G.; Buyers, A.; Coleman, C.; Nordin, H.; St Lawrence, S. [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)

2016-06-15

The pressure tubes in CANDU reactors typically operate for times up to about 30 years prior to refurbishment. The in-reactor performance of Zr-2.5Nb pressure tubes has been evaluated by sampling and periodic inspection. This paper describes the behavior and discusses the factors controlling the behaviour of these components. The Zr–2.5Nb pressure tubes are nominally extruded at 815{sup o}C, cold worked nominally 27%, and stress relieved at 400 {sup o}C for 24 hours, resulting in a structure consisting of elongated grains of hexagonal close-packed alpha-Zr, partially surrounded by a thin network of filaments of body-centred-cubic beta-Zr. These beta-Zr filaments are meta-stable and contain about 20% Nb after extrusion. The stress-relief treatment results in partial decomposition of the beta-Zr filaments with the formation of hexagonal close-packed alpha-phase particles that are low in Nb, surrounded by a Nb-enriched beta-Zr matrix. The material properties of pressure tubes are determined by variations in alpha-phase texture, alpha-phase grain structure, network dislocation density, beta-phase decomposition, and impurity concentration that are a function of manufacturing variables. The pressure tubes operate at temperatures between 250 {sup o}C and 310 {sup o}C with coolant pressures up to about 11 MPa in fast neutron fluxes up to 4 x 10{sup 17} n·m{sup -2}·s{sup -1} (E > 1 MeV) and the properties are modified by these conditions. The properties of the pressure tubes in an operating reactor are therefore a function of both manufacturing and operating condition variables. The ultimate tensile strength, fracture toughness, and delayed hydride-cracking properties (velocity (V) and threshold stress intensity factor (K{sub IH})) change with irradiation, but all reach a nearly limiting value at a fluence of less than 10{sup 25} n·m{sup -2} (E > 1 MeV). At this point the ultimate tensile strength is raised about 200 MPa, toughness is reduced by about 50%, V increases

Sample summary report for ARG 1 pressure tube sample

International Nuclear Information System (INIS)

Belinco, C.

2006-01-01

The ARG 1 sample is made from an un-irradiated Zr-2.5% Nb pressure tube. The sample has 103.4 mm ID, 112 mm OD and approximately 500 mm length. A punch mark was made very close to one end of the sample. The punch mark indicates the 12 O'clock position and also identifies the face of the tube for making all the measurements. ARG 1 sample contains flaws on ID and OD surface. There was no intentional flaw within the wall of the pressure tube sample. Once the flaws are machined the pressure tube sample was covered from outside to hide the OD flaws. Approximately 50 mm length of pressure tube was left open at both the ends to facilitate the holding of sample in the fixtures for inspection. No flaw was machined in this zone of 50 mm on either end of the pressure tube sample. A total of 20 flaws were machined in ARG 1 sample. Out of these, 16 flaws were on the OD surface and the remaining 4 on the ID surface of the pressure tube. The flaws were characterized in to various groups like axial flaws, circumferential flaws, etc

LPG based all plastic pressure sensor

DEFF Research Database (Denmark)

Bundalo, Ivan-Lazar; Lwin, R.; Leon-Saval, S.

2015-01-01

A prototype all-plastic pressure sensor is presented and characterized for potential use as an endoscope. The sensor is based on Long Period Gratings (LPG) inscribed with a CO2 laser in 6-ring microstructured PMMA fiber. Through a latex coated, plastic 3D-printed transducer pod, external pressure...... is converted to longitudinal elongation of the pod and therefore of the fiber containing the LPG. The sensor has been characterised for pressures of up to 160 mBar in an in-house built pressure chamber. Furthermore, the influence of the fiber prestrain, fiber thickness and the effect of different glues...

In situ sampling for pressure tube deuterium concentration

International Nuclear Information System (INIS)

Harrington, A.J.; Kittmer, C.A.

1988-01-01

The present method of assessing the useful life of pressure tubes in CANDU (CANada Deuterium Uranium) reactors requires the periodic removal and examination of a tube. Special tooling was developed at Atomic Energy of Canada Limited (AECL) to obtain a sample of material from a pressure tube without removing the tube from the reactor. The sampling tool concept has been successfully used by Ontario Hydro during scheduled outages at the Pickering Nuclear Generating Station (PNGS). (author)

Ballooning of CANDU pressure tube in local thermal transients

International Nuclear Information System (INIS)

Mihalache, Maria; Ionescu, Viorel

2008-01-01

In certain LOCA scenarios for the CANDU fuel channel, the ballooning of the pressure tube and contact with the calandria tube can occur. After the contact moment, a radial heat transfer from cooling fluid to moderator takes place through the contact area. If the temperature of channel walls increases, the contact area is drying and the heat transfer becomes inefficiently. In INR-Pitesti the DELOCA code was developed to simulate the mechanical behaviour of pressure tube during pre-contact transition, and mechanical and thermal behaviour of pressure tube and calandria tube after occurrence of the contact between the two tubes. The code contains few models: thermal creep of Zr-2.5%Nb alloy, the heat transfer by conduction through the cylindrical walls, channel failure criteria and calculus of heat transfer at the calandria tube - moderator interface. This code evaluates the contact and channel failure moments. This paper gives a DELOCA code description and the fuel channel behaviour analysis, in transient temperature conditions of the pressure tube, using the materials properties, time and temperature dependencies of these properties as obtained in the different laboratories of the world and in the INR - Pitesti in the last years. DELOCA computer code simulated the fuel channel response to the constant heating rates of inside pressure tube surface. The paper presents contact temperature and time dependencies on the heating rate, and the appropriate fitting functions. (authors)

Plastic deformation and fracture behavior of zircaloy-2 fuel cladding tubes under biaxial stress

International Nuclear Information System (INIS)

Maki, Hideo; Ooyama, Masatosi

1975-01-01

Various combinations of biaxial stress were applied on five batches of recrystallized zircaloy-2 fuel cladding tubes with different textures; elongation in both axial and circumferential directions of the specimen was measured continuously up to 5% plastic deformation. The anisotropic theory of plasticity proposed by Hill was applied to the resulting data, and anisotropy constants were obtained through the two media of plastic strain loci and plastic strain ratios. Comparison of the results obtained with the two methods proved that the plastic strain loci provide data that are more effective in predicting quantitatively the plastic deformation behavior of the zircaloy-2 tubes. The anisotropy constants change their value with progress of plastic deformation, and judicious application of the effective stress and effective strain obtained on anisotropic materials will permit the relationship between stress and strain under various biaxialities of stresses to be approximated by the work hardening law. The test specimens used in the plastic deformation experiments were then stressed to fracture under the same combination of biaxial stress as in the proceeding experiments, and the deformation in the fractured part was measured. The result proved that the tilt angle of the c-axis which serves as the index of texture is related to fracture ductility under biaxial stress. Based on this relationship, it was concluded that material with a tilt angle ranging from 10 0 to 15 0 is the most suitable for fuel cladding tubes, from the viewpoint of fracture ductility, at least in the case of unirradiated material. (auth.)

49 CFR 192.191 - Design pressure of plastic fittings.

Science.gov (United States)

2010-10-01

... 49 Transportation 3 2010-10-01 2010-10-01 false Design pressure of plastic fittings. 192.191... Components § 192.191 Design pressure of plastic fittings. (a) Thermosetting fittings for plastic pipe must conform to ASTM D 2517, (incorporated by reference, see § 192.7). (b) Thermoplastic fittings for plastic...

Rejection index for pressure tubes

International Nuclear Information System (INIS)

Mitchell, A.B.; Meneley, D.

1989-10-01

The objective of the present study was to establish a set of criteria (or Rejection Index) which could be used to decide whether a zirconium-2 1/2 w/o niobium pressure tube in a CANDU reactor should be removed from service due to in-service degradation. A critique of key issues associated with establishing a realistic rejection index was prepared. Areas of uncertainty in available information were identified and recommendations for further analysis and laboratory testing made. A Rejection Index based on the following limits has been recommended: 1) Limits related to design intent and normal operation: any garter spring must remain within the tolerance band specified for its design location; the annulus gas system must normally be operated in a circulating mode with a procedure in place for purging to prevent accumulation of deuterium. It must remain sensitive to leaks into any part of the systems; and pressure tube dimensions and distortions must be limited to maintain the fuel channels within the original design intent; 2) Limits related to defect tolerance: adequate time margins between occurrence of a leaking crack and unstable failure must be demonstrated for all fuel channels; long lap-type flaws are unacceptable; crack-like defects of any size are unacceptable; and score marks, frat marks and other defects with contoured profiles must fall below certain depth, length and stress intensity limits; and 3) Limits related to property degradation: at operating temperature each pressure tube must be demonstrated to have a critical length in excess of a stipulated value; the maximum equivalent hydrogen level in any pressure tube should not exceed a limit which should be defined taking into account the known history of that tube; the maximum equivalent hydrogen level in any rolled joint should not exceed a limit which is presently recommended as 200 ppm equivalent hydrogen; and the maximum diametral creep strain should be limited to less than 5%

Bulging of pressure tubes at hot spots under LOCA conditions

International Nuclear Information System (INIS)

Manu, C.; Shewfelt, R.S.W.; Wright, A.C.D.; Aboud, R.; Lau, J.H.K.; Sanderson, D.B.

1996-01-01

During certain postulated loss-of-coolant accidents (LOCA) in a CANDU reactor, some fuel channels can become highly voided within a very short time. Although the pressure tubes are heated mainly by convection and thermal radiation during the LOCA transient, additional heat flow occurs through the bearing pads that are in contact with the pressure tribe. This contact can lead to local hot spots and associated thermal stresses in the pressure tube wall. The two factors that affects the behavior of the pressure tubes during LOCA conditions are the internal pressure and the local heating. Although the effect of internal pressure and of axially uniform temperature has been studied elsewhere, the effect of the local heating on the pressure tube behavior has not been modelled before. This paper shows that the bulging of a pressure tube at a hot spot is the result of the thermal stresses that are developed in a pressure tube during a LOCA transient. To isolate the local heating effect from the internal pressure, a series of single-effect experiments was performed. In these experiments, sections of a CANDU pressure tube were subjected to local heating only. The thermal profile and the local deformation were measured function of time. To quantify the effect of the thermal stresses on the bulging of pressure tubes at hot spots and to develop numerical tools that can predict such bulging, finite element analyses were performed rising the ABAQUS finite element computer code. Use of the measured thermal profiles in the ABAQUS finite element analysis, resulted in very good agreement between the predicted and measured displacements. (author)

Endotracheal tube cuff pressure monitoring during neurosurgery - Manual vs. automatic method

Directory of Open Access Journals (Sweden)

Mukul Kumar Jain

2011-01-01

Full Text Available Background: Inflation and assessment of the endotracheal tube cuff pressure is often not appreciated as a critical aspect of endotracheal intubation. Appropriate endotracheal tube cuff pressure, endotracheal intubation seals the airway to prevent aspiration and provides for positive-pressure ventilation without air leak. Materials and Methods: Correlations between manual methods of assessing the pressure by an experienced anesthesiologists and assessment with maintenance of the pressure within the normal range by the automated pressure controller device were studied in 100 patients divided into two groups. In Group M, endotracheal tube cuff was inflated manually by a trained anesthesiologist and checked for its pressure hourly by cuff pressure monitor till the end of surgery. In Group C, endotracheal tube cuff was inflated by automated cuff pressure controller and pressure was maintained at 25-cm H 2 O throughout the surgeries. Repeated measure ANOVA was applied. Results: Repeated measure ANOVA results showed that average of endotracheal tube cuff pressure of 50 patients taken at seven different points is significantly different (F-value: 171.102, P-value: 0.000. Bonferroni correction test shows that average of endotracheal tube cuff pressure in all six groups are significantly different from constant group (P = 0.000. No case of laryngomalacia, tracheomalacia, tracheal stenosis, tracheoesophageal fistula or aspiration pneumonitis was observed. Conclusions: Endotracheal tube cuff pressure was significantly high when endotracheal tube cuff was inflated manually. The known complications of high endotracheal tube cuff pressure can be avoided if the cuff pressure controller device is used and manual methods cannot be relied upon for keeping the pressure within the recommended levels.

Frictional pressure drop of high pressure steam-water two-phase flow in internally helical ribbed tubes

International Nuclear Information System (INIS)

Tingkuan, C.; Xuanzheng, C.

1987-01-01

It is well known that the internally helical ribbed tubes are effective in suppressing the dry-out in boiling tubes at high pressures, so they are widely used as furnace water wall tubes in modern large steam power boilers. Design of the boilers requires the data on frictional pressure drop characteristics of the ribbed tubes, but they are not sufficient now. This paper describes the experimental results on the adiabatic frictional pressure drop in both horizontal ribbed tubes with measured mean inside diameter of 11.69 mm and 35.42 mm at high pressure from 10 to 21 MPa, mass flow rate from 350 to 3800 kg/m/sup 2/s and steam quality from 0 to 1 in our high pressure electrically heated water loop. Simultaneously, both smooth tubes under the same conditions for comparison. Based on the tests the correlation for determining the frictional pressure drop of internally ribbed tubes are proposed

Remotised sliver sample removal from irradiated pressure tube

Energy Technology Data Exchange (ETDEWEB)

Rupani, B B; Sharma, B S.V.G.; Shyam, T V; Sinha, R K [Bhabha Atomic Research Centre, Bombay (India). Reactor Engineering Div.

1994-12-31

The life of irradiated pressure tubes of Pressurised Heavy Water Reactor (PHWR) can be assessed by analysing hydrogen content in sliver samples of pressure tube material. The sample can be obtained in the form of small slivers by scraping at any specified locations within the bore of the pressure tube of operating reactor. A tool namely Hydride Scraping Tool (HST) has been developed to obtain very thin slivers of hydride bearing samples from irradiated pressure tube. In order to save man-rem consumption during scraping operation, a feeding mechanism has also been designed and developed for axial positioning of scraping tool in the channel remotely. This paper covers general details about constructional features of the scraping tool, feeding mechanism and their control system. It also highlights the operational salient features and capabilities of the system. Possible applications of the feeding mechanism in other fields are also indicated. (author). 4 figs.

Remotised sliver sample removal from irradiated pressure tube

International Nuclear Information System (INIS)

Rupani, B.B.; Sharma, B.S.V.G.; Shyam, T.V.; Sinha, R.K.

1994-01-01

The life of irradiated pressure tubes of Pressurised Heavy Water Reactor (PHWR) can be assessed by analysing hydrogen content in sliver samples of pressure tube material. The sample can be obtained in the form of small slivers by scraping at any specified locations within the bore of the pressure tube of operating reactor. A tool namely Hydride Scraping Tool (HST) has been developed to obtain very thin slivers of hydride bearing samples from irradiated pressure tube. In order to save man-rem consumption during scraping operation, a feeding mechanism has also been designed and developed for axial positioning of scraping tool in the channel remotely. This paper covers general details about constructional features of the scraping tool, feeding mechanism and their control system. It also highlights the operational salient features and capabilities of the system. Possible applications of the feeding mechanism in other fields are also indicated. (author). 4 figs

Pressure Tube and Pressure Vessel Reactors; certain comparisons

Energy Technology Data Exchange (ETDEWEB)

Margen, P H; Ahlstroem, P E; Pershagen, B

1961-04-15

In a comparison between pressure tube and pressure vessel type reactors for pressurized D{sub 2}O 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 D{sub 2}O 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.

Pressure Tube and Pressure Vessel Reactors; certain comparisons

International Nuclear Information System (INIS)

Margen, P.H.; Ahlstroem, P.E.; Pershagen, B.

1961-04-01

In a comparison between pressure tube and pressure vessel type reactors for pressurized D 2 O 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 D 2 O 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

Development of technology on the material surveillance of CANDU pressure tubes

Energy Technology Data Exchange (ETDEWEB)

Noh, Kye Hoh; Han, Jung Hoh; Lee, Duk Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-05-01

Material degradation of pressure tubes, which are the most important components in CANDU fuel channel, can only be evaluated by removing and examining them(material surveillance). This study aimed at establishment of overall evaluation technology including the evaluation of the material degradation for the integrity of pressure tubes of Wolsung units. Material tests for pressure tubes were performed as follows; (1) Evaluation on life limiting factors of pressure tubes (2) Review on leak-before-break and integrity maintenance technology of pressure tubes (3) Survey on selection criteria for tubes to be inspected and on related regulations for material surveillance (4) Analysis of material surveillance test procedure (5) Basic examinations of Wolsung unit 1 pressure tube material(TEM, texture, chemical component etc) (6) Manufacture of test equipments and test (DHCV, hydriding, grip and tensile specimen etc). 23 figs, 6 tabs, 59 refs. (Author).

Development of technology on the material surveillance of CANDU pressure tubes

International Nuclear Information System (INIS)

Noh, Kye Hoh; Han, Jung Hoh; Lee, Duk Hyun

1995-05-01

Material degradation of pressure tubes, which are the most important components in CANDU fuel channel, can only be evaluated by removing and examining them(material surveillance). This study aimed at establishment of overall evaluation technology including the evaluation of the material degradation for the integrity of pressure tubes of Wolsung units. Material tests for pressure tubes were performed as follows; (1) Evaluation on life limiting factors of pressure tubes (2) Review on leak-before-break and integrity maintenance technology of pressure tubes (3) Survey on selection criteria for tubes to be inspected and on related regulations for material surveillance (4) Analysis of material surveillance test procedure (5) Basic examinations of Wolsung unit 1 pressure tube material(TEM, texture, chemical component etc) (6) Manufacture of test equipments and test (DHCV, hydriding, grip and tensile specimen etc). 23 figs, 6 tabs, 59 refs. (Author)

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

Absorbed Dose Distributions in Irradiated Plastic Tubing and Wire Insulation

DEFF Research Database (Denmark)

Miller, Arne; McLaughlin, W. L.

1979-01-01

Plastic tubing and wire insulation were simulated by radiochromic dye dosimeter films having electron absorbing properties similar to the materials of interest (polyethylene and PVC). A 400-keV electron accelerator was used to irradiate from 1, 2, 3 and 4 sides simulating possible industrial...

Advanced pressure tube sampling tools

International Nuclear Information System (INIS)

Wittich, K.C.; King, J.M.

2002-01-01

Deuterium concentration is an important parameter that must be assessed to evaluate the Fitness for service of CANDU pressure tubes. In-reactor pressure tube sampling allows accurate deuterium concentration assessment to be made without the expenses associated with fuel channel removal. This technology, which AECL has developed over the past fifteen years, has become the standard method for deuterium concentration assessment. AECL is developing a multi-head tool that would reduce in-reactor handling overhead by allowing one tool to sequentially sample at all four axial pressure tube locations before removal from the reactor. Four sets of independent cutting heads, like those on the existing sampling tools, facilitate this incorporating proven technology demonstrated in over 1400 in-reactor samples taken to date. The multi-head tool is delivered by AECL's Advanced Delivery Machine or other similar delivery machines. Further, AECL has developed an automated sample handling system that receives and processes the tool once out of the reactor. This system retrieves samples from the tool, dries, weighs and places them in labelled vials which are then directed into shielded shipping flasks. The multi-head wet sampling tool and the automated sample handling system are based on proven technology and offer continued savings and dose reduction to utilities in a competitive electricity market. (author)

Neutron radiography of irradiated zircaloy coupons of pressure tubes from PHWR`s

Energy Technology Data Exchange (ETDEWEB)

Gangotra, S; Ouseph, P M; Tamhane, A B; Singh, H N; Sahoo, K C [Bhabha Atomic Research Centre, Bombay (India). Radiometallurgy Div.

1994-12-31

The Indian Pressurised Heavy Water Reactors (PHWR`s) are of CANDU type, consisting of 304 zircaloy-2 pressure tubes. These pressure tubes contain the fuel bundles, where the heat is generated and is removed by the heavy water flowing through these pressure tubes at high temperature and pressure. These pressure tubes are surrounded by the calandria tubes, and are separated from them by a pair of garter springs. Over a period of time, as a result of the irradiation creep and assisted by the displacement of the garter springs, the hot pressure tube may come in contact with the cold calandria tube. This would result in the hydrogen migrating to the cold contact location and formation of hydride blisters. These blisters could eventually rupture the pressure tube by the DHC (delayed hydrogen cracking) mechanism. 2 refs., 2 figs.

Innovation in pressure tube life assessment

International Nuclear Information System (INIS)

Guler, B.; Kalenchuk, D.; Celovsky, A.

2003-01-01

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

Simulation and analysis of the thermal and deformation behaviour of `as-received` and `hydrided` pressure tubes used in the circumferential temperature distribution experiments (end of life/pressure tube behaviour)

Energy Technology Data Exchange (ETDEWEB)

Muir, W C; Bayoumi, M H [Ontario Hydro, Toronto, ON (Canada)

1996-12-31

It is postulated that in-reactor pressure tubes may be subjected to radiation damage and dissolved deuterium which could change the pressure tube characteristics and lead to different behaviour than that of as-received pressure tubes under large LOCA (loss of coolant) conditions. A hydrided pressure tube was used to study the effect of dissolved hydrogen on thermal-mechanical behaviour. In the experiment, simulating an in-reactor (hydrided) pressure tube with circumferential differential temperature under boil-off conditions, the pressure tube ballooned into contact with the calandria tube. The pressure tube used in this experiment was hydrided in a furnace to a nominal value of 200 {mu}g/g dissolved hydrogen. This test was a repeat of the first supplementary boil-off test (S-5-1) which used an as-received pressure tube. The objective of this paper is to analyze the results obtained from the simulation of this Boil-Off test using the SMARTT computer code and to examine the effect of hydriding on the thermal and ballooning behaviour of the pressure tube by comparison with the results obtained from test S-5-1. A discussion of the results obtained from this comparison is presented together with an analysis of their application to the analysis of pressure tube behaviour in CANDU reactors. (author). 13 refs., 1 tab., 16 figs.

Simulation and analysis of the thermal and deformation behaviour of 'as-received' and 'hydrided' pressure tubes used in the circumferential temperature distribution experiments (end of life/pressure tube behaviour)

International Nuclear Information System (INIS)

Muir, W.C.; Bayoumi, M.H.

1995-01-01

It is postulated that in-reactor pressure tubes may be subjected to radiation damage and dissolved deuterium which could change the pressure tube characteristics and lead to different behaviour than that of as-received pressure tubes under large LOCA (loss of coolant) conditions. A hydrided pressure tube was used to study the effect of dissolved hydrogen on thermal-mechanical behaviour. In the experiment, simulating an in-reactor (hydrided) pressure tube with circumferential differential temperature under boil-off conditions, the pressure tube ballooned into contact with the calandria tube. The pressure tube used in this experiment was hydrided in a furnace to a nominal value of 200 μg/g dissolved hydrogen. This test was a repeat of the first supplementary boil-off test (S-5-1) which used an as-received pressure tube. The objective of this paper is to analyze the results obtained from the simulation of this Boil-Off test using the SMARTT computer code and to examine the effect of hydriding on the thermal and ballooning behaviour of the pressure tube by comparison with the results obtained from test S-5-1. A discussion of the results obtained from this comparison is presented together with an analysis of their application to the analysis of pressure tube behaviour in CANDU reactors. (author). 13 refs., 1 tab., 16 figs

Delayed hydrogen cracking of zirconium alloy pressure tubes

International Nuclear Information System (INIS)

Jackman, A.H.; Dunn, J.T.

1976-10-01

After several years of almost continuous service, Pickering Units 3 and 4 have both experienced long outages to replace cracked pressure tubes. This report summarizes the status of the investigation into the cause of the cracks as of May 1976. The basic cause of the cracking was the presence of very high residual tensile stresses in the pressure tubes due to improper rolling procedures. These residual stresses are being reduced to acceptable levels by local stress relieving techniques at Bruce G.S. and in future reactors improvements in rolling procedures and changes in pressure tube specifications will prevent a recurrence of this problem. (author)

Tube micro-fouling, boiling and steam pressure after chemical cleaning

International Nuclear Information System (INIS)

Hu, M.H.

1998-01-01

This paper presents steam pressure trends after chemical cleaning of steam generator tubes at four plants. The paper also presents tube fouling factor that serves as an objective parameter to assess tubing boiling conditions for understanding the steam pressure trend. Available water chemistry data helps substantiate the concept of tube micro-fouling, its effect on tubing boiling, and its impact on steam pressure. All four plants experienced a first mode of decreasing steam pressure in the post-cleaning operation. After 3 to 4 months of operation, the decreasing trend stopped for three plants and then restored to a pre-cleaning value or better. The fourth plant is soil in decreasing trend after 12 months of operation. Dissolved chemicals, such as silica, titanium can precipitate on tube surface. The precipitate micro-fouling can deactivate or eliminate boiling nucleation sites. Therefore, the first phase of the post-cleaning operation suffered a decrease in steam pressure or an increase in fouling factor. It appears that micro fouling by magnetite deposit can activate or create more bubble nucleation sites. Therefore, the magnetite deposit micro-fouling results in a decrease in fouling factor, and a recovery in steam pressure. Fully understanding the boiling characteristics of the tubing at brand new, fouled and cleaned conditions requires further study of tubing surface conditions. Such study should include boiling heat transfer tests and scanning electronic microscope examination. (author)

Database for Pressure Tube Diameter and Operation Data of Wolsong NPP

International Nuclear Information System (INIS)

Jung, Jong Yeob; Kim, W. Y.; Bae, J. H.; Park, J. H.

2010-12-01

Pressure tube of CANDU reactor which is a long cylindrical shape of its diameter about 10 cm and length of about 6m, can be expanded toward both radial and axial directions due to irradiation under the high pressure and temperature condition. As the irradiation period increases, the radial expansion due to creep of the pressure tube increases. The radial expansion of the pressure tube comes out the reduction of the coolability and it results in the power deration. The objectives of the current work is to establish the database for the measured diametral data of pressure tube and operational data from Wolsong NPP as a preliminary work of developing the prediction model for pressure tube diameter. In order to develop the database, measured data for total 86 channels were collected from Wolsong NPP 1, 2, 3 and 4 and analyzed. Based on the provided data, the operational conditions such as an axial temperature and a pressure of the channel and neutron fluxes were derived. All data were analysed to derive the correlation between the pressure tube diameter and the other operational parameters

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

International Nuclear Information System (INIS)

Wolfer, W.G.; Reiley, T.C.

1977-05-01

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

The role of strain localization in the fracture of irradiated pressure tube material

International Nuclear Information System (INIS)

Dutton, R.

1989-04-01

This report reviews those phenomena that lead to strain localization in zirconium alloys, with particular reference to the role played by the formation of shear bands in fracture processes. The important influence of plastic deformation, in general, on fracture mechanisms is emphasized. This is to be expected when elastic-plastic fracture mechanics is the chosen analytical technique. Intensely inhomogeneous characteristics of strain localization cause an abrupt bifurcation in the evolution of deformation strain and lead to plastic instability linked with intrinsic material behaviour (e.g., work softening) or of geometric origin (e.g., localized necking). Both of these effects are discussed in relation to measurable deformation parameters, such as the work hardening rate and strain rate sensitivity, which determine the degree of resistance to plastic instability. The modifying effect of irradiation on these quantities is given specific attention, the appropriate literature pertaining to Zircaloy and Zr-2.5% Nb being reviewed. Recommendations are made for a combined experimental and theoretical program to characterize strain localization and reduced ductility in irradiated cold-worked Zr-2.5% Nb pressure tube material. The relationship between the deformation properties and the fracture behaviour is discussed

Tensile strength of Zr-2.5 Nb pressure tubes: A statistical study

Energy Technology Data Exchange (ETDEWEB)

Shah, Priti Kotak, E-mail: pritik@barc.gov.in [Senior Scientist, Bhabha Atomic Research Centre, Mumbai, 400 085 (India); Dubey, J.S.; Datta, D.; Shriwastaw, R.S.; Rath, B.N.; Singh, R.N. [Senior Scientist, Bhabha Atomic Research Centre, Mumbai, 400 085 (India); Anantharaman, S. [Head, Post Irradiation Examination Division, Bhabha Atomic Research Centre, Mumbai (India); Chakravartty, J.K. [Director, Materials Group, Bhabha Atomic Research Centre, Mumbai (India)

2015-12-15

Highlights: • Tensile properties in axial and transverse direction for a number of Indian Zr-2.5 Nb PHWR pressure tubes. • Distribution of tensile properties of double-melted and quadruple-melted pressure tubes. • Tensile properties at front-end and back-end of the quadruple melted pressure tubes at room temperature and at 300 °C. - Abstract: In order to get an idea of the statistical variation in the tensile properties of the double-melted as well as quadruple melted Zr-2.5 Nb pressure tubes (PTs) and also the variation in tensile properties between the two ends of the pressure tubes, tension tests were carried out on around 50 pressure tube off-cuts. Longitudinal and transverse tensile specimens were prepared from these off-cuts of pressure tubes of double-melted and quadruple melted types. For quadruple melted pressure tubes the specimens were tested from both front-end and back-end off-cuts. Miniature flat tensile specimens having 1.8 mm width and 1.5 mm thickness and 7.6 mm gauge length were prepared from the pressure tube off-cuts without any flattening treatment. Tension tests were carried out in a screw-driven machine at room temperature and 300 °C for both front-end and back-end off-cuts of each of 16 pressure tubes. In general the transverse specimens showed higher yield strength (YS) and ultimate tensile strength (UTS) compared to the longitudinal specimens. Transverse specimens showed less strain hardening compared to the longitudinal specimens. The axial specimens showed higher uniform (UE) and total elongation (TE) compared to the transverse specimens. Double-melted pressure tubes showed relatively higher strength and lower elongation and larger standard deviation compared to the quadruple melted pressure tubes. Mean values of tensile properties showed that back-end off-cuts were relatively stronger and less ductile compared to the front-end off-cuts.

Operating envelope to minimize probability of fractures in Zircaloy-2 pressure tubes

International Nuclear Information System (INIS)

Azer, N.; Wong, H.

1994-01-01

The failure mode of primary concern with Candu pressure tubes is fast fracture of a through-wall axial crack, resulting from delayed hydride crack growth. The application of operating envelopes is demonstrated to minimize the probability of fracture in Zircaloy-2 pressure tubes based on Zr-2.5%Nb pressure tube experience. The technical basis for the development of the operating envelopes is also summarized. The operating envelope represents an area on the pressure versus temperature diagram within which the reactor may be operated without undue concern for pressure tube fracture. The envelopes presented address both normal operating conditions and the condition where a pressure tube leak has been detected. The examples in this paper are prepared to illustrate the methodology, and are not intended to be directly applicable to the operation of any specific reactor. The application of operating envelopes to minimized the probability of fracture in 80 mm diameter Zircaloy-2 pressure tubes has been discussed. Both normal operating and leaking pressure tube conditions have been considered. 3 refs., 4 figs

TEM examination of irradiated zircaloy-2 pressure tube material

International Nuclear Information System (INIS)

Srivastava, D.; Tewari, R.; Dey, G.K.; Sharma, B.P.; Sah, D.N.; Banerjee, Suparna; Sahoo, K.C.

2005-09-01

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)

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

Energy Technology Data Exchange (ETDEWEB)

Gangotra, S; Muralidhar, S; Raut, S D; Ouseph, P M; Ghosh, J K; Sahoo, K C [Bhabha Atomic Research Centre, Bombay (India). Radiometallurgy Div.

1994-12-31

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.

Probabilistic fracture mechanics applied for DHC assessment in the cool-down transients for CANDU pressure tubes

Energy Technology Data Exchange (ETDEWEB)

Radu, Vasile, E-mail: vasile.radu@nuclear.ro [Institute for Nuclear Research Pitesti, 1st Campului Street, 115400 Mioveni, Arges, P.O. Box 78, Mioveni (Romania); Roth, Maria [Institute for Nuclear Research Pitesti, 1st Campului Street, 115400 Mioveni, Arges, P.O. Box 78, Mioveni (Romania)

2012-12-15

irradiation in reactor. The paper describes a prospective way for the probabilistic approach of CANDU pressure tube failure by DHC mechanisms during cool-down cycles by using probabilistic fracture mechanics principles. The limit state functions are defined for fracture instability and plastic collapse according to the Canadian Standard N285.8-05 criteria. British Procedure R6 is used to define another limit state function based on reserve factors. Discussion of lifetime probability values obtained from both procedures is made.

Evaluation of hydride blisters in zirconium pressure tube in CANDU reactor

International Nuclear Information System (INIS)

Cheong, Y. M.; Kim, Y. S.; Gong, U. S.; Kwon, S. C.; Kim, S. S.; Choo, K.N.

2000-09-01

When the garter springs for maintaining the gap between the pressure tube and the calandria tube are displaced in the CANDU reactor, the sagging of pressure tube results in a contact to the calandria tube. This causes a temperature difference between the inner and outer surface of the pressure tube. The hydride can be formed at the cold spot of outer surface and the volume expansion by hydride dormation causes the blistering in the zirconium alloys. An incident of pressure tube rupture due to the hydride blisters had happened in the Canadian CANDU reactor. This report describes the theoretical development and models on the formation and growth of hydride blister and some experimental results. The evaluation methodology and non-destructive testing for hydride blister in operating reactors are also described

Evaluation of hydride blisters in zirconium pressure tube in CANDU reactor

Energy Technology Data Exchange (ETDEWEB)

Cheong, Y M; Kim, Y S; Gong, U S; Kwon, S C; Kim, S S; Choo, K N

2000-09-01

When the garter springs for maintaining the gap between the pressure tube and the calandria tube are displaced in the CANDU reactor, the sagging of pressure tube results in a contact to the calandria tube. This causes a temperature difference between the inner and outer surface of the pressure tube. The hydride can be formed at the cold spot of outer surface and the volume expansion by hydride dormation causes the blistering in the zirconium alloys. An incident of pressure tube rupture due to the hydride blisters had happened in the Canadian CANDU reactor. This report describes the theoretical development and models on the formation and growth of hydride blister and some experimental results. The evaluation methodology and non-destructive testing for hydride blister in operating reactors are also described.

Influence of deformation rate on plasticity of metals under pressure

International Nuclear Information System (INIS)

Churbaev, R.V.; Dobromyslov, A.V.; Kolmogorov, V.L.; Taluts, G.G.

1990-01-01

Change of polycrystalline molybdenum (BCC) and titanium (HCP) plasticity under pressure depeding on the deformation rate at the room temperature is studied. It is shown that the reduction of molybdenum and titanium deformation rate leads to a substantial growth of their plastic properties with the effect being increased with pressure growth. Production of several necks testifying to the transition to a superplastic state is observed at high pressures and low deformation rates. A functional dependence of plasticity change on the deformation rate under pressure is ascertained

An improved model to predict nonuniform deformation of Zr-2.5 Nb pressure tubes

International Nuclear Information System (INIS)

Lei, Q.M.; Fan, H.Z.

1997-01-01

Present circular pressure-tube ballooning models in most fuel channel codes assume that the pressure tube remains circular during ballooning. This model provides adequate predictions of pressure-tube ballooning behaviour when the pressure tube (PT) and the calandria tube (CT) are concentric and when a small (<100 degrees C) top-to-bottom circumferential temperature gradient is present on the pressure tube. However, nonconcentric ballooning is expected to occur under certain postulated CANDU (CANada Deuterium Uranium) accident conditions. This circular geometry assumption prevents the model from accurately predicting nonuniform pressure-tube straining and local PT/CT contact when the pressure tube is subjected to a large circumferential temperature gradient and consequently deforms in a noncircular pattern. This paper describes an improved model that predicts noncircular pressure-tube deformation. Use of this model (once fully validated) will reduce uncertainties in the prediction of pressure-tube ballooning during a postulated loss-of-coolant accident (LOCA) in a CANDU reactor. The noncircular deformation model considers a ring or cross-section of a pressure tube with unit axial length to calculate deformation in the radial and circumferential directions. The model keeps track of the thinning of the pressure-tube wall as well as the shape deviation from a reference circle. Such deviation is expressed in a cosine Fourier series for the lateral symmetry case. The coefficients of the series for the first m terms are calculated by solving a set of algebraic equations at each time step. The model also takes into account the effects of pressure-tube sag or bow on ballooning, using an input value of the offset distance between the centre of the calandria tube and the initial centre of the pressure tube for determining the position radius of the pressure tube. One significant improvement realized in using the noncircular deformation model is a more accurate prediction in

Delayed hydrogen cracking test design for pressure tubes

International Nuclear Information System (INIS)

Haddad, Roberto; Loberse, Antonio N.; Yawny, Alejandro A.; Riquelme, Pablo

1999-01-01

CANDU nuclear power stations pressure tubes of alloy Zr-2,5 % Nb present a cracking phenomenon known as delayed hydrogen cracking (DHC). This is a brittle fracture of zirconium hydrides that are developed by hydrogen due to aqueous corrosion on the metal surface. This hydrogen diffuses to the crack tip where brittle zirconium hydrides develops and promotes the crack propagation. A direct current potential decay (DCPD) technique has been developed to measure crack propagation rates on compact test (CT) samples machined from a non irradiated pressure tube. Those test samples were hydrogen charged by cathodic polarization in an acid solution and then pre cracked in a fatigue machine. This technique proved to be useful to measure crack propagation rates with at least 1% accuracy for DHC in pressure tubes. (author)

Full length channel Pressure Tube sagging under completely voided full length pressure tube of an Indian PHWR

Energy Technology Data Exchange (ETDEWEB)

Negi, Sujay, E-mail: negi.sujay@gmail.com [Indian Institute of Technology, Roorkee 247667 (India); Kumar, Ravi, E-mail: ravikfme@gmail.com [Indian Institute of Technology, Roorkee 247667 (India); Majumdar, P., E-mail: pmajum@barc.gov.in [Bhabha Atomic Research Centre, Mumbai 400085 (India); Mukopadhyay, D., E-mail: dmukho@barc.gov.in [Bhabha Atomic Research Centre, Mumbai 400085 (India)

2017-03-15

Highlights: • At 16 kW/m input, thermal stability was attained at 595 °C, without PT-CT contact. • At 20 kW/m step input, PT-CT contact occurred at 637 °C near bottom-center of the tube. • PT integrity was maintained throughout the experiment. - Abstract: An experimental investigation was conducted to simulate the sagging behavior of a full length Pressure Tube of a channel of 220 MWe Indian PHWR. The investigation aimed to recreate a condition resembling Loss of Coolant Accident (LOCA) with Emergency Core Cooling System (ECCS) failure in a nuclear power plant. A full length channel assembly immersed in moderator was subjected to electrical resistance heating of Pressure Tube (PT) to simulate the residual heat after shutting down of reactor. The temperature of PT started rising and the contact between PT and CT was established at the center of the tube where average bottom temperature was 637 °C. The integrity of PT was maintained throughout the experiment and the PT heat up was arrested on contact with the CT due to transfer of heat to the moderator.

The cracking of pressure tubes in the Pickering reactor

International Nuclear Information System (INIS)

Ross-Ross, P.A.

1978-01-01

Small cracks in 17 of the 390 pressure tubes in Unit 3 of the 2056 MW (electrical) Pickering Generating Station and of 52 tubes in Unit 4, resulted in each of these units being out of service for many months. The cracks originated at areas of extremely high residual tensile stress produced by improper positioning of the rolling tool used during construction to join the pressure tube to its end-fitting. The mechanism of failure was delayed hydrogen cracking. (author)

Plastic collapse and energy absorption of circular filled tubes under quasi-static loads by computational analysis

Energy Technology Data Exchange (ETDEWEB)

Beng, Yeo Kiam; Tzeng, Woo Wen [Universiti Malaysia Sabah, Sabah (Malaysia)

2017-02-15

This study presents the finite element analysis of plastic collapse and energy absorption of polyurethane-filled aluminium circular tubes under quasi-static transverse loading. Increasing focuses were given to impact damage of structures where energy absorbed during impact could be controlled to avoid total structure collapse of energy absorbers and devices designed to dissipate energy. ABAQUS finite element analysis application was utilized for modelling and simulating the polyurethane-filled aluminium tubes, different set of diameterto- thickness ratios and span lengths, subjected to transverse three-point-bending load. Different sets of polyurethane-filled aluminium tubes subjected to the transverse loading were modelled and simulated. The failure modes and mechanisms of filled tubes and its capabilities as energy absorbers to further improve and strengthening of empty tube were also identified. The results showed that plastic deformation response was affected by the geometric constraints and parameters of the specimens. The diameter-to-thickness ratio and span lengths had shown to play crucial role in optimizing the PU-filled tube as energy absorber.

Plastic tube hadron calorimeter: study of operation properties and particle separation

International Nuclear Information System (INIS)

Akopdzhanov, G.A.; Belousov, V.I.; Blik, A.M.; Romanovski, V.I.

1988-01-01

The DELPHI hadron calorimeter prototype plastic tubes were tested to show a long-term stability of the prototype operating with the gas mixture carbon dioxide isobutane. The operating properties of the prototype are investigated and presented as well as the results on particles separation. 5 refs.; 11 figs.; 9 tabs

Pressure distribution over tube surfaces of tube bundle subjected to two phase cross flow

International Nuclear Information System (INIS)

Sim, Woo Gun

2013-01-01

Two phase vapor liquid flows exist in many shell and tube heat exchangers such as condensers, evaporators and nuclear steam generators. To understand the fluid dynamic forces acting on a structure subjected to a two phase flow, it is essential to obtain detailed information about the characteristics of a two phase flow. The characteristics of a two phase flow and the flow parameters were introduced, and then, an experiment was performed to evaluate the pressure loss in the tube bundles and the fluid dynamic force acting on the cylinder owing to the pressure distribution. A two phase flow was pre mixed at the entrance of the test section, and the experiments were undertaken using a normal triangular array of cylinders subjected to a two phase cross flow. The pressure loss along the flow direction in the tube bundles was measured to calculate the two phase friction multiplier, and the multiplier was compared with the analytical value. Furthermore, the circular distributions of the pressure on the cylinders were measured. Based on the distribution and the fundamental theory of two phase flow, the effects of the void fraction and mass flux per unit area on the pressure coefficient and the drag coefficient were evaluated. The drag coefficient was calculated by integrating the measured pressure coefficient and the drag coefficient were evaluated. The drag coefficient was calculated by integrating the measured pressure on the tube by a numerical method. It was found that for low mass fluxes, the measured two phase friction multipliers agree well with the analytical results, and good agreement for the effect of the void fraction on the drag coefficients, as calculated by the measured pressure distributions, is shown qualitatively, as compared to the existing experimental results

A statistical approach to the prediction of pressure tube fracture toughness

International Nuclear Information System (INIS)

Pandey, M.D.; Radford, D.D.

2008-01-01

The fracture toughness of the zirconium alloy (Zr-2.5Nb) is an important parameter in determining the flaw tolerance for operation of pressure tubes in a nuclear reactor. Fracture toughness data have been generated by performing rising pressure burst tests on sections of pressure tubes removed from operating reactors. The test data were used to generate a lower-bound fracture toughness curve, which is used in defining the operational limits of pressure tubes. The paper presents a comprehensive statistical analysis of burst test data and develops a multivariate statistical model to relate toughness with material chemistry, mechanical properties, and operational history. The proposed model can be useful in predicting fracture toughness of specific in-service pressure tubes, thereby minimizing conservatism associated with a generic lower-bound approach

Vapor pressure measured with inflatable plastic bag

Science.gov (United States)

1965-01-01

Deflated plastic bag in a vacuum chamber measures initial low vapor pressures of materials. The bag captures the test sample vapors and visual observation of the vapor-inflated bag under increasing external pressures yields pertinent data.

Development of Zr-2.5Nb pressure tubes for Advanced CANDU Reactor

International Nuclear Information System (INIS)

Bickel, G.A.; Griffiths, M.; Douchant, A.; Douglas, S.; Woo, O.T.; Buyers, A.

2010-01-01

In an Advanced CANDU Reactor (ACR), pressure tubes of cold-worked Zr-2.5Nb materials will be used in the reactor core to contain the fuel bundles and the light water coolant. They will be subjected to higher temperature, pressure and flux than that in a CANDU reactor. In order to ensure that these tubes will perform acceptably over their 30-year design life in such an environment, a manufacturing process has been developed to produce 6.5 mm thick ACR pressure tubes with optimized chemical composition, improved mechanical properties and in-reactor behaviour. The test and examination results show that, when compared with current in-service pressure tubes, the mechanical properties of ACR pressure tubes are significantly improved. Based on previous experience with CANDU reactor pressure tubes an assessment of the grain structure and texture indicates that the in-reactor creep deformation will be improved also. Analysis of the distribution of texture parameters from a trial batch of 26 tubes shows that the variability is reduced relative to tubes fabricated in the past. This reduction in variability together with a shift to a coarser grain structure will result in a reduction in diametral creep design limits and thus a longer economic life for the fuel channels of the advanced CANDU reactor. (author)

Use of CATHENA to model calandria-tube/moderator heat transfer after pressure-tube/calandria-tube ballooning contact

International Nuclear Information System (INIS)

Fan, H.Z.; Bilanovic, Z.; Nitheanandan, T.

2004-01-01

A study was performed to assess the effect of the calandria-tube/moderator heat transfer after pressure-tube/calandria tube ballooning contact using CATHENA. Results of this study indicated that the analytical tool, CATHENA, can be applied for pool boiling heat transfer on the external surface of a large diameter tube, such as the calandria tube used in CANDU reactors. The methodology in such CANDU-generic study can be used to simulate the tube surface with multiple boiling regimes and to assess the benefits of closely coupling thermalhydraulics modelling and fuel/fuel channel behaviour modelling. CATHENA (Canadian Algorithm for THErmalhydraulic Network Analysis) is a one-dimensional, two-fluid thermalhydraulic simulation code designed by AECL to analyse two-phase flow and heat transfer in piping networks. The detailed heat transfer package in CATHENA allows a connection to be established from the multiple solid surfaces of tubes to the surrounding large amount of moderator water, which acts as a heat sink during a postulated loss of coolant event. The generalized heat transfer package within CATHENA allows the tube walls to be divided into several layers in the radial direction and several sectors in the circumferential direction, to account for heat transfer conditions in these two directions. The CATHENA code with the generalized heat transfer package is capable of capturing key pool-boiling phenomena such as nucleate, transition and film boiling heat transfer as well as an ability to model the rewet phenomenon to some extent. A CATHENA input model was generated and used in simulations of selected contact boiling experiment test cases. The transient wall temperatures have been calculated in different portions of the calandria tube. By using this model an adequate agreement was achieved between CATHENA calculation and experimental measurement The CATHENA code enables one to investigate the transient and local thermal-mechanical behaviour of the calandria tube

Consequences of pressure tube rupture on in-core components

International Nuclear Information System (INIS)

Hill, P.G.; Hauptmann, E.G.; Lee, V.

1982-12-01

An investigation has been made of the consequences of pressure tube rupture in calandria vessels of heavy water cooled and moderated reactors. The study included a review of previous experimental and analytical work, as well as supplementary investigations carried out to examine the validity of previous assumptions and findings. The central questions considered were: the possibility of a propagating pressure tube failure; damage to the calandria vessel; and damage to the shut-off-rod guide tubes of the reactor shut-down system. The results of the investigation do not indicate mechanisms of sufficient strength to cause propagating failure in a well-designed, well-operated reactor following a tube burst under normal operating conditions. However, not all the details of the physical processes involved in a tube burst have been revealed by existing experimental and analytical work

Experiment on the effects of contact between the pressure tube and the fuel assembly

Energy Technology Data Exchange (ETDEWEB)

Watanabe, Y; Fujii, Y [Electric Power Development Co. Ltd., Tokyo (Japan); Kato, K [Hitachi Ltd., Ibaraki (Japan). Hitachi Works

1996-12-31

The Advanced Thermal Reactor (ATR) is a pressure tube type reactor in which the fuel assembly is located close to the pressure tube. The ATR has a structure which is such that the thermal stretch of the fuel pin is not limited by the spacer if the fuel pin dries out. Accordingly. it is not thought that the fuel pin contacts the pressure tube due to large transformations around the Design Based Event (DBE). Nevertheless, the safety margin must be kept in case the over-DBE. We have confirmed in this experiment that the temperature of the pressure tube does not increase to the critical level when the fuel pin contacts the pressure tube and the functions of the pressure tube are maintained as a pressure boundary. Further, we analyzed the safety margin of the pressure tube using the data from this experiment and from code analysis. (author). 10 tabs., 32 figs.

Highlights of the metallurgical behaviour of CANDU pressure tubes

International Nuclear Information System (INIS)

Price, E.G.

1984-10-01

This paper is an overview of the service induced metallurgical changes that take place in Zircaloy-2 and Zr-2.5 wt. percent Nb pressure tubes in CANDU reactors. It incorporates the findings of an evaluation program, that followed a significant pressure tube failure at Ontario Hydro's Pickering Nuclear Generating Station, and also provides valid reasons for continued confidence in the current CANDU design

Assessment and management of ageing of major nuclear power plant components important to safety: CANDU pressure tubes

International Nuclear Information System (INIS)

1998-08-01

The report documents the current practices for assessment and management of the ageing of the pressure tubes in CANDU reactors and Indian PHWTRs. Chapter headings are: fuel channel and pressure tube description, design basis for the fuel channel and pressure tube, degradation mechanisms and ageing concerns for pressure tubes, inspection and monitoring methods for pressure tubes,assessment methods and fitness-for-service guidelines for pressure tubes, mitigation methods for pressure tubes, and pressure tube ageing management programme

Anisotropic deformation of Zr–2.5Nb pressure tube material at high temperatures

Energy Technology Data Exchange (ETDEWEB)

Fong, R.W.L., E-mail: fongr@aecl.ca [Fuel and Fuel Channel Safety Branch, Atomic Energy of Canada Limited, Chalk River Nuclear Laboratories, Chalk River, Ontario (Canada)

2013-09-15

Zr–2.5Nb alloy is used for the pressure tubes in CANDU® reactor fuel channels. In reactor, the pressure tube normally operates at 300 °C and experiences a primary coolant fluid internal pressure of approximately 10 MPa. Manufacturing and processing procedures generate an anisotropic state in the pressure tube which makes the tube stronger in the hoop (transverse) direction than in the axial (longitudinal) direction. This anisotropy condition is present for temperatures less than 500 °C. During postulated accident conditions where the material temperature could reach 1000 °C, it might be assumed that the high temperature and subsequent phase change would reduce the inherent anisotropy, and thus affect the deformation behaviour (ballooning) of the pressure tube. From constant-load, rapid-temperature-ramp, uniaxial deformation tests, the deformation rate in the longitudinal direction of the tube behaves differently than the deformation rate in the transverse direction of the tube. This anisotropic mechanical behaviour appears to persist at temperatures up to 1000 °C. This paper presents the results of high-temperature deformation tests using longitudinal and transverse specimens taken from as-received Zr–2.5Nb pressure tubes. It is shown that the anisotropic deformation behaviour observed at high temperatures is largely due to the stable crystallographic texture of the α-Zr phase constituent in the material that was previously observed by neutron diffraction measurements during heating at temperatures up to 1050 °C. The deformation behaviour is also influenced by the phase transformation occurring at high temperatures during heating. The effects of texture and phase transformation on the anisotropic deformation of as-received Zr–2.5Nb pressure tube material are discussed in the context of the tube ballooning behaviour. Because of the high temperatures in postulated accident scenarios, any irradiation damage will be annealed from the pressure tube material

Modelling of oxidation and hydriding behaviour of Zircaloy-2 pressure tubes in PHWR

International Nuclear Information System (INIS)

Sah, D.N.; Sunil Kumar; Khan, K.B.

2002-01-01

A computer model named DOCTOR (Deuteriding of Coolant Tubes during Operation of Reactor) has been developed for predicting the axial profile of oxide thickness and hydrogen (Deuterium) concentration in PHWR pressure tubes. This model is applicable to single channel or full core analysis. The main source of hydrogen is considered to be oxidation of pressure tube on the i.d. surface by high temperature coolant water. Three stages of oxidation is considered namely, pre- transition, post transition and accelerated. Oxidation rate is considered to be dependent on channel power, axial power/flux distribution, coolant temperature and pre-existing oxide thickness at the location. The kinetics parameters for oxidation model are derived from the actual measurement of oxide thickness on a number of pressure tubes examined in PIE Division. The input data required for the model are: channel power, channel power factor, axial flux distribution, coolant inlet temperature, critical oxide thickness, hydrogen pick up fraction, initial hydrogen in the material and time of operation (efpy). The model calculates the oxide layer thickness on the inside surface of the pressure tube along the length. The amount of hydrogen picked up by the pressure tube is calculated from the oxide thickness using hydrogen pick up fraction determined from the PIE data. The pressure tube length is divided into a number of axial segments for calculation. The temperature and fast neutron flux assumed to be constant in a given segment. The axial temperature profile calculated from the axial power profile in the channel is used for calculating the oxidation rate at various locations in the pressure tube. The model has been validated with PIE data of hydrogen equivalent measurement on a number of irradiated Zircaloy-2 pressure tubes of various PHWRs. The performance of the model in predicting the axial profile of hydrogen in the pressure tubes has been found to be good. (author)

Development and performance of inspection equipment for pressure tubes in Fugen

International Nuclear Information System (INIS)

Naruo, Kazuteru; Tanimoto, Ken-ichi; Ohta, Takeo; Nakamura, Takahisa; Imaizumi, Kiyoshi.

1984-01-01

The pressure tubes of Fugen are the important equipment as the many tubes compose the core, and since they are made of Zr-2.5% Nb alloy which has been used for the first time in Japan, they have become the object of monitoring (the follow-up investigation of the change of inside diameter, the presence of defects and so on) in addition to the in-service inspection. In this paper, on the inspection equipment for pressure tubes, that has been developed independently by the Power Reactor and Nuclear Fuel Development Corp. in order to carry out the ISI and monitoring, the course of development and the construction and the performance are reported, and the results of having used it for the fourth regular inspection of Fugen are described. The 10-year plan of the ISI and monitoring of pressure tubes is shown. The core of Fugen is composed of 224 pressure tubes, therefore, the inspection is carried out by sampling inspection. The monitoring is carried out on four tubes for the follow-up investigation and one tube that shows the severest operation history at the time of inspection. The equipment performs ultrasonic flaw detection, the measurement of inside diameter and the visual inspection of internal surface. (Kako, I.)

Leak-before-break concept for evaluation of flows detected in pressure tubes in a Candu type reactor

International Nuclear Information System (INIS)

Crespi, J.C.

1992-01-01

This paper reviews the role of the Leak-Before-Break concept for evaluation of flaws detected in cold-worked Zr 2.5% Nb pressure tubes in a CANDU type reactors. The acceptance criteria are intended to prevent failure by brittle fracture, plastic collapse of the ligament and delayed hydride cracking. The methodology developed here was of great help in order to establish the operative conditions for fuel channel garter springs repositioning by means of the SLA Rette tool at Embalse Nuclear Generating Station, Cordoba, Argentina. (author)

Evaluation of Fatigue Crack Initiation for Volumetric Flaw in Pressure Tube

International Nuclear Information System (INIS)

Choi, Sung Nam; Yoo, Hyun Joo

2005-01-01

CAN/CSA.N285.4-94 requires the periodic inservice inspection and surveillance of pressure tubes in operating CANDU nuclear power reactors. If the inspection results reveal a flaw exceeding the acceptance criteria of the Code, the flaw must be evaluated to determine if the pressure is acceptable for continued service. Currently, the flaw evaluation methodology and acceptance criteria specified in CSA-N285.05-2005, 'Technical requirements for in-service evaluation of zirconium alloy pressure tubes in CANDU reactors'. The Code is applicable to zirconium alloy pressure tubes. The evaluation methodology for a crack-like flaw is similar to that of ASME B and PV Sec. XI, 'Inservice Inspection of Nuclear Power Plant Components'. However, the evaluation methodology for a blunt volumetric flaw is described in CSA-N285.05-2005 code. The object of this paper is to address the fatigue crack initiation evaluation for the blunt volumetric flaw as it applies to the pressure tube at Wolsong NPP

Flooding of a large, passive, pressure-tube light water reactor

International Nuclear Information System (INIS)

Hejzlar, P.; Todreas, N.E.; Driscoll, M.J.

1997-01-01

A reactor concept has been developed which can survive loss of coolant accidents without scram and without replenishing primary coolant inventory, while maintaining safe temperature limits on the fuel and pressure tubes. The proposed concept is a pressure tube type reactor of similar design to CANDU reactors, but differing in three key aspects. First, a solid SiC-coated graphite fuel matrix is used in place of fuel pin bundles to enable the dissipation of decay heat from the fuel in the absence of primary coolant. Second, the heavy water coolant in the pressure tubes is replaced by light water, which also serves as the moderator. Finally, the calandria tank, surrounded by a graphite reflector, contains a low pressure gas instead of heavy water moderator, and this normally-voided calandria is connected to a light water heat sink. The cover gas displaces the light water from the calandria during normal operation, while during loss of coolant or loss of heat sink accidents it allows passive calandria flooding. Calandria flooding also provides redundant and diverse reactor shutdown. This paper describes the thermal hydraulic characteristics of the passively initiated, gravity driven calandria flooding process. Flooding the calandria space with light water is a unique and very important feature of the proposed pressure-tube light water reactor (PTLWR) concept. The flooding of the top row of fuel channels must be accomplished fast enough so that in the total loss of coolant, none of the critical components of the fuel channel, i.e. the pressure tube, the calandria tube, the matrix and the fuel, exceed their design limits. The flooding process has been modeled and shown to be rapid enough to maintain all components within their design limits. (orig.)

Evaluation of the crack initiation of curved compact tension specimens of a Zr-2.5Nb pressure tube using the unloading compliance and direct current potential drop methods

International Nuclear Information System (INIS)

Jeong, Hyeon Cheol; Ahn, Sang Bok; Park, Joong Chul; Kim, Young Suk

2005-01-01

Zr-2.5Nb pressure tubes, carrying fuel bundles and heavy water coolant inside, degrade due to neutron irradiation and hydrogen embrittlement during their operation in heavy water reactors. The safety criterion for the Zr-2.5Nb tubes to meet is a leak-before-break (LBB) requirement. To evaluate a safety margin related to the LBB criterion, facture toughness of the pressure tubes are to be determined periodically with their operational time. For a reliable evaluation of the LBB safety criterion of the pressure tubes, it is required to precisely determine their fracture toughness. Since the fracture toughness or J of the pressure tubes is determined only by the extended crack length, it is important to reliably and precisely evaluate the advanced crack length. However, the problem lies with the detection of the crack opening point because prior plastic deformation before a start of the crack makes it difficult. The aim of this work is to evaluate which method can define the crack initiation point in the Zr- 2.5Nb compact tension specimens more precisely between the unloading compliance method with a crack opening displacement (COD) gauge and the direct current potential drop (DCPD) methods

Inelastic analysis of finite length and depth cracked tubes

International Nuclear Information System (INIS)

Reich, M.; Gardner, D.; Prachuktam, S.; Chang, T.Y.

1977-01-01

Steam generator tube failure can at times result in reactor safety problems and subsequent premature reactor shutdown. This paper concerns itself with the prediction of the failure pressures for typical PWR steam generator tubes with longitudinal finite length and finite depth cracks. Only local plastic overload failure is considered since the material is non-notch sensitive. Non-linear finite element analyses are carried out to determine the burst pressures of steam generator tubes containing longitudinal cracks located on the outer surface of the tubes. The non-linearities considered herein include elastic-plastic material behaviour and large deformations. A non-proprietary general purpose non-linear finite element program, NFAP was adopted for the analysis. Due to the asymmetric nature of the cracks, two-dimensional as well as three-dimensional finite element analyses, were performed. The analysis clearly shows that for short cracks axial effects play a significant role. For long cracks, they are not important since two-dimensional conditions predominate and failure is governed by circumferential or hoop stress conditions. (Auth.)

Influence of texture on the analysis of thermoelastic/plastic ansisotropy and the initial yielding of Zircaloy tubes

International Nuclear Information System (INIS)

Eisenberg, M.A.; Yen, C.F.

1980-01-01

Averaging procedures are developed to obtain texture factors relating the thermoelastic properties of single crystals of zirconium to polycrystalline properties. The effect of various approximations to the elastic properties of polycrystalline material and various assumed yield criteria on the transition from elastic to plastic behavior for a tube subjected to various boundary conditions is examined. It is found that thermal anisotropy has a greater effect on the elastic-plastic transitions than does elastic anisotropy. The choice of yield condition can also affect appreciably the predicted yielding behavior of a tube. (orig.)

Pressure heat pumping in the orifice pulse-tube refrigerator

International Nuclear Information System (INIS)

Boer, P.C.T. de

1996-01-01

The mechanism by which heat is pumped as a result of pressure changes in an orifice pulse-tube refrigerator (OPTR) is analyzed thermodynamically. The thermodynamic cycle considered consists of four steps: (1) the pressure is increased by a factor π 1 due to motion of a piston in the heat exchanger at the warm end of the regenerator; (2) the pressure is decreased by a factor π 2 due to leakage out of the orifice; (3) the pressure is further decreased due to motion of the piston back to its original position; (4) the pressure is increased to its value at the start of the cycle due to leakage through the orifice back into the pulse tube. The regenerator and the heat exchangers are taken to be perfect. The pressure is assumed to be uniform during the entire cycle. The temperature profiles of the gas in the pulse tube after each step are derived analytically. Knowledge of the temperature at which gas enters the cold heat exchanger during steps 3 and 4 provides the heat removed per cycle from this exchanger. Knowledge of the pressure as a function of piston position provides the work done per cycle by the piston. The pressure heat pumping mechanism considered is effective only in the presence of a regenerator. Detailed results are presented for the heat removed per cycle, for the coefficient of performance, and for the refrigeration efficiency as a function of the compression ratio π 1 and the expansion ratio π 2 . Results are also given for the influence on performance of the ratio of specific heats. The results obtained are compared with corresponding results for the basic pulse-tube refrigerator (BPTR) operating by surface heat pumping

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

, 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, combinations...... to continuous regulation of smaller pressures, whereas the tube was related to intermittent regulation of higher pressures....

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

DEFF Research Database (Denmark)

Gaihede, Michael; Jacobsen, Henrik; Tveterås, Kjell

, 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, combinations...... to continuous regulation of smaller pressures, whereas the tube was related to intermittent regulation of higher pressures....

The examination of the ruptured Zircaloy-2 pressure tube from Pickering NGS Unit 2

International Nuclear Information System (INIS)

Cheadle, B.A.; Smith, A.D.; Baskin, C.C.

1985-07-01

On 1983 August 01 a Zircaloy-2 pressure tube in Pickering NGS Unit 2 ruptured. All the fuel channel components, the fuel bundles, pressure tube, end fittings, garter springs and calandria tubes were shipped to Chalk River Nuclear Laboratories for examination to determine the cause of the rupture. The examination showed that the rupture initiated at a series of hydride blisters on the outside surface of the pressure tube. The blisters formed because of the garter spring spacers between the pressure tube and calandria tube was about one metre out of position. This allowed the horizontal pressure tube to sag by creep and touch the cool calandria tube. The resulting thermal gradients in the pressure tube concentrated the hydrogen and deuterium at the cool zones and blisters of solid hydride formed. Cracks initiated at several of the blisters and linked together to form a partial through wall critical crack which initiated the final rupture. The video presentation shows how the examination of the fuel channel components was conducted in underwater bays and shielded cells and explains the sequence of events that caused the rupture

All-plastic fiber-based pressure sensor

DEFF Research Database (Denmark)

Bundalo, Ivan-Lazar; Lwin, Richard; Leon-Saval, Sergio

2016-01-01

We present a feasibility study and a prototype of an all-plastic fiber-based pressure sensor. The sensor is based on long period gratings inscribed for the first time to the best of our knowledge by a CO2 laser in polymethyl methacrylate (PMMA) microstructured fibers and coupled to a pod......-like transducer that converts pressure to strain. The sensor prototype was characterized for pressures up to 150 mbars, and various parameters related to its construction were also characterized in order to enhance sensitivity. We consider this sensor in the context of future applications in endoscopic pressure...... sensors....

On random pressure pulses in the turbine draft tube

Science.gov (United States)

Kuibin, P. A.; Shtork, S. I.; Skripkin, S. G.; Tsoy, M. A.

2017-04-01

The flow in the conical part of the hydroturbine draft tube undergoes various instabilities due to deceleration and flow swirling at off-design operation points. In particular, the precessing vortex rope develops at part-load regimes in the draft tube. This rope induces periodical low-frequency pressure oscillations in the draft tube. Interaction of rotational (asynchronous) mode of disturbances with the elbow can bring to strong oscillations in the whole hydrodynamical system. Recent researches on flow structure in the discharge cone in a regime of free runner had revealed that helical-like vortex rope can be unstable itself. Some coils of helix close to each other and reconnection appears with generation of a vortex ring. The vortex ring moves toward the draft tube wall and downstream. The present research is focused on interaction of vortex ring with wall and generation of pressure pulses.

Microstructural examination of irradiated zircaloy-2 pressure tube material

International Nuclear Information System (INIS)

Srivastava, D.; Tewari, R.; Dey, G.K.; Sah, D.N.; Banerjee, S.

2005-01-01

Irradiation induced microstructural changes in Zr alloys strongly influence the creep, growth and mechanical properties of pressure tube material. Since dimensional changes and mechanical property degradation can limit the life of pressure tube, it is essential to study and develop an understanding of the microstructure produced by neutron irradiation, by examining samples taken from the irradiated components. In the present work, an effort has been made to examine, microstructure of the Zircaloy-2 pressure tube material irradiated in the Indian Pressurized Heavy Water Reactor (PHWR). The present work is a first step towards a comprehensive program of characterization of microstructure of reactor materials after irradiation to different fluence levels in power reactors. In this study, samples 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), have been prepared and examined. The samples selected from the tube are expected to have a cumulative radiation damage of about 3 dpa. Samples prepared from the off cuts of RAPS-1 pressure tubes were also studied for examining the unirradiated microstructure of the material. The samples were examined in a 200kV JEOL 2000 FX microscope. This paper presents the distinct features observed in irradiated sample and a comprehensive comparison of the microstructures of the unirradiated and irradiated material. The effect of annealing on the annihilation of the defects generated during irradiation has been also studied. The bright field micrographs revealed that microstructure of the irradiated samples was different in many respects from the microstructure of the unirradiated samples. The presence of defect structure in the form of loops etc could be seen in the irradiated sample. These loops were mostly c-type loops lying in the basal plane. The dissolution and redistribution of the precipitates were

Feedwater heater tube-to-tubesheet connections

International Nuclear Information System (INIS)

Yokell, S.

1993-01-01

This paper discusses some practical aspects of expanded, welded, and welded-and-expanded feedwater heater tube-to-tubesheet joints. It outlines elastic-plastic tube expanding theory. It examines uniform-pressure-expanded tube joint strength and correlating roller-expanded joint strength with wall reduction and rolling torque. For materials subject to stress-corrosion cracking (SCC), it recommends heat treating tube ends before expanding. For materials subject to fatigue and tube-end cracking, it advocates two-stage expanding: (1) expanding enough to create firm tube-hole contact over the full tubesheet thickness; and (2) re-expanding at full pressure or torque. The paper emphasizes the desirability of segregating heats of tubing, mapping the tube-heat locations and making the heat map a permanent part of the heater maintenance file. It recommends when to provide TEMA/HEI Power Plant Standard annular grooves for roller-expanding and provides an equation for determining optimum groove width for uniform-pressure expanding. The paper also reviews welding requirements for welds of tubes to tubesheets. The review covers front-face welding before and after expanding and the reasons for welding first. It outlines current thinking about definitions of strength- and seal-welds of front-face welded joint in terms of their functions and load-carrying abilities. It presents a proposal for determining the required size of strength welds for use in Section VIII of the ASME Boiler and Pressure Vessel Code (the Code). It shows why welded-and-expanded feedwater heater tube-to-tubesheet joints should be full-strength and full-depth expanded. It makes recommendations for pressure- and leak-testing. This work also proposes the industry consider butt welding the tubes to the steam-side face of the tubesheet as a regular method of tube joining. The results of a survey of manufacturers practices are appended. 30 refs., 14 figs

Measurement and analysis of pressure tube elongation in the Douglas Point reactor

International Nuclear Information System (INIS)

Causey, A.R.; MacEwan, S.R.; Jamieson, H.C.; Mitchell, A.B.

1980-02-01

Elongations of zirconium alloy pressure tubes in CANDU reactors, which occur as a result of neutron-irradiation-induced creep and growth, have been measured over the past 6 years, and the consequences of thses elongations have recently been analysed. Elongation rates, previously deduced from extensive measurements of elongations of cold-worked Zircaloy-2 pressure tubes in the Pickering reactors, have been modified to apply to the pressure tubes in the Douglas Point (DP) reactor by taking into account measured diffences in texture and dislocation density. Using these elongation rates, and structural data unique to the DP reactor, the analysis predicts elongation behaviour which is in good agreement with pressure tube elongations measured during the ten years of reactor operation. (Auth)

49 CFR 192.619 - Maximum allowable operating pressure: Steel or plastic pipelines.

Science.gov (United States)

2010-10-01

... plastic pipelines. 192.619 Section 192.619 Transportation Other Regulations Relating to Transportation... Operations § 192.619 Maximum allowable operating pressure: Steel or plastic pipelines. (a) No person may operate a segment of steel or plastic pipeline at a pressure that exceeds a maximum allowable operating...

Design of experiments and equipment to test the ballooning characteristics of CANDU pressure tubes

International Nuclear Information System (INIS)

Forrest, C.F.; Stern, F.; Hart, R.G.

1992-01-01

Experiments have been planned and an apparatus has been designed to enable creep testing of end-of-life pressure tube specimens in a LOCA environment. Effects that could be studied include: annealing of irradiation damage during transient heating; effects of hydride blisters on pressure tube ballooning strains; and, effects of uniformly-distributed hydrogen content on pressure tube ballooning strains. The proposed experimental program will consist of separate effects creep tests on pressure tube sections under transient heating conditions

Tube Plugging Criteria for the High-pressure Heaters of Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyungnam; Cho, Nam-Cheoul; Lee, Kuk-hee [KHNP Central Research Institute, Daejeon (Korea, Republic of)

2015-10-15

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 a nuclear power plant. This method relies on the similar plugging criteria used in the steam generator tubes. 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. 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

Guarantee of reliability of devices complexes for plastic tube welding

International Nuclear Information System (INIS)

Voskresenskij, L.A.; Zajtsev, A.I.; Nelyubov, V.I.; Fedorov, M.A.

1988-01-01

Results of calculations and experimental studies on providing reliability of complex for plastic tube welding are presented. Choice of reliability indeces and standards is based. Reliability levels of components are determined. The most waded details are calculated. It is shown that they meet the reqrurements of strength and reliability. Service life tests supported the correct choice of springs. Recommendations on elevating reliability are given. Directions of further developments are shown. 8 refs.; 2 figs.; 1 tab

CT findings of foreign body reaction to retained endoloop ligature plastic tube mimicking acute appendicitis: A case report

Energy Technology Data Exchange (ETDEWEB)

Ahn, Jae Hong; Kang, Chae Hoon; Choi, Soo Jung; Park, Man Soo; Jung, Seung Mun; Ryu, Dae Shick; Shin, Dong Rock [Dept. of Radiology, GangNeung Asan Hospital, University of Ulsan College of Medicine, Gangneung (Korea, Republic of)

2016-07-15

Many hospitals experience one or more retained surgical instrument events with risk of patient morbidity and medicolegal problems. Identification of retained surgical instrument is important. The radiologists should be familiar with imaging finding of retained surgical instrument. In a 62-year-old female with a retained plastic tube, localized peritoneal infiltration around air-containing tubular structure mimicked acute appendicitis on abdomen computed tomography (CT), one year after laparoscopic cholecystectomy. We reported CT findings of foreign body reaction related to retained Endoloop ligature plastic tube mimicking acute appendicitis.

CT findings of foreign body reaction to retained endoloop ligature plastic tube mimicking acute appendicitis: A case report

International Nuclear Information System (INIS)

Ahn, Jae Hong; Kang, Chae Hoon; Choi, Soo Jung; Park, Man Soo; Jung, Seung Mun; Ryu, Dae Shick; Shin, Dong Rock

2016-01-01

Many hospitals experience one or more retained surgical instrument events with risk of patient morbidity and medicolegal problems. Identification of retained surgical instrument is important. The radiologists should be familiar with imaging finding of retained surgical instrument. In a 62-year-old female with a retained plastic tube, localized peritoneal infiltration around air-containing tubular structure mimicked acute appendicitis on abdomen computed tomography (CT), one year after laparoscopic cholecystectomy. We reported CT findings of foreign body reaction related to retained Endoloop ligature plastic tube mimicking acute appendicitis

Method of detecting leakage from sealing attached to pressure tube

International Nuclear Information System (INIS)

Tomomatsu, Ken-ichi; Hayashi, Ken-ichi.

1990-01-01

The present invention provides a detection method for measuring the amount of water leaked from sealings attached to the lower end of a pressure tube. That is, the lower end of the pressure tube is sealed only by a metal sealing. A capturing vessel is placed under the pressure tube for capturing the leaked water dropping from the lower end of the pressure tube and the weight of the leaked water is measured on every capturing vessels to determine the amount of the leaked water. The leakage detection method based on the weight measurement has higher accuracy compared with a conventional volume measuring method using a water level gauge as described below. For example, if the volume of the captured water is 10cc, an error of about 0.1cc is caused by the volume measuring method using the water level gauge, whereas if 10g (10cc) weight of water is measured by using an accurate balance, error is only about 10 -4 g (10 -4 cc). Accordingly, the method of the present invention can measure at an accuracy about 1000 times as high as the conventional method. (I.S.)

CFD analysis of multiphase coolant flow through fuel rod bundles in advanced pressure tube nuclear reactors

International Nuclear Information System (INIS)

Catana, A.; Turcu, I.; Prisecaru, I.; Dupleac, D.; Danila, N.

2010-01-01

The key component of a pressure tube nuclear reactor core is pressure tube filled with a stream of fuel bundles. This feature makes them suitable for CFD thermal-hydraulic analysis. A methodology for CFD analysis applied to pressure tube nuclear reactors is presented in this paper, which is focused on advanced pressure tube nuclear reactors. The complex flow conditions inside pressure tube are analysed by using the Eulerian multiphase model implemented in FLUENT CFD computer code. Fuel rods in these channels are superheated but the liquid is under high pressure, so it is sub-cooled in normal operating conditions on most of pressure tube length. In the second half of pressure tube length, the onset of boiling occurs, so the flow consists of a gas liquid mixture, with the volume of gas increasing along the length of the channel in the direction of the flow. Limited computer resources enforced us to use CFD analysis for segments of pressure tube. Significant local geometries (junctions, spacers) were simulated. Main results of this work are: prediction of main thermal-hydraulic parameters along pressure tube including CHF evaluation through fuel assemblies. (authors)

Performance evaluation of reactor operated zircaloy-2 pressure tubes of RAPS-1

International Nuclear Information System (INIS)

Chatterjee, S.; Ramadasan, E.; Balakrishnan, K.S.; Bahl, J.K.

1992-01-01

Detailed post irradiation examination was carried out on pressure tube sections from E-10, F-9 and F-10 locations of RAPS-1 after an in-reactor residence equivalent to 3.6 effective full power years. The F-10 pressure tube was studied in detail on sections obtained from one end to the other, whereas in the case of E-9 and F-9 pressure tubes only the end sections were examined. The studies carried out were visual examination, metallography, hydrogen i.e. H(D) analysis and mechanical testing at 300 C. Microstructural observations revealed uniform and random hydride/deuteride platelet distribution and absence of blisters or hydride segregation. The H(D) content in the F-10 pressure tube was found to vary in the range 6-12 ppm. The typical H(D) content in the three tubes was around 1 ppm. The H(D) pick-up evaluated from the observed oxide layer thickness was 8 ppm. Longitudinal tensile specimens fabricated from the F-10 pressure tube section and tested at 300 C exhibited increase in yield strength and tensile strength of 39% and 30% respectively. The residual uniform elongation was typically 1.8%. The observed changes in the tensile properties were found to be lower than those reported on unstressed specimens irradiated to similar neutron fluences. The observed hydrogen content and tensile properties obtained in F-10 pressure tube would not be detrimental under normal reactor operating conditions. (author). 10 refs., 4 figs., 2 tabs., 1 annexure

First Research Coordination Meeting on Prediction of Axial and Radial Creep in HWR Pressure Tubes. Presentations

International Nuclear Information System (INIS)

2013-01-01

Pressure tube deformation is a critical aging issue in operating Heavy Water Reactors (HWRs). According to the service year, horizontal pressure tubes have three kinds of deformation: diametral creep leading to the flow bypass and the penalty to critical heat flux for fuel rods, longitudinal creep leading to the interference of feeder pipes and/or with fuelling machine, and sagging leading to the interference with in-core components and potential contact between the pressure tube and calandria tube. The CRP scope includes the establishment of a database for pressure tube deformation, microstructure characterization of pressure tube materials collected from HWRs currently operating in Member States and development of a prediction model for pressure tube deformation

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

International Nuclear Information System (INIS)

Ardhapurkar, P. M.; Sridharan, Arunkumar; Atrey, M. D.

2014-01-01

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

Analysis of plastic residues in maple sap and syrup collected from tubing systems sanitized with isopropyl alcohol

Directory of Open Access Journals (Sweden)

Luc Lagacé

2017-05-01

Full Text Available A plastic tubing system operated under vacuum is usually used to collect sap from maple trees during spring time to produce maple syrup. This system is commonly sanitized with isopropyl alcohol (IPA to remove microbial contamination colonizing the system during the sugar season. Questions have been raised whether IPA would contribute to the leaching of plastic residues in maple sap and syrup coming from sanitized systems. First, an extraction experiment was performed in the lab on commercial plastic tubing materials that were submitted to IPA under harsh conditions. The results of the GC-MS analysis revealed the presence of many compounds that served has target for further tests. Secondly, tests were done on early and mid-season maple sap and syrup coming from many sugarbushes using IPA or not to determine potential concentrations of plastic residues. Results obtained from sap and syrup samples showed that no quantifiable (< 1–75 μg/L concentration of any plastic molecules tested was determined in all samples coming from IPA treated or not treated systems. However, some samples of first sap run used as a rinse solution to be discarded before the season start and that were coming from non sanitized or IPA sanitized systems, showed quantifiable concentrations of chemical residue such as ultraviolet protector (octabenzone. These results show that IPA can be safely used to sanitize maple sap collection system in regards to the leaching of plastic residues in maple sap and syrup and reinforced the need to thoroughly rinse the tubing system at the beginning of the season for both sanitized and non sanitized systems. Keywords: Food science, Food safety, Materials chemistry

Status and Plans for work on pressure tube creep at AECL

International Nuclear Information System (INIS)

Bickel, Grant A.

2013-01-01

AECL research goals: • Develop empirical models to: – regress out operating conditions/extrinsic factors – rank relative strain behavior of measured in-service pressure tubes; • Correlate the ranked strains to manufacturing variables and the microstructure to: – Develop mechanistic insights – Optimize manufacturing/microstructure for improved pressure tube performance

Gamma sensitivity of pressurized drift tubes

International Nuclear Information System (INIS)

Baranov, S.A.; Bojko, I.R.; Shelkov, G.A.; Ignatenko, M.A.

1995-01-01

Using a set of commonly used radioactive sources, the efficiency of pressurized drift tubes for gammas with energy from 5.9 keV up to 1.3 MeV has been measured. The tube was made of aluminium and filled with Ar, 15%CO 2 and 2.5%iC 4 H 10 gas mixture at 3 atm. The measured efficiency is compared with the results of the calculations in the frame of our simple model as well as with that of the Monte Carlo simulation using GEANT code. The results of our calculations are in agreement with experimental data, while GEANT simulation tends to give lower efficiency in the energy range of 200 keV γ <1300 keV. The average efficiency of the tube in the field of ATLAS gamma background is about 0.45%. 8 refs., 7 figs., 1 tab

Refrigerant charge, pressure drop, and condensation heat transfer in flattened tubes

Energy Technology Data Exchange (ETDEWEB)

Wilson, M J; Newell, T A; Chato, J C [University of Illinois, Urbana, IL (United States). Dept. of Mechanical and Industrial Engineering; Infante Ferreira, C A [Delft University of Technology (Netherlands). Laboratory for Refrigeration and Indoor Climate Control

2003-06-01

Horizontal smooth and microfinned copper tubes with an approximate diameter of 9 mm were successively flattened in order to determine changes in flow field characteristics as a round tube is altered into a flattened tube profile. Refrigerants R134a and R410A were investigated over a mass flux range from 75 to 400 kg m{sup -2} s{sup -}2{sup 1} and a quality range from approximately 10-80%. For a given refrigerant mass flow rate, the results show that a significant reduction in refrigerant charge is possible. Pressure drop results show increases of pressure drop at a given mass flux and quality as a tube profile is flattened. Heat transfer results indicate enhancement of the condensation heat transfer coefficient as a tube is flattened. Flattened tubes with an 18{sup o} helix angle displayed the highest heat transfer coefficients. Smooth tubes and axial microfin tubes displayed similar levels of heat transfer enhancement. Heat transfer enhancement is dependent on the mass flux, quality and tube profile. (author)

Delayed hydride cracking in irradiated Zr-2.5 % Nb pressure tubes

International Nuclear Information System (INIS)

Cirimello, Pablo; Coronel, Pascual; Haddad, Roberto; Lafont, Claudio; Mizrahi, Rafael

2003-01-01

Pressure tubes in CANDU nuclear power plants are made of Zr-2.5 % Nb alloy, which is susceptible to a cracking process called Delayed Hydride Cracking (DHC). Measurement of DHC velocity on irradiated pressure tubes is essential to assure the validity of the Leak Before Break criterion. This work was performed on samples from two pressure tubes taken out of the Embalse NPP in 1995, belonging to fuel channels A-14 and L-12. DHC velocity in the axial direction was measured at 211 C degrees for samples taken from different axial positions, which allowed to study its dependence on fast neutron fluency and irradiation temperature. Non-irradiated material was also tested. It was found that DHC velocity results for the tested material were similar to those obtained for a great number of tubes irradiated in other CANDU plants. (author)

Pressure relief experiments on a cyclindrical carbon brick tube

International Nuclear Information System (INIS)

Lang, H.; Weise, H.J.; Ennen, P.

1978-08-01

Pressure relief experiments have been carried out on a carbon brick tube. The outer diameter of the specimen was 580 mm, the inner diameter 280 mm, the length 800 mm. The experiments were made with helium at the temperature of the environment. The measurements were carried out in the pressure range from 15 upto 39 bar. The pressure loss was measured dependent on the initial pressure and on time at 5 positions uniformly distributed over the thickness of the tube wall and in the pressure vessel. The maximum pressure transients occurred amounted to approximately 60 bar/second. The maximum overpressure with respect to the environment which occurred in the carbon brick during the relief experiments was about 3.3 bar. The measurements distinctly showed the presence and the effects of inhomogeneities in the sample material. No damages or changes in the carbon brick, which could be regarded as a consequence of the experiments, were found. (orig./GSC) [de

Characterization of the surface physico-chemistry of plasticized PVC used in blood bag and infusion tubing.

Science.gov (United States)

Al Salloum, H; Saunier, J; Dazzi, A; Vigneron, J; Etcheberry, A; Marlière, C; Aymes-Chodur, C; Herry, J M; Bernard, M; Jubeli, E; Yagoubi, N

2017-06-01

Commercial infusion tubing and blood storage devices (tubing, blood and platelets bags) made of plasticized PVC were analyzed by spectroscopic, chromatographic and microscopic techniques in order to identify and quantify the additives added to the polymer (lubricants, thermal stabilizers, plasticizers) and to put into evidence their blooming onto the surface of the devices. For all the samples, deposits were observed on the surface but with different kinds of morphologies. Ethylene bis amide lubricant and metallic stearate stabilizers were implicated in the formation of these layers. In contact with aqueous media, these insoluble deposits were damaged, suggesting a possible particulate contamination of the infused solutions. Copyright © 2017 Elsevier B.V. All rights reserved.

[Characterizing the passive opening of the eustachian tube in a hypo-/hyperbaric pressure chamber].

Science.gov (United States)

Meyer, M F; Mikolajczak, S; Luers, J C; Lotfipour, S; Beutner, D; Jumah, M D

2013-09-01

Beside arbitrary and not arbitrary active pressure equalization systems there is a passive equalization system via the Eustachian tube (ET) at pressure difference between the epipharyngeal space and the middle ear. Aim of this study was to characterize this passive equalization system in a hypobaric/hyperbaric pressure chamber by continuously measuring the tympanic impedance. In contrast to other studies, which are measured only in a hypobaric pressure chamber it is possible to include participants with Eustachian tube dysfunction (ETD). Following a fixed pressure profile 39 participants were exposed to phases of pressure rising and decompression. By continuously measuring the tympanic impedance in the pressure chamber it was possible to measure data of the Eustachian Tube opening Pressure (ETOP), Eustachian Tube closing pressure (ETCP) and Eustachian Tube opening duration (ETOD). In addition it was possible to characterize the gradient of pressure during decompression, while the ET was open. Beside the measurement of the arithmetic average of the ETOP (30.2 ± 15.1 mbar), ETCP (9.1 ± 7.7 mbar) and ETOD (0.65 ± 0.38 s) it was obvious that there are recurrent samples of pressure progression during the phase of tube opening. Generally it is possible to differentiate between the type of complete opening and partial opening. The fundamental characterization of the action of the passive tube opening, including the measurement of the ETOP, ETCP and ETOD, is a first step in understanding the physiological and pathophysiological function of the ET. © Georg Thieme Verlag KG Stuttgart · New York.

Delayed hydride cracking in Zr-2.5Nb pressure tubes

International Nuclear Information System (INIS)

Mieza, Juan I.; Domizzi, Gladys; Vigna, Gustavo L.

2007-01-01

Zr-2.5 Nb alloy from CANDU pressure tubes are prone to failure by hydrogen intake. One of the degradation mechanisms is delayed hydride cracking, which is characterized by the velocity of cracking. In this work, we study the effect of beta zirconium phase transformation over delayed hydride cracking velocity in Zr-2.5 Nb alloy from pressure tubes. Acoustic emission technique was used for cracking detection. (author) [es

An ice rink refrigeration system based on CO{sub 2} as secondary fluid in copper tubes

Energy Technology Data Exchange (ETDEWEB)

Shahzad, Khuram [Royal Inst. of Technology, Stockholrn (Sweden). Dept. of Energy Technology

2006-06-15

This report is a study of the use of copper tubes with CO{sub 2} as heat transfer fluid in ice rink applications. Copper tubes can be rolled rather easily up to the required length which decreases installation cost and simplifies the procedure. A test ice rink was built at IUC Ref Centre, Katrineholm with copper tubes. FEMLAB and EES are two softwares that were used for analysis. The comparison between 12.7 mm diameter copper tubes with and without plastic foil cover, 9.5 mm diameter copper tubes with and without plastic foil cover, 21.3 mm diameter steel pipes and 25 mm diameter plastic pipes is presented in the report. The reason to have plastic foil over copper tubes is to avoid the minor risk of chemical corrosion. Furthermore the foil serves as mechanical wear protection as well, which in this case could appear if rubbing would occur due to thermal expansion and contraction. It is found that 12.7 mm copper tube with plastic foil is good choice in terms of heat transfer. At rated heat flux of 100 W/m{sup 2} and with a pitch of 100 mm, it is 0.18 deg C better than 9.5 mm copper tube with plastic foil. This report includes the investigation which shows that there is no danger of movement of copper tubes inside the rink bed due to thermal expansion and contraction during operation. It also includes the comparison of average Friedel pressure drop model and average homogeneous pressure drop model with experimental results. Average Friedel pressure drop method gave good results. It predicted 20 to 25 % higher pressure drop at lower CR and about 60 % at higher CR than the experimental results for 120 meter long and 12.7 mm diameter copper tubes. 120 meter long copper tubes are good choice; as header can be placed on short side of the ice rink. It will reduce the header length and connections to half. FEMLAB modelling for conduction heat transfer gave good results and can be used as a tool for design and optimization. The optimization of the pitch of the copper tubes

Pressure tube replacement in Pickering NGS A units 1 and 2

International Nuclear Information System (INIS)

Irvine, H.S.; Bennett, E.J.; Talbot, K.H.

1986-10-01

Being able to technically and economically replace the most radioactive components (excluding the nuclear fuel) in operating reactors will help to ensure the ongoing acceptance of nuclear power as a viable energy source for the future. Ontario Hydro is well along the path to meeting the above objective for its CANDU-PHW reactors. Following the failure of a Zircaloy-II pressure tube in unit 2 of Pickering NGS A in August, 1983, Ontario Hydro has embarked on a program to replace all Zircaloy-II pressure tubes in units 1 and 2 at Pickering. This program integrates the in-house research, design, construction, and operating skills of a large utility (Ontario Hydro) with the skills of a national nuclear organization (Atomic Energy of Canada Limited) and the private engineering sector of the Canadian nuclear industry. The paper describes the background to the pressure tube failure in Pickering unit 2 and to the efforts incurred in understanding the failure mechanism and how similar failures are not expected for the zirconium-niobium pressure tube material used in all other large CANDU-PHW units after units 1 and 2 of Pickering NGS A. The tooling developed for the pressure tube replacement program is described as well as the organization to undertake the program in an operating nuclear station. The retubing of units 1 and 2 at Pickering NGS A is nearing a successful completion and shows the benefits of being able to integrate the various skills required for this success. Pressure tube replacement in a CANDU-PHW reactor is equivalent to replacement of the reactor vessel in a LWR. The fact that this replacement can be done economically and with acceptable radiation dose to workers augurs well for the continued viability of the use of nuclear energy for the benefit of mankind. (author)

Decontamination and recycle of zirconium pressure tubes from Pressurized Heavy Water Reactor

International Nuclear Information System (INIS)

Gantayet, L.M.; Verma, R.; Remya Devi, P.S.; Banerjee, S.; Kotak, V.; Raha, A.; Sandeep, K.C.; Joshi, Shreeram W.; Lali, A.M.

2009-01-01

An ion exchange process has been developed for decontamination of zirconium pressure tubes from Pressurized Heavy Water Reactor and recycling of neutronically improved zirconium. Distribution coefficient, equilibrium isotherm, kinetic and breakthrough data were used to develop the separation process. Effect of gamma radiation on indigenous resins was also studied to assess their suitability in high radiation field. (author)

Ring shaped laser for tape winding of an endless tube

NARCIS (Netherlands)

Leonardus, Lucky

2012-01-01

AFPT is a start-up company that provides laser-assisted technology for the production of a pressurized component such as tube, pressure vessel, etc made from or strengthened by fiber reinforced plastic (FRP) tape. The tape is laid precisely by a machine head (connected to a robot), melted with a

Flooding of a large, passive, pressure-tube LWR

Energy Technology Data Exchange (ETDEWEB)

Hejzlar, P.; Todreas, N.E.; Driscoll, M.J. [Massachusetts Institute of Technology, Cambridge, MA (United States)

1995-09-01

A reactor concept has been developed which can survive LOCA without scram and without replenishing primary coolant inventory. The proposed concept is a pressure tube type reactor similar to CANDU reactors, but differing in three key aspects: (1) a solid SiC-coated graphite fuel matrix is used in place of fuel pin bundles, (2) the heavy water coolant in the pressure tubes is replaced by light water, and (3) the calandria tank contains a low pressure gas instead of heavy water moderator. The gas displaces the light water from the calandria during normal operation, while during loss of coolant or loss of heat sink accidents, it allows passive calandria flooding. This paper describes the thermal hydraulic characteristics of the gravity driven calandria flooding process. Flooding the calandria space with light water is a unique and very important feature of the proposed pressure-tube LWR concept. The flooding of the top row of fuel channels must be accomplished fast enough so that none of the critical components of the fuel channel exceed their design limits. The flooding process has been modeled and shown to be rapid enough to maintain all components within their design limits. Two other considerations are important. The thermal shock experienced by the calandria and pressure tubes has been evaluated and shown to be within acceptable bounds. Finally, although complete flooding renders the reactor deeply subcritical, various steam/water densities can be hypothesized to be present during the flooding process which could cause reactivity to increase from the initially voided calandria case. One such hypothesis which leads to the maximum possible density of the steam/water mixture in the still unflooded calandria space is entrainment from the free surface. It is shown that the steam/water mixture density yielding the maximum reactivity peak cannot be achieved by entrainment because it exceeds thermohydraulically attainable densities of steam/water by an order of magnitude.

Calandria cooling structure in pressure tube reactor

International Nuclear Information System (INIS)

Hyugaji, Takenori; Sasada, Yasuhiro.

1976-01-01

Purpose: To contrive the structure of a heavy water distributing device in a pressure tube reactor thereby to reduce the variation in the cooling function thereof due to the welding deformation and installation error. Constitution: A heating water distributing plate is provided at the lower part of the upper tubular plate of a calandria tank to form a heavy water distributing chamber between both plates and a plurality of calandria tubes. Heavy water which has flowed in the upper part of the heavy water distributing plate from the heavy water inlet nozzle flows down through gaps formed around the calandria tubes, whereby the cooling of the calandria tank and the calandria tubes is carried out. In the above described calandria cooling structure, a heavy water distributing plate support is provided to secure the heavy water distributing plate and torus-shaped heavy water distributing rings are fixed to holes formed in the heavy water distributing plate penetrating through the calandria tubes thereby to form torus-shaped heavy water outlet ports each having a space. (Seki, T.)

CT Findings of Foreign Body Reaction to a Retained Endoloop Ligature Plastic Tube Mimicking Acute Appendicitis: A Case Report

Energy Technology Data Exchange (ETDEWEB)

Ahn, Jae Hong; Kang, Chae Hoon; Choi, Soo-Jung; Park, Man Soo; Jung, Seung Mun; Ryu, Dae Shick; Shin, Dong Rock [Department of Radiology, Asan Foundation, GangNeung Asan Hospital, University of Ulsan College of Medicine, Gangneung 25440 (Korea, Republic of)

2016-11-01

Many hospitals experience one or more retained surgical instrument events with risk of patient morbidity and medicolegal problems. Identification of retained surgical instrument is important. The radiologists should be familiar with imaging finding of retained surgical instrument. In a 62-year-old female with a retained plastic tube, localized peritoneal infiltration around air-containing tubular structure mimicked acute appendicitis on abdomen computed tomography (CT), one year after laparoscopic cholecystectomy. We reported CT findings of foreign body reaction related to retained Endoloop ligature plastic tube mimicking acute appendicitis.

High-voltage test and training of plastic streamer tubes for the DELPHI hadron calorimeter

International Nuclear Information System (INIS)

Alekseev, G.D.; Cellar, S.; Khomenko, B.A.; Korytov, A.V.; Kulinich, P.A.; Micelmacher, G.V.; Sedykh, Yu.V.; Toledo, R.

1987-01-01

The results of high-voltage test and training of plastic streamer tubes of the DELPHI hadron calorimeter are presented. The testing technique is considered in detail. The equipment for high-voltage training consists of a mini-computer, CAMAC-electronics, a controllable high-voltage supply and a digital ampermeter. The experimental results shows that high-voltage training of streamer tubes improves their characteristics. The value of dark current decreased up to 1 μA. The operational voltage range increased by a value more than 300 V

Useful effectiveness of plastic pipes for gas pipelines operating at low pressure

Energy Technology Data Exchange (ETDEWEB)

Zayutsev, K I

1976-06-01

Because a significant portion of the feeder lines in the Soviet union operate at relatively low pressures up to 75 to 150 psi, it is economically feasible to replace the conventional pipeline material (steel) with plastic. Cost savings result from lower material costs and ease of laying plastic pipe. Because of stiffness and corrosion requirements, the steel pipe used for these low-pressure pipelines is much thicker than needed to withstand the pressure used. Data are tabulated on the comparative costs and manpower requirements for the construction of 1 km of steel, polyvinyl chloride, and polyethylene gas pipelines ranging in diameter from 3 to 14 in. Generally, the plastic pipelines required 15 to 30% less man-days and were 20 to 35% lower in cost to build. The plastic pipelines can operate at 150 to 175 psi pressure and at temperatures up to 100 to 140/sup 0/F. In research conducted at VNIIST (All-Union Research Institute for the Construction of Trunk Pipelines) on plastic pipelines, a 2.5-mile test section has been operating successfully for 2 years, and new techniques and equipment for joining plastic pipe up to 25-in. diameter are being developed.

Moderator mixing after a pressure tube failure

International Nuclear Information System (INIS)

MacKinnon, J.C.; Fortman, R.A.; Hadaller, G.I.

1997-01-01

During a guaranteed shutdown state (GSS) in a CANDU reactor, there must be sufficient negative reactivity to ensure subcriticality in the event of a process failure. In one of the acceptable states, the reactor is kept subcritical by a high concentration of a neutron-absorbing chemical (the poison gadolinium nitrate) dissolved in the moderator (i.e., the moderator is guaranteed overpoisoned). A postulated accident scenario which is considered as a part of reactor safety analysis is the rupture of a fuel channel (i.e., a pressure tube/calandria tube break) when the reactor is in a GSS. If one of the channels in the core breaks (requiring a simultaneous failure of both the pressure tube and the surrounding calandria tube), coolant from the primary heat transport system will be discharged into the moderator, causing an associated displacement of fluid through relief ducts at the top of the calandria vessel. The incoming (unpoisoned) coolant may mix quickly with the moderator, or may mix slowly while displacing poisoned moderator through the relief ducts. The effectiveness of mixing generally depends on the break location, the coolant discharge rate and the moderator circulation. If an in-core loss of coolant accident occurred while the reactor is in this overpoisoned state, it must be guaranteed that even with the dilution of the poison by the incoming coolant the reactor will remain subcritical on both a local and global basis. This paper presents an overview of an experimental program in progress at the Moderator Test Facility at Stern Laboratories to investigate coolant/poison mixing for a simulated in-core fishmouth pressure tube/calandria tube rupture. The nominal system conditions investigated are of a reactor in a GSS, with coolant in the primary heat transport system at the same temperature as the heavily poisoned moderator, i.e., a depressurised 'cold' state. The results presented are those obtained during the commissioning of the modified Test Facility. The

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.

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

International Nuclear Information System (INIS)

Takikawa, N.

1982-01-01

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

Development of rolled joints for zirconium-2.5 wt % niobium pressure tubes

International Nuclear Information System (INIS)

Madhusoodanan, K.; Sinha, R.K.; Samuel, K.A.; Joeman, V.

1992-01-01

Due to its higher strength and lower deuterium pick-up rate, as compared to the existing cold worked zircaloy-2 material, cold worked zirconium-2.5 wt% niobium (Zr-2.5%Nb) alloy is to be used as the pressure tube material in all forthcoming Indian PHWRs starting with KAPP-2. These pressure tubes, which carry the fuel bundles are to be joined to the S.S 403 end-fittings through rolled joints. Since the new pressure tubes have a lower wall thickness and higher room temperature yield stress, than zircaloy-2 tubes the design parameters of the rolled joint had to be developed afresh. Further, since Zr-2.5%Nb is susceptible to delayed hydride cracking, it is necessary to limit the residual stress near the rolled joint to a minimum. Since the high residual stress is due to the initial assembly clearance between the pressure tube and end-fitting, a modified rolled joint had to be developed, referred to as zero clearance rolled joint. This paper provides details of the work carried out at Reactor Engineering Division of Bhabha Atomic Research Centre, Bombay towards the development of the design of the rolled joint as well as the tooling and procedures required for achieving zero-clearance fit-ups at site. The requirements to be met by the Zr-2.5% Nb pressure tubes for achieving acceptable rolled joints are highlighted. (author). 5 refs., 6 figs., 3 tabs

Evaluation of techniques for inspection and diagnostics of HWR pressure tubes

International Nuclear Information System (INIS)

Choi, Jong-Ho

2008-01-01

Efficient and accurate inspection and diagnostic techniques for various reactor components and systems, especially pressure tubes for Heavy Water Reactors (HWRs), 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 Project (CRP) on Inter-comparison of Techniques for HWR Pressure Tube Inspection and Diagnostics. The objective of the CRP was to inter-compare inspection and diagnostic techniques, in use and being developed, for structural integrity assessment of HWR pressure tubes. During the first phase of the CRP, participants investigated the capability of different techniques to detect and characterize flaws. During the second phase, participants collaborated to determine the hydrogen concentration and to detect and characterize hydride blisters in zirconium alloy pressure tubes. Eight organizations from six countries, which operate HWRs, have participated in this CRP, Most of the techniques examined are well established and many of them are regularly used during in-service inspection of pressure tubes. The inter-comparison of these techniques provides a platform for identifying a particular technique (or a set of techniques), which is more accurate and reliable as compared to others for a specified task. The CRP also witnessed some new methodologies, which can be implemented on in-service inspection tools. These new techniques could complement the existing ones to overcome their limitations, thereby improving the reliability and accuracy of in-service inspection. This CRP also identified future areas of research and development. (author)

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....

Collapse of experimental capsules under external pressure

International Nuclear Information System (INIS)

Simonen, F.A.; Shippell, R.J. Jr.

1980-01-01

Stress analyses and developmental tests of capsules fabricated from thick-walled tubing were performed for an external pressure design condition. In the design procedure no credit was taken for the expected margin in pressure between yielding of the capsule wall and catastrophic collapse or flattening. In tests of AISI-1018 low carbon steel capsules, a significant margin was seen between yield and collapse pressure. However, the experimental yield pressures were significantly below predictions, essentially eliminating the safety margin present in the conservative design approach. The differences between predictions and test results are attributed to deficiencies in the plasticity theories commonly in use for engineering stress analyses. The results of this study show that the von Mises yield condition does not accurately describe the yield behavior of the AISI-1018 steel tubing material for the triaxial stress conditions of interest. Finite element stress analyses successfully predicted the transition between uniform inward plastic deformation and ovalization that leads to catastrophic collapse. After adjustments to correct for the unexpected yield behavior of the tube material, the predicted pressure-deflection trends were found to follow the experimental data

Conservatism in methodologies for moderator subcooling sufficiency for fuel channel integrity upon pressure tube and calandria tube contact

Energy Technology Data Exchange (ETDEWEB)

Sun, L., E-mail: LSun@nbpower.com [Point Lepreau Generating Station, Lepreau, NB, (Canada)

2015-07-01

During a postulated large LOCA event in CANDU reactors, the pressure tube may balloon to contact with its surrounding calandria tube to transfer heat to the moderator. To confirm the integrity of the fuel channel in this case, many experiments have been performed in the last three decades. Based on the extant database of the pressure tube/calandria tube (PT/CT) contact, an analytical methodology was developed by Canadian Nuclear Industry to determine the sufficiency of moderator subcooling for fuel channel integrity. At the same time a semi-empirical methodology with an idea of Equivalent Moderator Subcooling (EMS) was also developed to judge the sufficiency of the moderator. In this work, some discussions were made over the two methodologies on their conservatism and it is demonstrated that the analytical approach is over conservative comparing with the EMS methodology. By using the EMS methodology, it is demonstrated that applying glass-peened calandria tubes, the requirement to moderator subcooling can be reduced by 10{sup o}C from that for smooth calandria tubes. (author)

Holographic NDE of pressure tubes for Cirene nuclear reactor

International Nuclear Information System (INIS)

Di Chirico, G.; Pirodda, L.; Villani, A.

1985-01-01

Pressure tubes for CIRENE nuclear reactor can be subjected to fretting corrosion of the inner walls. The resulting marks exhibit different geometries, whose influence on the structural behaviour of the tubes has been evaluated by means of a real time holographic technique. The paper shows the results of this investigation. Position and shape of internal defects have been directly visualized by observing holographic fringe distorsions on the outside surface of the tubes. Furthermore, through the fringe patterns, circumferential stress values have also been obtained. (Author) [pt

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

Electrochemical noise measurements under pressurized water reactor conditions

International Nuclear Information System (INIS)

Van Nieuwenhove, R.

2000-01-01

Electrochemical potential noise measurements on sensitized stainless steel pressure tubes under pressurized water reactor (PWR) conditions were performed for the first time. Very short potential spikes, believed to be associated to crack initiation events, were detected when stressing the sample above the yield strength and increased in magnitude until the sample broke. Sudden increases of plastic deformation, as induced by an increased tube pressure, resulted in slower, high-amplitude potential transients, often accompanied by a reduction in noise level

Experiments on vibration of heat exchanger tube arrays in cross flow

International Nuclear Information System (INIS)

Blevins, R.D.; Gibert, R.J.; Villard, B.

1981-08-01

A series of tests have been made at the Commissariat a l'Energie Atomique, in cooperation with General Atomic Company, SAN DIEGO (U.S.A.) on the flow-induced vibration of heat exchanger tube bundles in cross flow. These tests were made in air on tube bundles which simulated heat exchangers in the high temperature gas cooled reactors. The tests were of two types. In the first type, an instrumented tube was inserted at various locations into a tube bundle. Measurements were made of pressure at a number of points along the tube and about the circumference of the tube. These measurements were processed to obtain the spectra of turbulent pressure fluctuations on the tube, the spanwise correlation and the lift force. The second set of tests was made on tube bundles with flexible tubes. As the flow velocity was increased, these tests clearly show an instability. Nine tube configurations were tested with both plastic and metallic tubes and the effect of tube-to-tube difference in natural frequency was investigated

Steam generator tube rupture simulation using extended finite element method

Energy Technology Data Exchange (ETDEWEB)

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

2016-08-15

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

Steam generator tube rupture simulation using extended finite element method

International Nuclear Information System (INIS)

Mohanty, Subhasish; Majumdar, Saurin; Natesan, Ken

2016-01-01

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

Propagation of atmospheric-pressure ionization waves along the tapered tube

Science.gov (United States)

Xia, Yang; Wang, Wenchun; Liu, Dongping; Yan, Wen; Bi, Zhenhua; Ji, Longfei; Niu, Jinhai; Zhao, Yao

2018-02-01

Gas discharge in a small radius dielectric tube may result in atmospheric pressure plasma jets with high energy and density of electrons. In this study, the atmospheric pressure ionization waves (IWs) were generated inside a tapered tube. The propagation behaviors of IWs inside the tube were studied by using a spatially and temporally resolved optical detection system. Our measurements show that both the intensity and velocity of the IWs decrease dramatically when they propagate to the tapered region. After the taper, the velocity, intensity, and electron density of the IWs are improved with the tube inner diameter decreasing from 4.0 to 0.5 mm. Our analysis indicates that the local gas conductivity and surface charges may play a role in the propagation of the IWs under such a geometrical constraint, and the difference in the dynamics of the IWs after the taper can be related to the restriction in the size of IWs.

Inelastic analysis of finite length and depth cracked tubes

International Nuclear Information System (INIS)

Reich, M.; Gardner, D.; Prachuktam, S.; Chang, T.Y.

1977-01-01

Steam generator tube failure can at times result in reactor safety problems and subsequent premature reactor shutdowns. This paper concerns itself with the prediction of the failure pressures for typical PWR steam generator tubes with longitudinal finite length and finite depth cracks. Only local plastic overload failure is considered since the material is non-notch sensitive. Non-linear finite element analyses are carried out to determine the burst pressures of steam generator tubes containing longitudinal cracks located on the outer surface of the tubes. The non-linearities considered herein include elastic-plastic material behavior and large deformations. A non-proprietary general purpose non-linear finite element program, NFAP was adopted for the analysis. Due to the asymmetric nature of the cracks, two-dimensional, as well as three-dimensional finite element analyses, were performed. The two-dimensional element and its formulations are similar to those of NONSAP. The three-dimensional isoparametric element with elastic-plastic material characteristics together with the large deformation formulations used in NFAP are described in the Report BNL-20684. The numerical accuracy of the program was investigated and checked with known solutions of benchmark problems. In addition to the three-dimensional element which was specifically inserted into NFAP for this problem, other features such as direct pressure inputs for isoparametric elements, automatic load increment adjustments for convergent non-linear solutions, and automatic bandwidth reduction schemes are incorporated into the program thus allowing for a more economical evaluation of three-dimensional inelastic analysis. In summary the analysis clearly shows that for short cracks axial effects play a significant role. For long cracks, they are not important since two-dimensional conditions predominate and failure is governed by circumferential or hoop stress conditions

A software tool for evaluation of hydrogen ingress in CANDU pressure tubes

International Nuclear Information System (INIS)

Mihalache, Maria; Vasile, Radu; Deaconu, Mariea

2009-01-01

The prediction of hydrogen isotopes concentration into the body and in the rolled joints of operating pressure tubes as a function of reactor hot hours is very important in many fitness-for-service assessments and end of life estimates. The rolled joints are high stress zones with potential for delayed hydride cracking. Predictive models for assessing the long-term deuterium ingress in both body and rolled joint of the pressure tubes have been implemented in a software tool, ROHID, developed in INR-Pitesti. ROHID is a PC-based Windows application with a user-friendly interface that predicts the equivalent hydrogen ingress for Zr-2.5Nb pressure tubes. It uses colour-coded reactor core maps to display the predicted deuterium concentration as a function of time for selected axial locations. Plots of deuterium versus axial location and time for individual pressure tubes are also available. Also, the software tool can predict the exceeding of hydrogen terminal solid solubility (HTSS) from hydrides during precipitation and dissolving processes as a function of time and axial location. (authors)

Fracture toughness of irradiated Zr-2.5Nb pressure tube from Indian PHWR

Science.gov (United States)

Shah, Priti Kotak; Dubey, J. S.; Shriwastaw, R. S.; Dhotre, M. P.; Bhandekar, A.; Pandit, K. M.; Anantharaman, S.; Singh, R. N.; Chakravartty, J. K.

2015-03-01

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.

Thermal behaviour of pressure tube under fully and partially voided heating conditions using 19 pin fuel element simulator

International Nuclear Information System (INIS)

Yadav, Ashwini K.; Kumar, Ravi; Gupta, Akhilesh; Chatterjee, B.; Mukhopadhya, D.; Lele, H.G.

2011-01-01

In a nuclear reactor temperature can rise drastically during LOCA due to failure of heat transportation system and subsequently leads to mechanical deformations like sagging, ballooning and breaching of pressure tube. To understand the phenomenon an experiment has been carried out using 19 pin fuel element simulator. Main purpose of the experiment was to trace temperature profiles over the pressure tube, calandria tube and clad tubes of 220 MWe Indian Pressurised Heavy Water Reactor (IPHWR). The symmetrical heating of pressure tube of 1 m length was done through resistance heating of 19 pins under 13.5 kW power using a rectifier and the variation of temperatures over the circumference of pressure tube (PT), calandria tube (CT) and clad tubes were measured. The sagging of pressure tube was initiated at 460 deg C temperature and highest temperature attained was 650 deg C. The highest temperature attained by clad tubes was 680 deg C (over outer ring) and heat is dissipated to calandria vessel mainly due to radiation and natural convection. Again to simulate partially voided conditions, asymmetrical heating of pressure was carried out by injecting 8 kW power to upper 8 pins of fuel simulator. A maximum temperature difference of 295 deg C was observed over the circumference of pressure tube which highlights the magnitude of thermal stresses and its role in breaching of pressure tube under partially voided conditions. Integrity of pressure tube was retained during both symmetrical and asymmetrical heatup conditions. (author)

Stress relaxation analysis and irradiation creep and swelling in pressure tubes

International Nuclear Information System (INIS)

Beeston, J.M.; Burr, T.K.

1979-01-01

An analysis is presented of slit width test information on two pressure tubes that had been irradiated in test reactors. The analysis showed that differential swelling stresses and thermal stresses undergo relaxation. The mechanism responsible for the stress relaxation at temperatures less than 700 K was irradiation creep. Irradiation creep in thermal test reactor pressure tubes is evidently greater than it would be at equivalent conditions in fast reactors. The residual stresses observed in the slit width tests varied between 30 and 257 MPa and would act to reduce the operating stresses, thus allowing for increased service life of the tubes as compared with no stress relaxation

Garter spring location of pressure tube for PHWR using eddy current testing methods

International Nuclear Information System (INIS)

Lee, Y. S.; Yang, D. J.; Jeong, H. K.

2001-01-01

There are garter springs between pressure tube and calandria tube for PHWR. If the space of these garter springs become to be changed, the sagging of tube is caused and the contact between the pressure tube and calandria tube will cause the tube to be failed. AECL has applied the eddy current testing methods using send-receive type probe for this purpose, but this study apply eddy current testing methods using bobbin differential type probe to detection of garter spring location. And we did the computer simulation using VIC-3D code and compared it with experiments results for inspection 1 ∼ 11kHz. The results was that the garter spring signal was successfully detected for every frequency, and 5 kHz was best

Shock tubes: compressions in the low pressure chamber

International Nuclear Information System (INIS)

Schins, H.; Giuliani, S.

1986-01-01

The gas shock tube used in these experiments consists of a low pressure chamber and a high pressure chamber, divided by a metal-diaphragm-to-rupture. In contrast to the shock mode of operation, where incident and reflected shocks in the low pressure chamber are studied which occur within 3.5 ms, in this work the compression mode of operation was studied, whose maxima occur (in the low pressure chamber) about 9 ms after rupture. Theoretical analysis was done with the finite element computer code EURDYN-1M, where the computation was carried out to 30 ms

AMPTRACT: an algebraic model for computing pressure tube circumferential and steam temperature transients under stratified channel coolant conditions

International Nuclear Information System (INIS)

Gulshani, P.; So, C.B.

1986-10-01

In a number of postulated accident scenarios in a CANDU reactor, some of the horizontal fuel channels are predicted to experience periods of stratified channel coolant condition which can lead to a circumferential temperature gradient around the pressure tube. To study pressure tube strain and integrity under stratified flow channel conditions, it is, necessary to determine the pressure tube circumferential temperature distribution. This paper presents an algebraic model, called AMPTRACT (Algebraic Model for Pressure Tube TRAnsient Circumferential Temperature), developed to give the transient temperature distribution in a closed form. AMPTRACT models the following modes of heat transfer: radiation from the outermost elements to the pressure tube and from the pressure to calandria tube, convection between the fuel elements and the pressure tube and superheated steam, and circumferential conduction from the exposed to submerged part of the pressure tube. An iterative procedure is used to solve the mass and energy equations in closed form for axial steam and fuel-sheath transient temperature distributions. The one-dimensional conduction equation is then solved to obtain the pressure tube circumferential transient temperature distribution in a cosine series expansion. In the limit of large times and in the absence of convection and radiation to the calandria tube, the predicted pressure tube temperature distribution reduces identically to a parabolic profile. In this limit, however, radiation cannot be ignored because the temperatures are generally high. Convection and radiation tend to flatten the parabolic distribution

Methodologies for assessment of the service life of pressure tubes in Indian PHWRs

International Nuclear Information System (INIS)

Sinha, R.K.; Sharma, A.; Madhusoodanan, K.; Sinha, S.K.; Malshe, U.D.

1997-01-01

For estimating safe service life of pressure tubes in Indian PHWRs, analytical methodologies have been developed to evaluate creep deformation, deuterium pick-up rate, blister growth at cold spot, and operating domain required for achieving leak-before-break. The paper provides an overview of these methodologies, and results of some studies carried out towards evolution of proposed fitness-for-service criteria for a pressure tube in contact with its calandria tube. (author)

Heat transfer and pressure drop of condensation of hydrocarbons in tubes

Science.gov (United States)

Fries, Simon; Skusa, Severin; Luke, Andrea

2018-03-01

The heat transfer coefficient and pressure drop are investigated for propane. Two different mild steel plain tubes and saturation pressures are considered for varying mass flux and vapour quality. The pressure drop is compared to the Friedel-Correlation with two different approaches to determine the friction factor. The first is calculation as proposed by Friedel and the second is through single phase pressure drop investigations. For lower vapour qualities the experimental results are in better agreement with the approach of the calculated friction factor. For higher vapour qualities the experimental friction factor is more precise. The pressure drop increases for a decreasing tube diameter and saturation pressure. The circumferential temperature profile and heat transfer coefficients are shown for a constant vapour quality at varying mass fluxes. The subcooling is highest for the bottom of the tube and lowest for the top. The average subcooling as well as the circumferential deviation decreases for rising mass fluxes. The averaged heat transfer coefficients are compared to the model proposed by Thome and Cavallini. The experimental results are in good agreement with both correlations, however the trend is better described with the correlation from Thome. The experimental heat transfer coefficients are under predicted by Thome and over predicted by Cavallini.

Detection of pressure tube leaks relying on moisture beetles only

International Nuclear Information System (INIS)

Kenchington, J.M.; Choi, A.; Jin, Y.

2004-01-01

A major decision was made for Pickering NGS A Annulus Gas System (ACS) that detection of a pressure tube (PT) leak should be achieved by using only moisture beetles and that dew point monitors would provide 'early warning' without status to shut down the reactor. Experience with Unit 3 has shown that dew point monitoring of pressure tube leaks was particularly subject to gas leaks and surface adsorption effects. Unit 4 was the first one to be converted during the full scale pressure tube replacement programme. Because of the fundamental change in design philosophy, moisture injection tests were carried out during commissioning to demonstrate that performance matched design. In particular it was necessary to show that leak before break (LBB) would be achieved if a leak occurred in the limiting string. Units 1 and 3 have since been converted. No decision has been taken to convert Pickering B units as gas leaks are small and no significant adsorption effects are anticipated. Hence dew point monitoring will not be impaired. (author)

Tensile and fracture toughness characteristics of Zr-2.5Nb pressure tube

International Nuclear Information System (INIS)

Jung, H. C.; Kim, Y. S.; Ahn, S. B.; Kim, S. S.; Im, K. S.

2004-01-01

The object of this study is to evaluate the characteristics of tensile and fracture toughness of Zr-2.5Nb pressure tube. The transverse tensile tests were performed at various temperatures and the fracture toughness tests were carried out at room temperature using the CCT (curved compact tension) specimen. These specimens were directly machined from the pressure tube retaining original curvatures. Also, the fracture toughness of two sets of Zr-2.5Nb manufactured at different time was compared. The chemical analysis and the Vicker's hardness tests were performed at two sets of Zr-2.5Nb pressure tube. The Vicker's hardness value of SET-2 containing more oxygen and carbon relatively was higher about 11 than that of SET-1

Study on the manufacturing process, causes of the pressure tube failure and methods for improving its performance

Energy Technology Data Exchange (ETDEWEB)

You, Ho Sik; Jeong, Jin Kon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-07-01

Manufacturing processes of Zr-2.5Nb pressure tube used in CANDU reactor, effects of impurities on the properties of the pressure tube, experiences and causes of the pressure tube cracking accident and the development programs on the fuel channel at AECL have been described. Fabrication processes on the pressure tube have been explained in detail from the sponge production step to the final product. Test methods that are performed to verify the integrity of the final product have also been described. Most of the pressure tube rupture accidents were caused by DHC (Delayed Hydride Cracking). In cases of Pickering units 3 and 4 and Bruce unit 2, excessive residual stresses induced by improper rolled joint process had played a role to cause DHC. In Pickering unit 2, cracks formed by contact between pressure and calandria tubes due to the movement of garter spring were direct cause of failure. After the accidents, a lot of R and D programs on each component of the fuel channel have been carried out. The study on the improvement of manufacturing processes such as increasing cold working rate, performing the intermediate and final annealing and adding the third element like Fe, V, Cr for enhancing the pressure tube performance are on progress. To suppress hydrogen uptake into the pressure tube, the methods such as zirconia coating on the pressure tube, Cr-plating on the end fitting and placing the yttrium getter on the pressure tube are considered. Experiments on each test specimen are currently under way. Owing to such an effort, more advanced fuel channel can be installed in the next CANDU reactor. 6 tabs., 20 figs., 20 refs. (Author).

Application of the VAW tube digester for metallurgical pressure-leaching processes

International Nuclear Information System (INIS)

Kaempf, F.; Pietsch, H.B.

1978-01-01

Problems associated with the treatment of complex and refractory ores or concentrates, as well as those related to environmental factors, have led to increased interest in hydrometallurgy under elevated temperatures and pressures. Pressure leaching can be carried out in vertical, horizontal or spherical autoclaves equipped with mechanical agitators. If high throughput capacities are catered for, the division of a conventional plant into several units is inevitable. By contrast, the VAW (Vereinigte Aluminium-Werke Aktiengesellschaft) tube digester enables hydrometallurgical processes to be carried out under pressure and at a high temperature with the use of a basically simple technology, extremely high specific throughput and improved thermal economics being achieved. The advantages of the tube digester over vessel autoclaves are described, and details of laboratory investigations into the applicability of tube digesters to various metallurgical applications are given. Test results are given for the leaching of refractory uranium ores. (author)

Pressure tube reactor

International Nuclear Information System (INIS)

Matsumoto, Tomoyuki; Fujino, Michihira.

1980-01-01

Purpose: To equalize heavy water flow distribution by providing a nozzle for externally injecting heavy water from a vibration preventive plate to the upper portion to feed the heavy water in a pressure tube reactor and swallowing up heavy water in a calandria tank to supply the heavy water to the reactor core above the vibration preventive plate. Constitution: A moderator injection nozzle is mounted on the inner wall of a calandria tank. Heavy water is externally injected above the vibration preventive plate, and heavy water in the calandria tank is swallowed up to supply the heavy water to the core reactor above the vibration preventive plate. Therefore, the heavy water flow distribution can be equalized over the entire reactor core, and the distribution of neutron absorber dissolved in the heavy water is equalized. (Yoshihara, H.)

Plastic limit pressure of spherical vessels with combined hardening involving large deformation

International Nuclear Information System (INIS)

Leu, S.-Y.; Liao, K.-C.; Lin, Y.-C.

2014-01-01

The paper aims to investigate plastic limit pressure of spherical vessels of nonlinear combined isotropic/kinematic hardening materials. The Armstrong-Frederick kinematic hardening model is adopted and the Voce hardening law is incorporated for isotropic hardening behavior. Analytically, we extend sequential limit analysis to deal with combined isotropic/kinematic hardening materials. Further, exact solutions of plastic limit pressure were developed analytically by conducting both static and kinematic limit analysis. The onset of instability was also derived and solved iteratively by Newton's method. Numerically, elastic–plastic analysis is also performed by the commercial finite-element code ABAQUS incorporated with the user subroutine UMAT implemented with user materials of combined hardening. Finally, the problem formulation and the solution derivations presented here are validated by a very good agreement between the numerical results of exact solutions and the results of elastic–plastic finite-element analysis by ABAQUS. -- Highlights: • Sequential limit analysis is extended to consider combined hardening. • Exact solutions of plastic limit pressure are developed. • The onset of instability of a spherical vessel is derived and solved numerically

The impact law of confining pressure and plastic parameter on Dilatancy of rock

Science.gov (United States)

Wang, Bin; Zhang, Zhenjie; Zhu, Jiebing

2017-08-01

Based on cyclic loading-unloading triaxle test of marble, the double parameter dilation angle model is established considering confining pressure effect and plastic parameter. Research shows that not only the strength but also the militancy behavior is highly depended on its confining pressure and plastic parameter during process of failure. Dilation angle evolution law shows obvious nonlinear characteristic almost with a rapid increase to the peak and then decrease gradually with plastic increasing, and the peak dilation angle value is inversely proportional with confining pressure. The proposed double parameter nonlinear dilation angle model can be used to well describe the Dilatancy of rock, which helps to understand the failure mechanism of surrounding rock mass and predict the range of plastic zone.

Effect of tubing condensate on non-invasive positive pressure ventilators tested under simulated clinical conditions.

Science.gov (United States)

Hart, Diana Elizabeth; Forman, Mark; Veale, Andrew G

2011-09-01

Water condensate in the humidifier tubing can affect bi-level ventilation by narrowing tube diameter and increasing airflow resistance. We investigated room temperature and tubing type as ways to reduce condensate and its effect on bi-level triggering and pressure delivery. In this bench study, the aim was to test the hypothesis that a relationship exists between room temperature and tubing condensate. Using a patient simulator, a Res-med bi-level device was set to 18/8 cm H(2)O and run for 6 h at room temperatures of 16°C, 18°C and 20°C. The built-in humidifier was set to a low, medium or high setting while using unheated or insulated tubing or replaced with a humidifier using heated tubing. Humidifier output, condensate, mask pressure and triggering delay of the bi-level were measured at 1 and 6 h using an infrared hygrometer, metric weights, Honeywell pressure transducer and TSI pneumotach. When humidity output exceeded 17.5 mg H(2)O/L, inspiratory pressure fell by 2-15 cm H(2)O and triggering was delayed by 0.2-0.9 s. Heating the tubing avoided any such ventilatory effect whereas warmer room temperatures or insulating the tubing were of marginal benefit. Users of bi-level ventilators need to be aware of this problem and its solution. Bi-level humidifier tubing may need to be heated to ensure correct humidification, pressure delivery and triggering.

Hierarchically structured nanoporous carbon tubes for high pressure carbon dioxide adsorption

Directory of Open Access Journals (Sweden)

Julia Patzsch

2017-05-01

Full Text Available Mesoscopic, nanoporous carbon tubes were synthesized by a combination of the Stoeber process and the use of electrospun macrosized polystyrene fibres as structure directing templates. The obtained carbon tubes have a macroporous nature characterized by a thick wall structure and a high specific surface area of approximately 500 m²/g resulting from their micro- and mesopores. The micropore regime of the carbon tubes is composed of turbostratic graphitic areas observed in the microstructure. The employed templating process was also used for the synthesis of silicon carbide tubes. The characterization of all porous materials was performed by nitrogen adsorption at 77 K, Raman spectroscopy, infrared spectroscopy, thermal gravimetric analysis (TGA, scanning electron microscopy (SEM as well as transmission electron microscopy (TEM. The adsorption of carbon dioxide on the carbon tubes at 25 °C at pressures of up to 30 bar was studied using a volumetric method. At 26 bar, an adsorption capacity of 4.9 mmol/g was observed. This is comparable to the adsorption capacity of molecular sieves and vertically aligned carbon nanotubes. The high pressure adsorption process of CO2 was found to irreversibly change the microporous structure of the carbon tubes.

Calculation of fast neutron flux in reactor pressure tubes and experimental facilities

Energy Technology Data Exchange (ETDEWEB)

Barnett, P. C. [Canadian General Electric (Canada)

1968-07-15

The computer program EPITHET was used to calculate the fast neutron flux (>1 MeV) in several reactor pressure tubes and experimental facilities in order to compare the fast neutron flux in the different cases and to provide a self-consistent set of flux values which may be used to relate creep strain to fast neutron flux . The facilities considered are shown below together with the calculated fast neutron flux (>1 MeV). Fast flux 10{sup 13} n/cm{sup 2}s: NPD 1.14, Douglas Point 2.66, Pickering 2.89, Gentilly 2.35, SGHWR 3.65, NRU U-1 and U-2 3.25'' pressure tube - 19 element fuel 3.05, NRU U-1 and U-2 4.07'' pressure tube - 28 element fuel 3.18, NRU U-1 and U-2 4.07'' pressure tube - 18 element fuel 2.90, NRX X-5 0.88, PRTR Mk I fuel 2.81, PRTR HPD fuel 3.52, WR-1 2.73, Mk IV creep machine (NRX) 0.85, Mk VI creep machine (NRU) 2.04, Biaxial creep insert (NRU U-49) 2.61.

Deformation of in-service pressure tubes

International Nuclear Information System (INIS)

Sarce, A.L.

1993-01-01

Candu type nuclear reactor pressure tubes suffer deformations during operation. This are consequences of irradiation growth and creep. By means of a computer code which takes into account the material microstructure, the above mentioned deformations are calculated, and results are compared with corresponding values measured at Embalse nuclear power plant. The calculations make explicit inclusion of intergranular stresses caused by an isotropy in the material. (author). 1 ref

A statistical method for draft tube pressure pulsation analysis

International Nuclear Information System (INIS)

Doerfler, P K; Ruchonnet, N

2012-01-01

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.

Fuel-element vibration and bearing pad to pressure tube fretting

International Nuclear Information System (INIS)

Fisher, N.J.; Taylor, C.E.; Pettigrew, M.J.

1990-08-01

Fuel channel operation under boiling condition results in increased flow velocities, which may lead to unacceptable fuel-element vibration and bearing pad to pressure tube fretting. The existing endurance test database does not fully cover the range of future channel operating conditions. In particular, after refuelling, some channels for future designs may operate with two-phase flow conditions outside the range of endurance test conditions. Full-scale endurance testing at realistic steam-water conditions involves substantial energy costs. Therefore, fundamental laboratory investigations were conducted to define and endurance test matrix which adequately envelops the future range of operating conditions while minimizing both the number of tests and the energy requirement of individual tests. The main focus of the laboratory investigations was to establish the relationships between: fuel channel flow conditions and fuel-element vibration; and fuel-element vibration and bearing pad to pressure tube fretting. The vibration response of a single fuel element was measured over a wide range of operating conditions covering realistic fuel channel conditions and simulated endurance testing conditions. For higher void fractions, the vibration amplitudes measured in air/water were much higher than in steam/water, while for low void fractions, the amplitudes were similar. The measured amplitudes in steam/water varied very little over the range of temperature and pressure investigated. The effects of temperature, pressure tube oxide thickness, vibration amplitude and bearing pad manufacturer on pressure tube fretting were investigated. The fretting rate is extremely temperature dependent. For vibration amplitudes about three or four times greater than expected in-reactor conditions, peak fretting rates were observed in the 225 to 286 degrees C temperature range. Fretting rates were seven times less at the higher temperatures of 300 and 315 degrees C, and the lower temperatures

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

International Nuclear Information System (INIS)

Agostini, F.

2008-07-01

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 T sat = 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)

Dynamic Measurements of Plastic Deformation in a Water-Filled Aluminum Tube in Response to Detonation of a Small Explosives Charge

Directory of Open Access Journals (Sweden)

Harold Sandusky

1999-01-01

Full Text Available Experiments have been conducted to benchmark computer code calculations for the dynamic interaction of explosions in water with structures. Aluminum cylinders with a length slightly more than twice their diameter were oriented vertically, sealed on the bottom by a thin plastic sheet, and filled with distilled water. An explosive charge suspended in the center of the tube plastically deformed but did not rupture the wall. Tube wall velocity, displacement, and strain were directly measured. The agreement among the three sets of dynamic data and the agreement of the terminal displacement measurements with the residual deformation were excellent.

Statistical analysis and modelling of in-reactor diametral creep of Zr-2.5Nb pressure tubes

Energy Technology Data Exchange (ETDEWEB)

Jyrkama, Mikko I., E-mail: mjyrkama@uwaterloo.ca [Department of Civil and Environmental Engineering, University of Waterloo, 200 University Avenue West, Waterloo, ON, Canada N2L 3G1 (Canada); Bickel, Grant A., E-mail: grant.bickel@cnl.ca [Canadian Nuclear Laboratories, Chalk River Laboratories, Chalk River, ON, Canada K0J 1J0 (Canada); Pandey, Mahesh D., E-mail: mdpandey@uwaterloo.ca [Department of Civil and Environmental Engineering, University of Waterloo, 200 University Avenue West, Waterloo, ON, Canada N2L 3G1 (Canada)

2016-04-15

Highlights: • New and simple statistical model of pressure tube diametral creep. • Based on surveillance data of 328 pressure tubes from eight different CANDU reactors. • Uses weighted least squares (WLS) to regress out operating conditions. • The shape of the diametral creep profiles are predicted very well. • Provides insight and relative ranking of strain behaviour of in-service tubes. - Abstract: This paper presents the development of a simplified regression approach for modelling the diametral creep over time in Zr-2.5 wt% Nb pressure tubes used in CANDU reactors. The model is based on a large dataset of in-service inspection data of 328 different pressure tubes from eight different CANDU reactor units. The proposed weighted least squares (WLS) regression model is linear in time as a function of flux and temperature, with a temperature-dependent variance function. The model predicts the shape of the observed diametral creep profiles very well, and is useful not merely for prediction, but also for assessing tube-to-tube variability and manufacturing properties among the inspected tubes.

Investigation of the failure of a reactor pressure vessel by plastic instability

International Nuclear Information System (INIS)

Laemmer, H.; Ritter, B.

1994-01-01

A possible consequence of a core meltdown accident in a pressurized water reactor is the failure of the reactor pressure vessel under high internal pressure. With the aid of the finite element program ABAQUS and using a material model of the thermo-plasticity for large deformation, the failure of the reactor pressure vessel due to plastic instability was examined. It was apparent from the finite element calculations that solely due to reduction in strength of the material, even for internal wall temperatures clearly below the core melt; of about 2000 C, the critical internal pressure can fall to values which are lower than the working pressure. With the aid of simplified geometry, a lower limit for the pressure at failure of the reactor pressure vessel can be calculated. (orig./HP) [de

High pressure studies of fluorenone emission in plastic media

International Nuclear Information System (INIS)

Mitchell, D.J.; Schuster, G.B.; Drickamer, H.G.

1977-01-01

The energy and the quantum efficiency for fluorenone fluorescence in the crystalline state and in polymeric matrices was measured as a function of external pressure over the range 0--140 kbar. The application of high pressure induces changes in the quantum yield, which ranges from 0.001 at low pressure to a maximum of approx.0.1 at high pressure in hydrocarbon plastics. These results are interpreted as arising from the decrease in the energy of the lowest ππ excited singlet state relative to other relevant states as the external pressure is increased

Prediction of pressure tube fretting-wear damage due to fuel vibration

International Nuclear Information System (INIS)

Yetisir, M.; Fisher, N.J.

1997-01-01

Fretting marks between fuel bundle bearing pads and pressure tubes have been observed at the inlet end of some Darlington Nuclear Generating Station (NGS) and Bruce NGS fuel channels. The excitation mechanisms that lead to fretting are not fully understood. In this paper, the possibility of bearing pad-to-pressure tube fretting due to turbulence-induced motion of the fuel element is investigated. Numerical simulations indicate that this mechanism by itself is not likely to cause the level of fretting experienced in Darlington and Bruce NGSs. (orig.)

Development of out-of-core concepts for a supercritical-water, pressure-tube reactor

International Nuclear Information System (INIS)

Diamond, W.T.

2010-01-01

One of the Generation IV programs at Chalk River Laboratories has as its prime focus the development of out-of-core concepts for the SuperCritical Water (SCW) pressure tube reactor under development in Canada. A number of technical issues associated with the interface of out-of-core components and the pressure tubes of a SCW pressure tube reactor are being investigated. This article focuses on several aspects of out-of-core components and layouts, building upon concepts that have been developed during the past few years. The efforts are strongly focused on concepts for a fuel channel that can be fabricated with the tight lattice pitch (typically 230 to 250 mm) that may be required for some applications such as utilization of a thorium fuel cycle. It is not practical to adapt concepts with a tight lattice pitch while using the thicker materials required for the higher temperatures and pressures required for supercritical operation. A change in lattice pitch or configuration is required to accommodate the component size increases. This presentation will cover a number of new concepts developed to produce feeders and end fittings for the harsh conditions of a SCW pressure tube reactor. These components are then developed into conceptual models of a Gen IV pressure tube reactor mounted in both horizontal and vertical orientations. Full 3-D solid models of both concepts will be demonstrated as well as a 1/10th-scale model of one face of a horizontal concept that has been built from components made with a 3-D printer. (author)

Integrated probabilistic assessment for DHC initiation, growth and leak-before-break of PHWR pressure tubes

Energy Technology Data Exchange (ETDEWEB)

Oh, Young-Jin [Power Engineering Research Institute, KEPCO Engineering and Construction, 188 Gumi-dong, Bundang-gu, Seongnam-si, Gyeonggi-do 463-870 (Korea, Republic of); Chang, Yoon-Suk, E-mail: yschang@khu.ac.kr [Department of Nuclear Engineering, Kyung Hee University, 1732 Deogyeong-daero, Giheung-gu, Yongin-si, Gyeonggi-do 446-701 (Korea, Republic of)

2014-08-15

Highlights: • We develop an integrated approach for probabilistic assessment of PHWR pressure tube. • We examine probabilities of DHC initiation, growth, penetration and LBB failure. • The proposed approach is helpful to calculate rupture probabilities in reactor flaws even in the case of very low rupture probability. - Abstract: A few hundred zirconium alloy pressure tubes in a pressurized heavy water reactor (PHWR) serve as the nuclear fuel channel, as well as the reactor coolant pressure boundary. The pressure tubes are inspected periodically and a fitness-for-service assessment (FFSA) must be conducted if any flaw is detected in the inspection. A Canadian standard provides FFSA procedures of PHWR pressure tubes, which include probabilistic assessment for flaws considering delayed hydride cracking (DHC) and leak-before-break (LBB). In the present study, an integrated approach with detailed stepwise calculation procedures and integration methodology for probabilistic assessment of pressure tube was developed. In the first step of this approach, a probability of the DHC initiation, growth and penetration for single initial flaw is calculated. In the next step, a probability of LBB failure, which means tube rupture, for single through-wall crack (TWC) is calculated. Finally, a rupture probability for all initial flaws in a reactor can be calculated using the penetration probability for single flaw and the LBB failure probability for single TWC, as well as the predicted total number of initial flaw in the reactor.

Integrated probabilistic assessment for DHC initiation, growth and leak-before-break of PHWR pressure tubes

International Nuclear Information System (INIS)

Oh, Young-Jin; Chang, Yoon-Suk

2014-01-01

Highlights: • We develop an integrated approach for probabilistic assessment of PHWR pressure tube. • We examine probabilities of DHC initiation, growth, penetration and LBB failure. • The proposed approach is helpful to calculate rupture probabilities in reactor flaws even in the case of very low rupture probability. - Abstract: A few hundred zirconium alloy pressure tubes in a pressurized heavy water reactor (PHWR) serve as the nuclear fuel channel, as well as the reactor coolant pressure boundary. The pressure tubes are inspected periodically and a fitness-for-service assessment (FFSA) must be conducted if any flaw is detected in the inspection. A Canadian standard provides FFSA procedures of PHWR pressure tubes, which include probabilistic assessment for flaws considering delayed hydride cracking (DHC) and leak-before-break (LBB). In the present study, an integrated approach with detailed stepwise calculation procedures and integration methodology for probabilistic assessment of pressure tube was developed. In the first step of this approach, a probability of the DHC initiation, growth and penetration for single initial flaw is calculated. In the next step, a probability of LBB failure, which means tube rupture, for single through-wall crack (TWC) is calculated. Finally, a rupture probability for all initial flaws in a reactor can be calculated using the penetration probability for single flaw and the LBB failure probability for single TWC, as well as the predicted total number of initial flaw in the reactor

Characterizing the active opening of the eustachian tube in a hypobaric/hyperbaric pressure chamber.

Science.gov (United States)

Mikolajczak, Stefanie; Meyer, Moritz Friedo; Hahn, Moritz; Korthäuer, Christine; Jumah, Masen Dirk; Hüttenbrink, Karl-Bernd; Grosheva, Maria; Luers, Jan Christoffer; Beutner, Dirk

2015-01-01

Active and passive opening of the Eustachian tube (ET) enables direct aeration of the middle ear and a pressure balance between middle ear and the ambient pressure. The aim of this study was to characterize standard values for the opening pressure (ETOP), the opening frequency (ETOF), and the opening duration (ETOD) for active tubal openings (Valsalva maneuver, swallowing) in healthy participants. In a hypobaric/hyperbaric pressure chamber, 30 healthy participants (19 women, 11 men; mean age, 25.57 ± 3.33 years) were exposed to a standardized profile of compression and decompression. The pressure values were recorded via continuous impedance measurement during the Valsalva maneuver and swallowing. Based on the data, standard curves were identified and the ETOP, ETOD, and ETOF were determined. Recurring patterns of the pressure curve during active tube opening for the Valsalva maneuver and for active swallowing were characterized. The mean value for the Valsalva maneuver for ETOP was 41.21 ± 17.38 mbar; for the ETOD, it was 2.65 ± 1.87 seconds. In the active pressure compensation by swallowing, the mean value for the ETOP was 29.91 ± 13.07 mbar; and for the ETOD, it was 0.82 ± 0.53 seconds. Standard values for the opening pressure of the tube and the tube opening duration for active tubal openings (Valsalva maneuver, swallowing) were described, and typical curve gradients for healthy subjects could be shown. This is another step toward analyzing the function of the tube in compression and decompression.

Design of a high-pressure single pulse shock tube for chemical kinetic investigations

International Nuclear Information System (INIS)

Tranter, R. S.; Brezinsky, K.; Fulle, D.

2001-01-01

A single pulse shock tube has been designed and constructed in order to achieve extremely high pressures and temperatures to facilitate gas-phase chemical kinetic experiments. Postshock pressures of greater than 1000 atmospheres have been obtained. Temperatures greater than 1400 K have been achieved and, in principle, temperatures greater than 2000 K are easily attainable. These high temperatures and pressures permit the investigation of hydrocarbon species pyrolysis and oxidation reactions. Since these reactions occur on the time scale of 0.5--2 ms the shock tube has been constructed with an adjustable length driven section that permits variation of reaction viewing times. For any given reaction viewing time, samples can be withdrawn through a specially constructed automated sampling apparatus for subsequent species analysis with gas chromatography and mass spectrometry. The details of the design and construction that have permitted the successful generation of very high-pressure shocks in this unique apparatus are described. Additional information is provided concerning the diaphragms used in the high-pressure shock tube

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

Energy Technology Data Exchange (ETDEWEB)

Jarron, D.; Trelinski, M.; Kretz, S. [Ontario Power Generation, Ajax, Ontario (Canada)]. E-mail: don.jarron@opg.com; mike.trelinski@opg.com; steve.kretz@opg.com

2006-07-01

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)

Study of the performances of acoustic emission testing for glass fibre reinforced plastic pipes containing defects; Etude des performances du controle par emission acoustique de tubes en materiaux composites resine-fibre de verre contenant des defauts

Energy Technology Data Exchange (ETDEWEB)

Villard, D.; Vidal, M.C.

1995-08-01

Glass fibre reinforced plastic pipes are more and more often used, in nuclear power plants, for building or replacement of water pipings classified `nuclear safety`. Tests have been performed to evaluate the performances of acoustic emission testing for in service inspection of these components. The tests were focused on glass fibre reinforced polyester and vinyl-ester pipes, in as received conditions or containing impacts, and intentionally introduced defects. They have been carried out by CETIM, following the ASTM Standard E 1118 (code CARP), to a maximum pressure lever of 25 Bar The results show that the CARP procedure can be used for detection of defects and evaluation of their noxiousness towards internal pressure: most of the tubes containing low energy impacts could not be distinguished from tubes without defect; on the other hand the important noxiousness of lacks of impregnation of roving layer appeared clearly. Complementary tests have been performed on some tubes at a more important pressure lever, for which the damage of the tubes in enough to deteriorate there elastic properties. The results showed that CARP procedure give valuable informations on damage level. It would be interesting to evaluate acoustic emission on tubes containing realistic in-service degradations. (author). 11 refs., 15 figs., 6 tabs., 2 appends.

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...

Fuel bundle to pressure tube fretting in Bruce and Darlington

Energy Technology Data Exchange (ETDEWEB)

Norsworthy, A G; Ditschun, A [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

1996-12-31

As the fuel channel elongates due to creep, the fuel string moves relative to the inlet until the fuel pads at the inboard end eventually separate from the spacer sleeve, and the fuel resides on the burnish mark of the pressure tube. The bundle is then supported in a fashion which contributes to increased levels of vibration. Those pads which (due to geometric variation) have contact loads with the pressure tube within a certain range, vibrate, and cause significant fretting on the burnish mark, and further along at the midplane of the bundle. Inspection of the pressure tubes in Bruce A, Bruce B, and Darlington has revealed fret damage up to 0.55 mm at the burnish mark and slightly lower than this at the inlet bundle midplane. To date, all fret marks have been dealt with successfully without the need for tube replacement, but a program of work has been initiated to understand the mechanism and reduce the fretting. Such understanding is necessary to guide future design changes to the fuel bundle, to guide future inspection programs, to guide maintenance programs, and for longer term strategic planning. This paper discusses how the understanding of fretting has evolved and outlines a current hypothesis for the mechanism of fretting. The role of bundle geometry, excitation forces, and reactor conditions are reviewed, along with options under consideration to mitigate damage. (author). 4 refs., 2 tabs., 13 figs.

Fuel bundle to pressure tube fretting in Bruce and Darlington

International Nuclear Information System (INIS)

Norsworthy, A.G.; Ditschun, A.

1995-01-01

As the fuel channel elongates due to creep, the fuel string moves relative to the inlet until the fuel pads at the inboard end eventually separate from the spacer sleeve, and the fuel resides on the burnish mark of the pressure tube. The bundle is then supported in a fashion which contributes to increased levels of vibration. Those pads which (due to geometric variation) have contact loads with the pressure tube within a certain range, vibrate, and cause significant fretting on the burnish mark, and further along at the midplane of the bundle. Inspection of the pressure tubes in Bruce A, Bruce B, and Darlington has revealed fret damage up to 0.55 mm at the burnish mark and slightly lower than this at the inlet bundle midplane. To date, all fret marks have been dealt with successfully without the need for tube replacement, but a program of work has been initiated to understand the mechanism and reduce the fretting. Such understanding is necessary to guide future design changes to the fuel bundle, to guide future inspection programs, to guide maintenance programs, and for longer term strategic planning. This paper discusses how the understanding of fretting has evolved and outlines a current hypothesis for the mechanism of fretting. The role of bundle geometry, excitation forces, and reactor conditions are reviewed, along with options under consideration to mitigate damage. (author). 4 refs., 2 tabs., 13 figs

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

International Nuclear Information System (INIS)

Majumdar, S.; Kasza, K.

2009-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2009-07-01

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

The EL-4 reactor. Changing of a pressure tube on a test loop

International Nuclear Information System (INIS)

Foulquier, H.; Clara, P.

1964-01-01

Right from the beginning of the EL-4 project, the research convected with the overall design of the reactor was guided by the various technical specifications resulting from a justifiable concern about the reliability. The external and internal tubes of each layer situated in the reactor block had in particular to be interchangeable. The research alone into the dismantling of the external tube, i.e in fact the pressure tube, justified a certain number of full-scale tests on a model. The tests carried out under relevant conditions on a non-irradiated structure made it possible to define a complete ranger of of positioning and un-positioning sequences at a distance for such a pressure tube. (authors) [fr

Axial and transverse tensile properties Zr-2.5Nb pressure tube off-cuts

International Nuclear Information System (INIS)

Shah, Priti K.; Dubey, J.S.; Shriwastaw, R.S.; Balakrishnan, K.S.; Anantharaman, S.; Chakravartty, J.K.

2011-01-01

Zr-2.5Nb alloys in cold worked and stress relieved (CWSR) condition serves as pressure boundary for hot coolant in Indian Pressurized Heavy Water Reactor (IPHWR). Due to both microstructural and crystallographic anisotropy, the mechanical properties in general and fracture behavior in particular are anisotropic for this material. To understand the anisotropic mechanical behavior of the Zr-2.5Nb pressure tubes of IPHWRs, tension tests were carried out on the RAPS-2 and KAPS-2 pressure tube off-cuts. Off-cuts are the small pieces cut from the two ends of a pressure tube before installing it into the PHWR. Miniature flat tensile specimens (without applying any flattening treatment to the pressure tube) of both longitudinal (or axial) and transverse orientation were fabricated from the off-cuts. Tension tests were carried out both at room temperature and 300 deg C. The transverse specimens showed higher strength and lower elongation than that of the axial ones. The back-end off-cuts showed higher strength compared to front-end off-cuts along axial direction whereas the same is not true for transverse direction specimens. One typical plot obtained in the testing of one of the off-cut is shown. The paper will discuss about the importance of the work carried out, test specimens, test method and results obtained in detail

Experimental and visual study on flow patterns and pressure drops in U-tubes

International Nuclear Information System (INIS)

Da Silva Lima, J. R.

2011-01-01

In single- and two-phase flow heat exchangers (in particular 'coils'), besides the straight tubes there are also many singularities, in particular the 180° return bends (also called return bends or U-bends). However, contrary to the literature concerning pressure drops and heat transfer in straight tubes, where many experimental data and predicting methods are available, only a limited number of studies concerning U-bends can be found. Neither reliable experimental data nor proven prediction methods are available. Indeed, flow structure, pressure drop and heat transfer in U-bends are an old unresolved design problem in the heat transfer industry. Thus, the present study aims at providing further insight on two-phase pressure drops and flows patterns in U-bends. Based on a new type of U-bend test section, an extensive experimental study was conducted. The experimental campaign covered five test sections with three internal diameters (7.8, 10.8 and 13.4 mm), five bend diameters (24.8, 31.7, 38.1, 54.8 and 66.1 mm), tested for three orientations (horizontal, vertical upflow and vertical downflow), two fluids (R134a and R410A), two saturation temperatures (5 and 10 °C) and mass velocities ranging from 150 to 1000 kg s -1 m -2 . The flow pattern observations identified were stratified-wavy, slug-stratified-wavy, intermittent, annular, dryout and mist flows. The effects of the U-bend on the flow patterns were also observed. A total of 5655 pressure drop data were measured at seven different locations in the test section ( straight tubes and U-bend) providing a total of almost 40,000 data points. The straight tube data were first used to improve the actual two-phase straight tube model of Moreno-Quibén and Thome. This updated model was then used to developed a two-phase U-bend pressure drop model. Based on a comparison between experimental and predicted values, it is concluded that the new two-phase frictional pressure drop model for U-bends successfully

Crack growth of throughwall flaw in Alloy 600 tube during leak testing

International Nuclear Information System (INIS)

Bahn, Chi Bum; Majumdar, Saurin

2015-01-01

Graphical abstract: - Highlights: • A series of leak testing was conducted at a constant pressure and room temperature. • The time-dependent increase in the leak rate was observed. • The fractography revealed slip offsets and crystallographic facets. • Time-dependent plasticity at the crack tip caused the slip offsets. • Fatigue by jet/structure interaction caused the crystallographic facets. - Abstract: We examined the issue of whether crack growth in a full thickness material can occur in a leaking crack. A series of leak tests was conducted at a room temperature and constant pressure (17.3 MPa) with Alloy 600 tube specimens containing a tight rectangular throughwall axial fatigue crack. To exclude a potential pulsation effect by a high pressure pump, the test water was pressurized by using high pressure nitrogen gas. Fractography showed that crack growth in the full thickness material can occur in the leaking crack by two mechanisms: time-dependent plasticity at the crack tip and fatigue induced by jet/structure interaction. The threshold leak rate at which the jet/structure interaction was triggered was between 1.3 and 3.3 L/min for the specific heat of the Alloy 600 tube tested

Development of NDT techniques for the inspection of WSGHWR pressure tubes

International Nuclear Information System (INIS)

Gray, B.S.; Highmore, P.J.; Rudlin, J.R.; Cooper, A.G.

1979-01-01

The fuel for the Steam Generating Heavy Water Reactor at Winfrith Heath is contained in vertical Zircaloy pressure tubes and is cooled by boiling light water. This paper describes the development of NDT techniques for the inservice examination of the pressure tubes to provide continuing assurance of the absence of axial crack-like defects. The resultant equipment has to operate in water-filled tubes in the presence of the radiation field due to the irradiated fuel elements in adjacent tubes. Also, a layer of surface oxide on the inside of the tubes has been found to significantly affect the behaviour of a prototype inspection device. To provide adequate sensitivity in these conditions, without the occurrence of unnecessary spurious indications, a combination of techniques has been developed. This involves the use of ultrasonics in both pulse-echo and 'pitch and catch' mode together with a single frequency eddy current technique. Laboratory work using artificial defects is described and also how the development programme was modified to accommodate the results of in-reactor tests using a prototype device. Reference is also made to the development of CCTV equipment to provide a supplementary visual examination. (author)

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

International Nuclear Information System (INIS)

Royen Van, E.

2011-11-01

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 T sat of 5 °C and 15 °C, mass flow rates from 4 to 35 kg/m 2 s and heat fluxes from 15 to 70 kW/m 2 , 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

Feeding tube - infants

Science.gov (United States)

... this page: //medlineplus.gov/ency/article/007235.htm Feeding tube - infants To use the sharing features on this page, please enable JavaScript. A feeding tube is a small, soft, plastic tube placed ...

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

International Nuclear Information System (INIS)

Urbanic, V.F.; Warr, B.D.; Manolescu, A.; Chow, C.K.; Shanahan, M.W.

1989-01-01

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/dm 2 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)

PPFAE preparation for pressure pipes production to CNE

International Nuclear Information System (INIS)

Azzinari, D; Carus, G; Sabio Calvete, M; Ilarri, M; Segato, M; Perazzo, M; Duarte; J; Duarte, A; Rojas, J; Fleitas, G; Szieber, W C; Bianchi, A D; Flores, A V; Bianchi, D R; Vizcaino, P

2012-01-01

The Life Extension Program for the PHWR Embalse power reactor of Argentina includes the replacement of de original set of 2,5% Nb pressure tubes by a new one It was also decided to made the new set de pressure tubes locally from imported Zr-2,5% Nb extruded tubes. Based on the degraded processes that develop in service a new set of metallurgical requirements was set by the Designer to be follow for new suppliers. As the mechanical properties of pressure tubes in operation depend on the micro-structural deformation, it is necessary to add to the normal macroscopic quality control same additional mechanical and microstructural studies, especially during the development stage. In the present paper the current status of delay hydrogen cracking, full width at half maximum (FWHM) of x-ray diffraction peaks, alpha phase thickness log-normal distributions and Kearns texture factors are summarized to estimate the magnitude of the micro-structural characteristics of pressure tubes obtained from imported Zr-2,5% Nb extruded tubes that were processes following designer specifications. The pressure tubes were obtained by two step of plastic deformation in a HPTR 60-120 tube rolling machine without intermediate annealing and a final thermal treatment (author)

Two and dimensional heat analysis inside a high pressure electrical discharge tube

International Nuclear Information System (INIS)

Aghanajafi, C.; Dehghani, A. R.; Fallah Abbasi, M.

2005-01-01

This article represents the heat transfer analysis for a horizontal high pressure mercury steam tube. To get a more realistic numerical simulation, heat radiation at different wavelength width bands, has been used besides convection and conduction heat transfer. The analysis for different gases with different pressure in two and three dimensional cases has been investigated and the results compared with empirical and semi empirical values. The effect of the environmental temperature on the arc tube temperature is also studied

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%.

Method of reactivity control in pressure tube reactor

International Nuclear Information System (INIS)

Fukumura, Nobuo.

1988-01-01

Purpose: To provide a method of controlling reactivity in a pressure tube reactor at high conversion ratio intended for high burn-up degree. Method: Control tubes are inserted in heavy water moderator. Light water is filled in the tubes at the initial burning stage. Along with the advance of the burning, the light water is gradually removed and replaced with gases of less reactive nuclear reactivity with neutrons such as air or gaseous carbon dioxide. The tubes are made of less neutron absorbing material such as aluminum. By filling light water, infinite multiplication factor is reduced to suppress the reactivity at the initial burning stage. As light water is gradually removed and replaced with air, etc., it provides an effect like that elimination of heavy water moderator to increase the conversion ratio. Accordingly, nuclear fission materials are produced additionally by so much to extend the burn-up degree. In this way, it can provide excellent effect in realizing high burn-up ratio and high conversion ratio. (Kamimura, M.)

Manufacture of thin-walled clad tubes by pressure welding of roll bonded sheets

Science.gov (United States)

Schmidt, Hans Christian; Grydin, Olexandr; Stolbchenko, Mykhailo; Homberg, Werner; Schaper, Mirko

2017-10-01

Clad tubes are commonly manufactured by fusion welding of roll bonded metal sheets or, mechanically, by hydroforming. In this work, a new approach towards the manufacture of thin-walled tubes with an outer diameter to wall thickness ratio of about 12 is investigated, involving the pressure welding of hot roll bonded aluminium-steel strips. By preparing non-welded edges during the roll bonding process, the strips can be zip-folded and (cold) pressure welded together. This process routine could be used to manufacture clad tubes in a continuous process. In order to investigate the process, sample tube sections with a wall thickness of 2.1 mm were manufactured by U-and O-bending from hot roll bonded aluminium-stainless steel strips. The forming and welding were carried out in a temperature range between RT and 400°C. It was found that, with the given geometry, a pressure weld is established at temperatures starting above 100°C. The tensile tests yield a maximum bond strength at 340°C. Micrograph images show a consistent weld of the aluminium layer over the whole tube section.

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

Time-dependent leak behavior of flawed Alloy 600 tube specimens at constant pressure

Energy Technology Data Exchange (ETDEWEB)

Bahn, Chi Bum, E-mail: bahn@anl.gov [Argonne National Laboratory, Argonne, IL 60439 (United States); Majumdar, Saurin [Argonne National Laboratory, Argonne, IL 60439 (United States); Harris, Charles [United States Nuclear Regulatory Commission, Rockville, MD 20852 (United States)

2011-10-15

Leak rate testing has been performed using Alloy 600 tube specimens with throughwall flaws. Some specimens have shown time-dependent leak behavior at constant pressure conditions. Fractographic characterization was performed to identify the time-dependent crack growth mechanism. The fracture surface of the specimens showed the typical features of ductile fracture, as well as the distinct crystallographic facets, typical of fatigue crack growth at low {Delta}K level. Structural vibration appears to have been caused by the oscillation of pressure, induced by a high-pressure pump used in a test facility, and by the water jet/tube structure interaction. Analyses of the leak behaviors and crack growth indicated that both the high-pressure pump and the water jet could significantly contribute to fatigue crack growth. To determine whether the fatigue crack growth during the leak testing can occur solely by the water jet effect, leak rate tests at constant pressure without the high-pressure pump need to be performed. - Highlights: > Leak rate of flawed Alloy 600 tubing increased at constant pressure condition. > Fractography revealed two cases: ductile tearing and crystallographic facets. > Crystallographic facets are typical features of fatigue crack growth at low {Delta}K. > Fatigue source could be water jet-induced vibration and/or high-pressure pump pulsation.

An assessment of the waterside corrosion and hydrogen pick-up in the zircaloy-2 pressure tubes of PHWR

International Nuclear Information System (INIS)

Sah, D.N.

1992-01-01

In view of the deleterious effect of hydriding on the operating life of zircaloy-2 pressure tubes in PHWRs there is an urgent need for the assessment of the status of the pressure tubes with respect to corrosion and hydrogen pick-up in the operating PHWRs. A model has been developed for analysing the waterside corrosion and hydrogen pick-up in the zircaloy-2 pressure tubes under reactor operating conditions. This model predicts the axial profiles of oxide layer thickness and hydrogen pick-up in the pressure tubes as a function of the operating time of the reactor. The prediction of hydrogen pick-up by the model in the F-10 pressure tube of RAPS-I have been found to be in good agreement with the measured value of hydrogen content. This report gives a brief description of the model and its predictions on the present status of hydrogen pick-up in the pressure tubes of lead reactor RAPS-II. (author). 6 refs., 5 figs., 2 tabs

Prediction of pressure tube fretting-wear damage due to fuel vibration

Energy Technology Data Exchange (ETDEWEB)

Yetisir, M; Fisher, N J [Atomic Energy of Canada Ltd., Chalk River, ON (Canada)

1996-12-31

Fretting marks between fuel bundle bearing pads and pressure tubes have been observed at the inlet end of some Darlington NGS (nuclear generating station) and Bruce NGS fuel channels. The excitation mechanisms that lead to fretting are not fully understood. In this paper, the possibility of bearing pad-to-pressure tube fretting due to turbulence-induced motion of the fuel element is investigated. Numerical simulations indicate that this mechanism by itself is not likely to cause the level of fretting experienced in Darlington and Bruce NGS`s (nuclear generating stations). (author). 12 refs., 2 tabs., 11 figs.

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.

Study on frictional pressure drop of steam-water two phase flow in optimized four-head internal-ribbed tube

International Nuclear Information System (INIS)

Wang Weishu; Zhu Xiaojing; Bi Qincheng; Wu Gang; Yu Shuiqing

2012-01-01

The optimized internal-ribbed tube is different from the normal internal-ribbed tube on the frictional pressure drop characteristics. The frictional pressure drop characteristics of steam-water two phase flow in horizontal four-head optimized internal-ribbed were studied under adiabatic condition. According to the experimental and calculation results, the two-phase multiplier is greatly affected by the steam quality and pressure. The two-phase multiplier increases with increasing quality, and decreases with increasing pressure. In the near-critical pressure region, the two-phase multiplier is close to 1. The frictional pressure drop of two phase flow in optimized tube is less than that in the normal tube under the same work condition. The good hydrodynamic condition could be achieved when the optimized internal-ribbed tube is used in the heat transfer equipment because the self-compensating characteristics exist due to the reduction of frictional pressure drop. (authors)

Final report on development evaluation of Task Group 3 pressure tubes

International Nuclear Information System (INIS)

Fleck, R.G.; Price, E.G.; Cheadle, B.A.

1983-11-01

This report describes the production and evaluation of pressure tubes manufactured to the recommendations of Task Group 3 (TG3) of the Creep Engineering Design Plan. The Zr-2.5 wt percent Nb tubes were manufactured by modified production route to change their metallurgical structure and so reduce the in-service elongation rates. Three modified routes were investigated and a total of twenty-eight tubes produced. There were no difficulties in manufacture and the tubes satisfied the quality assurance and design specifications of reactor grade tubes. Metallurgical evaluation showed that the expected changes in microstructure had occurred but not to the extent anticipated. The TG3 tubes were found to have comparable properties to current tubes when tested for: tensile strength (irradiated and unirradiated); hydride cracking; stress to reorient hydrides; hydrogen diffusion; flaw tolerance; corrosion (irradiated and unirradiated); wear; rolled joint characteristics; irradiation creep and growth. Lower in-service elongation rates are expected for tubes produced by two of the modified routes

A technique of taking samples from inside the pressure tube along the axis of the fuel channel

International Nuclear Information System (INIS)

Gyongyosi, T.

2005-01-01

Full text: The ageing process through its complex mechanisms affects in time more or less the component parts, the systems and the structures of the nuclear power plant. For CANDU type nuclear power plant the main component part in operation is the pressure tube, made from Zr - 2,5% Nb alloy, used in extreme hard operation conditions (static and dynamic loading from high pressure and temperature and high neutron flux). The pressure tube endures, in time, changing of the material and of the geometry. Additionally, the excessive hydrogen uptake initiates the Delayed Hydride Cracking (DHC) mechanism on the pressure tubes. For checking the evolution of the hydride / deuteration phenomena in the material of the pressure tube and especially to establish the real lifetime as compared to design lifetime it is useful to initiate, develop and apply a technology and a complex equipment for taking samples directly from inside the pressure tube, this enabling the determination of the hydrogen content. In the paper are showed briefly: - the evolution in time of the techniques for axial taking of the samples from inside of the pressure tube used in the CANDU 6 NPP; - the reasons that determined us to develop one of these technologies; - the technological facilities needed to apply it. (author)

A technique of taking samples from inside the pressure tube along the axis of the fuel channel

International Nuclear Information System (INIS)

Gyongyosi, T.

2005-01-01

The ageing process through its complex mechanisms affects in time more or less the component parts, the systems and the structures of the nuclear power plant. For CANDU type nuclear power plant the main component part in operation is the pressure tube, made from Zr - 2,5% Nb alloy, used in extreme hard operation conditions (static and dynamic loading from high pressure and temperature and high neutron flux). The pressure tube endures, in time, changing of the material and of the geometry. Additionally, the excessive hydrogen uptake initiates the Delayed Hydride Cracking (DHC) mechanism on the pressure tubes. For checking the evolution of the hydride/deuteration phenomena in the material of the pressure tube and especially to establish the real lifetime as compared to design lifetime it is useful to initiate, develop and apply a technology and a complex equipment for taking samples directly from inside the pressure tube, this enabling the determination of the hydrogen content. In the paper are showed briefly: - the evolution in time of the techniques for axial taking of the samples from inside of the pressure tube used in the CANDU 6 NPP; - the reasons that determined us to develop one of these technologies; - the technological facilities needed to apply it. (author)

Inelastic analysis of finite length and depth cracked tubes

International Nuclear Information System (INIS)

Reich, M.; Prachuktam, S.; Gardner, D.

1977-01-01

Steam generator tube failure can at times result in reactor safety problems and subsequent premature reactor shutdowns. Typical PWR steam generator units contain thousands of long straight tubes with U-bend sections. These tubes are primarily made from alloy 600 and their sizes vary between 3 / 4 '' and 7 / 8 '' (1.905 cm and 2.223 cm) in diameter with nominal thicknesses of 0.043'' to 0.053'' (0.109 cm to 0.135 cm). Since alloy 600 (and the previously used 304-SS tubes) are ductile, high toughness materials LEFM (linear elastic fracture mechanics) criteria do not apply. This paper concerns itself with the prediction of the failure pressures for typical PWR steam generator tubes with longitudinal finite length and finite depth cracks. Only local plastic overload failure is considered

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.

Detection of steam generator tube leaks in pressurized water reactors

International Nuclear Information System (INIS)

Roach, W.H.

1985-01-01

This report addresses the early detection of small steam generator tube leaks in pressurized water reactors. It discusses the third, and final, year's work on an NRC-funded project examining diagnostic instrumentation in water reactors. The first two years were broad in coverage, concentrating on anticipatory measurements for detection of potential problems in both pressurized- and boiling-water reactors, with recommendations for areas of further study. One of these areas, the early detection of small steam tube leaks in PWRs, formed the basis of study for the last year of the project. Four tasks are addressed in this study of the detection of steam tube leaks. (1) Determination of which physical parameters indicate the onset of steam generator tube leaks. (2) Establishing performance goals for diagnostic instruments which could be used for early detection of steam generator tube leaks. (3) Defining the diagnostic instrumentation and their location which satisfy Items 1 and 2 above. (4) Assessing the need for diagnostic data processing and display. Parameters are identified, performance goals established, and sensor types and locations are specified in the report, with emphasis on the use of existing instrumentation with a minimum of retrofitting. A simple algorithm is developed which yields the leak rate as a function of known or measurable quantities. The conclusion is that leak rates of less than one-tenth gram per second should be detectable with existing instrumentation. (orig./HP)

High pressure thimble/guide tube seal fitting with built-in low pressure seal especially suitable for facilitated and more efficient nuclear reactor refueling service

International Nuclear Information System (INIS)

Bhatt, P.N.; Blaushield, R.M.

1991-01-01

This patent describes a HP/LP seal arrangement for an elongated guide tube and an elongated thimble disposed therein. The guide tube and thimble extending outwardly from the core of a nuclear reactor to a seal table where the guide tube is welded to the seal table to provide a high pressure seal relative thereto. It comprises: a tubular seal fitting disposed in alignment with the guide tube with the thimble extending therethrough on the low pressure side of the seal table; first high pressure sealing means coupling one end of the fitting to an end of the guide tube to prevent leakage from within the guide tube; inwardly facing thread means disposed adjacent the other and outer end of the seal fitting; a nut having an opening through which the thimble extends and further having outwardly facing threading in mating engagement with the fitting thread means; the fitting having a seal seat spaced longitudinally inwardly from the thread means and facing the fitting outer end and further disposed annularly about the inner surface of the fitting; deformable ring seal means; second releasable high pressure sealing means coupling the thimble to the outer end portion of the guide tube

Fracture Toughness Round Robin Test International in pressure tube materials

International Nuclear Information System (INIS)

Villagarcia, M.P.; Liendo, M.F.

1993-01-01

Part of the pressure tubes surveillance program of CANDU type reactors is to determine the fracture toughness using a special fracture specimen and test procedure. Atomic Energy of Canada Limited decided to hold a Round Robin Test International and 9 laboratories participated worldwide in which several pressure tube materials were selected: Zircaloy-2, Zr-2.5%Nb cold worked and Zr-2.5%Nb heat treated. The small specimens used held back the thickness and curvature of the tube. J-R curves at room temperature were obtained and the crack extension values were determined by electrical potential drop techniques. These values were compared with results generated from other laboratories and a bid scatter was founded. It could be due to slight variations in the test method or inhomogeneity of the materials and a statistical study must be done to see if there is any pattern. The next step for the Round Robin Test would be to make some modifications in the test method in order to reduce the scatter. (Author)

Elastic-plastic failure analysis of pressure burst tests of thin toroidal shells

International Nuclear Information System (INIS)

Jones, D.P.; Holliday, J.E.; Larson, L.D.

1998-07-01

This paper provides a comparison between test and analysis results for bursting of thin toroidal shells. Testing was done by pressurizing two toroidal shells until failure by bursting. An analytical criterion for bursting is developed based on good agreement between structural instability predicted by large strain-large displacement elastic-plastic finite element analysis and observed burst pressure obtained from test. The failures were characterized by loss of local stability of the membrane section of the shells consistent with the predictions from the finite element analysis. Good agreement between measured and predicted burst pressure suggests that incipient structural instability as calculated by an elastic-plastic finite element analysis is a reasonable way to calculate the bursting pressure of thin membrane structures

Correlations of CO2 at supercritical pressures in a vertical circular tube

International Nuclear Information System (INIS)

Li Zhihui; Jiang Peixue

2010-01-01

The experiment results of convection heat transfer of CO 2 at supercritical pressures in a 2 mm diameter vertical circular tube for upward flow and downward flow were analyzed for pressures ranging from 78 to 95 bar, inlet temperatures from to 25 to 40 degree C, and inlet Re numbers from 3000 to 20000. The results were compared with some well known empirical correlations for the heat transfer without buoyancy effects and the heat transfer with strong buoyancy effects. It is found that there is a big deviation between the experiment results and empirical correlations. Based on the experiment data, correlations are developed for the local Nusselt correlations of CO 2 at supercritical pressures in vertical circular tubes.(authors)

Predicting crack instability behavior of burst tests from small specimens for irradiated Zr-2.5Nb pressure tubes

International Nuclear Information System (INIS)

Davies, P.H.

1997-01-01

A scaling approach, based on the deformation J-integral at maximum load obtained from small specimens, is proposed for predicting the crack instability behavior of burst tests on irradiated Zr-2.5Nb pressure tubes. An assessment of this approach is carried out by comparison with other toughness criteria such as the modified J-integral and the plastic work dissipation rate approach. The largest discrepancy between the different parameters occurs for materials of intermediate toughness which exhibit the most stable crack growth and tunnelling up to maximum load. A study of one material of intermediate toughness suggests crack-front tunnelling has a significant influence on the results obtained from the 17-mm-wide specimens. It is shown that for a tube of intermediate toughness the different approaches can significantly underpredict the extent of stable crack growth before instability in a burst test even after correcting for tunnelling. The usefulness of a scaling approach in reducing the discrepancy between the small- and large-scale specimen results for this material is demonstrated

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%.

The Sensitivity Analysis of Axial Pressure Tube Creep Profile for Dryout Power in PHWR

Energy Technology Data Exchange (ETDEWEB)

Ryu, Euiseung; Kim, Youngae [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

The Stern Laboratory performed the CHF tests with only one axial pressure tube creep profile per 3.3%, 5.1% peak crept channel and made CHF correlation including creep factor from the CHF test results. Wolsong nuclear power plants also have utilized the same CHF correlation derived by CNL. Pressure tube diameter creep rate is function of fast neutron, coolant temperature, and coolant pressure in a channel. It means that various axial pressure tube creep profiles exist in PHWR due to the history of operating conditions. Usually, CHF correlation is used during ROP(Regional Overpower Protection) Trip Setpoint Analysis or Safety Analysis in PHWR. The sensitivity analysis for CHF effects using various creep profiles is needed. This paper summarizes the comparison results of dryout power between CHF test creep profile and estimated creep profiles of Wolsong units. The effect of axial pressure tube creep profile for dryout power in fuel channel is evaluated by using Stern Lab. CHF test creep profile and 380 channel creep profiles of Wolsong. The dryout powers at 3.3% and 5.1% test conditions are slightly smaller when using 380 Wolsong channels creep profiles. These also show that the simulated dryout powers maintain consistency regardless of flow conditions.

Flow instability research on steam generator with straight double-walled heat transfer tube for FBR. Pressure drop under high pressure condition

International Nuclear Information System (INIS)

Liu, Wei; Tamai, Hidesada; Yoshida, Hiroyuki; Takase, Kazuyuki; Hayafune, Hiroki; Futagami, Satoshi; Kisohara, Naoyuki

2008-01-01

For the Steam Generator (SG) with straight double-walled heat transfer tube that used in sodium cooled Faster Breeder Reactor, flow instability is one of the most important items need researching. As the first step of the research, thermal hydraulics experiments were performed under high pressure condition in JAEA with using a straight tube. Pressure drop, heat transfer coefficients and void fraction data were derived. This paper evaluates the pressure drop data with TRAC-BF1 code. The Pffan's correlation for single phase flow and the Martinelli-Nelson's two-phase flow multiplier are found can be well predicted the present pressure drop data under high pressure condition. (author)

Burst pressure and leak rate from fretted SG tubes

International Nuclear Information System (INIS)

Hwang, Seong Sik; Jung, Man Kyo; Kim, Hong Pyo; Kim, Joung Soo

2005-01-01

Steam generator(SG) tubes of a pressurized water reactor(PWR) have suffered from various types of corrosion, such as pitting, wastage and stress corrosion cracking (SCC) on both the primary and secondary side. Recently, fretting/wear degradation at the tube support region has been reported in some Korean nuclear power plants. In order to prevent the primary coolant from leaking to the secondary side, the tubes are repaired by a sleeving or plugging. It is important to establish the repair criteria to assure a reactor integrity and yet maintain the plugging ratio within the limits needed for an efficient operation. The objective of the burst test is to obtain a relationship between the burst/leak rate and the shape of the fretted flaws machined with an electro discharge machining (EDM)

Brittle-fracture potential of irradiated Zircaloy-2 pressure tubes

Science.gov (United States)

Huang, F. H.

1993-12-01

Neutron irradiation can degrade the fracture toughness of Zircaloy-2 and may cause highly irradiated reactor components of this material to fail in a brittle manner. The effects of radiation embrittlement on the structural integrity of N Reactor pressure tubes are studied by performing KIc and JIc fracture toughness testing on samples cut from the Zircaloy-2 tubes periodically removed from the reactor. A fluence of 6 × 10 25n/ m2 ( E > 1 MeV) reduced the fracture toughness of the material by 40 to 50%. The fracture toughness values appear to saturate at 260°C with fluences above 3 × 10 25n/ m2 ( E > 1 MeV), but continue to decline with increasing fluence at temperatures below 177°C. Present and previous results obtained from irradiated pressure tubes indicate that the brittle-fracture potential of Zircaloy-2 increases with decreasing temperature and increasing fluence. Fractographic examinations of the fracture surfaces of irradiated samples reveal that circumferential hydride formation significantly influenced fracture morphology by providing sites for easy crack nucleation and leaving deep cracks. However, the deep cracks created at the hydride platelets in specimens containing less than 220 ppm hydrogen are not believed to be the major cause of degradation in postirradiation fracture toughness.

Heat transfer test in a tube using CO2 at supercritical pressures

International Nuclear Information System (INIS)

Kim, Hwan Yeol; Kim, Hyungrae; Song, Jin Ho; Cho, Bong Hyun; Bae, Yoon Yeong

2005-01-01

Heat transfer test facility, which is named as SPHINX (Supercritical Pressure Heat Transfer Investigation for NeXt Generation), has been constructed in KAERI for the study of heat transfer and pressure drop characteristics in a single tube, single rod and rod bundle at supercritical CO 2 conditions. The tests with supercritical water are difficult it terms of cost and effort, since the critical pressure and temperature of water are as high as 22.12 MPa and 374.14degC. As a substitute for water, CO 2 is selected for the test since the critical pressure and temperature of CO 2 are 7.38 MPa and 31.05degC that are much lower than those of water. This paper describes the design characteristics of the SPHINX and the experimental investigations on the heat transfer and pressure drop of a vertical single tube with an inside diameter of 4.4 mm with upward flow of supercritical CO 2 . The geometry of the single tube is the same as that of Kyushu University test performed with Freon (R22) for the direct comparison of a medium effect. The tests were performed with various heat and mass fluxes at a given pressure. The range of mass flux is 400∼1200 kg/m 2 s and the heat flux is chosen up to 150 kW/m 2 . The selected pressure are 7.75, 8.12, and 8.85 MPa. The test results are investigated and compared with the previous tests. (author)

Apparatus for inspecting and repairing a pressurized-water reactor's steam generator heat exchanger tubes

International Nuclear Information System (INIS)

Mueller, O.; Roettger, H.; Kasti, H.; Hagen, H.G.

1976-01-01

Described is an apparatus provided for use with a pressurized-water reactor' steam generator having a manifold chamber enclosing the bottom side of a horizontal tube sheet having holes therethrough in which are mounted the tubes of a heat exchanger tube bundle. The manifold chamber has a manhole giving access to the tube's bottom side to permit internal inspection or repair of the tubes by registration of an end of a flexible guide conduit with the tube sheet holes and through which a flexible carrier can be guided for insertion via these holes in the tube sheet and through the tubes extending from the tube sheet's other side

Fuel rod bundles proposed for advanced pressure tube nuclear reactors

International Nuclear Information System (INIS)

Prodea, Iosif; Catana, Alexandru

2010-01-01

The paper aims to be a general presentation for fuel bundles to be used in Advanced Pressure Tube Nuclear Reactors (APTNR). The characteristics of such a nuclear reactor resemble those of known advanced pressure tube nuclear reactors like: Advanced CANDU Reactor (ACR TM -1000, pertaining to AECL) and Indian Advanced Heavy Water Reactor (AHWR). We have also developed a fuel bundle proposal which will be referred as ASEU-43 (Advanced Slightly Enriched Uranium with 43 rods). The ASEU-43 main design along with a few neutronic and thermalhydraulic characteristics are presented in the paper versus similar ones from INR Pitesti SEU-43 and CANDU-37 standard fuel bundles. General remarks regarding the advantages of each fuel bundle and their suitability to be burned in an APTNR reactor are also revealed. (authors)

Burst strength of tubing and casing based on twin shear unified strength theory.

Science.gov (United States)

Lin, Yuanhua; Deng, Kuanhai; Sun, Yongxing; Zeng, Dezhi; Liu, Wanying; Kong, Xiangwei; Singh, Ambrish

2014-01-01

The internal pressure strength of tubing and casing often cannot satisfy the design requirements in high pressure, high temperature and high H2S gas wells. Also, the practical safety coefficient of some wells is lower than the design standard according to the current API 5C3 standard, which brings some perplexity to the design. The ISO 10400: 2007 provides the model which can calculate the burst strength of tubing and casing better than API 5C3 standard, but the calculation accuracy is not desirable because about 50 percent predictive values are remarkably higher than real burst values. So, for the sake of improving strength design of tubing and casing, this paper deduces the plastic limit pressure of tubing and casing under internal pressure by applying the twin shear unified strength theory. According to the research of the influence rule of yield-to-tensile strength ratio and mechanical properties on the burst strength of tubing and casing, the more precise calculation model of tubing-casing's burst strength has been established with material hardening and intermediate principal stress. Numerical and experimental comparisons show that the new burst strength model is much closer to the real burst values than that of other models. The research results provide an important reference to optimize the tubing and casing design of deep and ultra-deep wells.

Benchmark of WIMS-IST against MCNP for CANDU pressure tube fast fluxes

International Nuclear Information System (INIS)

Donders, R.E.; Douglas, S.R.

2002-01-01

Pressure tube fast-flux data in CANDU are currently calculated using the multi-group neutron transport code WIMS-IST. In this study, the WIMS-IST fast flux calculations are benchmarked against MCNP calculations (a Monte Carlo particle transport code), over the range of fuel burnup and coolant density in CANDU. The comparison shows good agreement between WIMS and MCNP, with WIMS fast fluxes being 1.5% to 4% lower than the MCNP values. The difference is smallest for fresh fuel, and increases with burnup. The fast flux gradient across the pressure tube (factor of 1.23 from inner edge to outer edge) is accurately calculated by WIMS. When reporting fast fluxes in pressure tubes, these are generally given as >1.000 MeV fluxes. For WIMS, this requires an extra conversion step, since the WIMS ENDF/B libraries do not have a group boundary at 1 MeV. The conversion step is based on a fictitious isotope ONEMEV in the WIMS nuclear data library. The conversion factor in WIMS was found to be about one percent too high. When providing >1 MeV fluxes from WIMS, this partially compensates for the slight under prediction of the fast flux. Pressure tube >1 MeV fluxes from WIMS are therefore 0.5% to 3% lower than MCNP values. To obtain accurate fast flux data, neutron transport calculations must be performed on a critical cell. For this study, all calculations were performed with radial albedo boundary conditions giving a critical cell. This required the use of an albedo version of MCNP, developed at AECL. (author)

Factors influencing leaching of PBDEs from waste cathode ray tube plastic housings.

Science.gov (United States)

Stubbings, William A; Harrad, Stuart

2016-11-01

Samples of waste cathode ray tube (CRT) plastic housings were exposed to Milli-Q® water containing dissolved humic matter at concentrations of 0, 100 and 1000mgL(-1) as leaching fluid under laboratory conditions, and polybrominated diphenyl ethers (PBDEs) determined in the resulting leachate. Despite the relatively hydrophobic physicochemical properties of PBDEs, concentrations of ƩPBDEs in the leachate from the leaching experiments in this study ranged from 14,000 to 200,000ngL(-1). PBDE leaching appears to be a second order process, whereby a period of initially intense dissolution of more labile PBDEs is followed by a slower stage corresponding to external diffusion of the soluble residue in the material. The bulk of transfer of PBDEs to the leaching fluid occurs within the first 6h of contact, during which time we suggest that the most labile PBDEs are "washed" off the surface of the CRT plastics. The predominant congeners in the chips were BDE-209 (2600mgkg(-1)) and BDE 183 (220mgkg(-1)). The impacts on PBDE leaching of leachate pH and temperature were also examined. Increasing the temperature of leaching fluids from 20 to 80°C, enhances the leachability of BDE-209 and BDE-99 from plastics. In all cases, the alkaline pH8.5 examined, resulted in the greatest PBDE concentrations in leachate. Agitation of the waste/leachate mixture enhances PBDE leaching from CRT plastics. Potential evidence for debromination of heavy congeners to the lower brominated and more bioavailable BDEs was observed. Specifically, BDEs-47, -85 and -100 were detected in the leachates, but were absent from the CRT plastics themselves. Copyright © 2016. Published by Elsevier B.V.

The development of an auto-sealing system using an electrically shrinkable tube under a low-pressure condition

Energy Technology Data Exchange (ETDEWEB)

Okano, Yoshihiro; Kitagawa, Takao [NKK Corp, Tsu, Mie (Japan); Shoji, Norio [NKK Corp., Yokohama (Japan); Namioka, Toshiyuki [Nippon Kokan Koji Corp., Yokohama (Japan). Research and Development Dept.; Komura, Minoru [Nitto Denko Corp., Fukaya, Saitama (Japan)

1997-04-01

This article describes the development of a system to create high quality, automatic sealing of field joints of polyethylene coated pipelines. The system uses a combination of an electrically heated shrinkable tube and a low-pressure chamber. The self-heating shrinkable tube includes electric heater wires that heat when connected to electricity. A method was developed to eliminate air trapped between the tube and the steel pipe by shrinking the tube under a low-pressure condition. The low-pressure condition was automatic and easily attained by using a vacuum chamber. It was verified that the system produced high quality sealing of the field joints.

Hydrogen concentration determination in pressure tube samples using differential scanning calorimetry (dsc)

International Nuclear Information System (INIS)

Marinescu, R.; Mincu, M.

2015-01-01

Zirconium alloys are widely used as a structural material in nuclear reactors. It is known that zirconium based cladding alloys absorb hydrogen as a result of service in a pressurized water reactor. Hydrogen absorbed (during operation of the reactor) in the zirconium alloy, out of which the pressure tube is made, is one of the major factors determining the life time of the pressure tube. For monitoring the hydrides, samples of the pressure tube are periodically taken and analyzed. At normal reactor operating temperature, hydrogen has limited solubility in the zirconium lattice and precipitates out of solid solution as zirconium hydride when the solid solubility is exceeded. As a consequences material characterization of Zr-2.5Nb CANDU pressure tubes is required after manufacturing but also during the operation to assess its structural integrity and to predict its behavior until the next in-service inspection. Hydrogen and deuterium concentration determination is one of the most important parameters to be evaluated during the experimental tests. Hydrogen present in zirconium alloys has a strong effect of weakening. Following the zirconium-hydrogen reaction, the resulting zirconium hydride precipitates in the mass of material. Weakening of the material, due to the presence of 10 ppm of precipitated hydrogen significantly affects some of its properties. The concentration of hydrogen in a sample can be determined by several methods, one of them being the differential scanning calorimetry (DSC). The principle of the method consists in measuring the difference between the amount of heat required to raise the temperature of a sample and a reference to a certain value. The experiments were made using a TA Instruments DSC Q2000 calorimeter. This paper contains experimental work for hydrogen concentration determination by Differential Scanning Calorimetry (DSC) method. Also, the reproducibility and accuracy of the method used at INR Pitesti are presented. (authors)

Development of heat treated Zr-2.5 Wt% Nb pressure tube and its microstructural characterization using electron microscopy techniques

International Nuclear Information System (INIS)

Saibaba, N.

2010-01-01

Two phase Zr-2.5 wt % Nb alloy is widely used for manufacture of pressure tubes for pressurized heavy water reactors (PHWRs). These tubes are used in cold worked and stress relieved (CWSRs) condition and are manufactured by cold drawing or pilgering routes. The microstructure of the CWSR tube is characterized with presence of discontinuous β phase stringers sandwiched between elongated α-phase. Pressure tube undergoes dimensional changes and micro structural deterioration under the reactor operating conditions of temperature, pressure and neutron flux. This limits the life of the component and the availability of the power reactors. There is renewed interest in increasing the life of the pressure tube by bringing about a change in the microstructure of Zr-2.5 Nb material using various thermo mechanical processes during its manufacturing. Heat treatment of this two-phase alloy has been understood to uniquely stabilize the microstructure, which prevents degradation, under in-reactor service condition. This paper illustrates various heat treatment cycles carried out at intermediate cold working stage. Heat treatment involves solutionization of the Zr-2.5 wt % Nb tube from different temperatures followed by two types of quenching process viz, gas quenching and water quenching. The OIM-TEM studies were carried out for characterization of final tube. The technique confirmed the presence of β-phase relatively enriched in Nb content. The resulting SEM microstructures after ageing treatment at different soaking temperatures and time have been presented. Mechanical properties of heat treated pressure tubes, both at room temperature and elevated temperature have been compared with conventional CWSR pressure tube used in PHWRs. (author)

Numerical study on turbulent heat transfer and pressure drop of nanofluid in coiled tube-in-tube heat exchangers

International Nuclear Information System (INIS)

Aly, Wael I.A.

2014-01-01

Highlights: • The performance of helically coiled tube heat exchanger using nanofluid is modeled. • The 3D turbulent flow and conjugate heat transfer of CTITHE are solved using FVM. • The effects of nanoparticle concentration and curvature ratio are investigated. • The Gnielinski correlation for Nu for turbulent flow in helical tubes can be used for water-based Al 2 O 3 nanofluid. - Abstract: A computational fluid dynamics (CFD) study has been carried out to study the heat transfer and pressure drop characteristics of water-based Al 2 O 3 nanofluid flowing inside coiled tube-in-tube heat exchangers. The 3D realizable k–ε turbulent model with enhanced wall treatment was used. Temperature dependent thermophysical properties of nanofluid and water were used and heat exchangers were analyzed considering conjugate heat transfer from hot fluid in the inner-coiled tube to cold fluid in the annulus region. The overall performance of the tested heat exchangers was assessed based on the thermo-hydrodynamic performance index. Design parameters were in the range of; nanoparticles volume concentrations 0.5%, 1.0% and 2.0%, coil diameters 0.18, 0.24 and 0.30 m, inner tube and annulus sides flow rates from 2 to 5 LPM and 10 to 25 LPM, respectively. Nanofluid flows inside inner tube side or annular side. The results obtained showed a different behavior depending on the parameter selected for the comparison with the base fluid. Moreover, when compared at the same Re or Dn, the heat transfer coefficient increases by increasing the coil diameter and nanoparticles volume concentration. Also, the friction factor increases with the increase in curvature ratio and pressure drop penalty is negligible with increasing the nanoparticles volume concentration. Conventional correlations for predicting average heat transfer and friction factor in turbulent flow regime such as Gnielinski correlation and Mishra and Gupta correlation, respectively, for helical tubes are also valid for

Study on transport safety of fresh MOX fuel. Performance of the cladding tube of fresh MOX fuel against external water pressure

International Nuclear Information System (INIS)

Ito, Chihiro

1999-01-01

It is important to know the ability of the cladding tube for fresh MOX fuel against external water pressure when they were hypothetically sunk into the sea for unknown reasons. In order to evaluate the ability of cladding tubes for MOX fresh fuel against external water pressure, external water pressure tests were carried out. Resistible limit of cladding tubes against external water pressure is defined when cladding tubes are deformed largely due to buckling etc. The test results show cladding tube of BWR type can resist an external water pressure of 69 MPa (a depth of water of 7,000 m) and that of PWR type fuel can resist an external water pressure of 54 MPa (a depth of water of 5,500 m). Moreover, leak tightness is maintained at an external water pressure of 73 MPa (a depth of water of 7,400 m) for BWR type cladding tubes and at an external water pressure of 98 MPa (a depth of water of 10,000 m) for PWR type cladding tubes. (author)

Endotracheal tube cuff pressures during general anaesthesia while using air versus a 50% mixture of nitrous oxide and oxygen as inflating agents

Directory of Open Access Journals (Sweden)

Jesni Joseph Manissery

2007-01-01

Full Text Available The present study was aimed at assessing the efficacy of filling a 50% mixture of nitrous oxide : oxygen (50%N 2 O:O 2 in the endotracheal tube cuff to provide stable cuff pressures during general anaesthesia with 67%N 2 O. The endotracheal tube cuff pressures with air (control as the inflating agent in the tubes were found to have a total mean pressure of 62.60±12.33 at the end of one hour of general anaesthesia. When comparing the endotracheal tube cuff pressures in the Mallinckrodt tubes with that of the Portex tubes, with air as the inflating agent, the Portex tubes showed a significantly lower cuff pressures at the end of one hour. The endotracheal tube cuff pressures with 50%N 2 O:O 2 as the inflating agent showed a total mean pressure of 27.63 ± 3.221 at the end of one hour of general anaesthesia. This indicates that inflation of the cuff of the endotracheal tubes with a 50%N 2 O:O 2 rather than air maintains a stable intra cuff pressure. Therefore, the method of using a 50%N 2 O:O 2 for filling endotracheal tube cuff can be adopted for endotracheal tubes with high-volume, low-pressure cuffs to prevent both excessive cuff pressure and disruption of cuff seal, during general anaesthesia lasting up to one hour.

Characterization of plastic strains and crystallographic properties surrounding defects in steam generator tubes by orientation imaging microscopy

International Nuclear Information System (INIS)

Lehockey, E.M.; Brennenstuhl, A.M.

2002-01-01

Orientation Imaging Microscopy (OIM) has become a valuable technique for characterizing grain boundary structure, texture, and grain size distribution, which govern material susceptibility to degenerative effects (e.g. IGSCC). Methods recently developed, by Kinectrics, have extended OIM capabilities toward mapping and quantifying residual plastic strains in materials. OIM is applied in the present work to characterize the distribution of plastic strains, that accumulate in CANDU steam generator tubing during installation and service potentially undermining the performance, reliability, and fitness-for-service of these components. Plastic strain that evolves in response to roller-expansion was characterized in simulated roll joints constructed from Alloy 600 tubing. Results underscore the effect of over-rolling in generating intense gradients with broad variations in strain that extend significant distances through the wall thickness. Of greater relevance is the orientation of these gradients in the transverse direction, relative to the tube axis and potential for the development of abnormal grain growth during post-expansion heat treatments. The magnitude and distribution of strain measured by OIM are remarkably consistent with Finite Element Analysis (FEA) predictions offering compelling evidence as to the reliability of the OIM technique. OIM offers superior resolution than can be practically achieved with FEA having particular relevance in identifying highly localized concentrations of strain surrounding metallurgical defects that can serve as precursors to stress-related degenerative effects (e.g. IGSCC). The spatial distribution of residual plastic strain was also characterized within the context of localized texture, and grain size morphology surrounding (OD) 'pits' and indentations found in ex-service Monel 400 and Alloy 800 SG tubes, respectively. An absence of strain surrounding these surface defects suggests their propensity for promoting more deleterious

Plastic strain accumulation during asymmetric cyclic loading of Zircaloy-2 at room temperature

International Nuclear Information System (INIS)

Rajpurohit, R.S.; Santhi Srinivas, N.C.; Singh, Vakil

2016-01-01

Asymmetric cyclic loading leads to accumulation of cyclic plastic strain and reduces the fatigue life of components. This phenomenon is known as ratcheting fatigue. Zircaloy-2 is a important structural material in nuclear reactors and used as pressure tubes and fuel cladding in pressurized light and heavy water nuclear reactors. Due to power fluctuations, these components experience plastic strain cycles in the reactor and their life is reduced due to strain cycles. Power fluctuations also cause asymmetric straining of the material and leads to accumulation of plastic strain. The present investigation deals with the effect of the magnitude of mean stress, stress amplitude and stress rate on hardening/softening behavior of Zircaloy-2 under asymmetric cyclic loading, at room temperature. It was observed that plastic strain accumulation increased with mean stress and stress amplitude; however, it decreased with stress rate. (author)

Closed-form plastic collapse loads of pipe bends under combined pressure and in-plane bending

International Nuclear Information System (INIS)

Oh, Chang Sik; Kim, Yun Jae

2006-01-01

Based on three-dimensional (3-D) FE limit analyses, this paper provides plastic limit, collapse and instability load solutions for pipe bends under combined pressure and in-plane bending. The plastic limit loads are determined from FE limit analyses based on elastic-perfectly plastic materials using the small geometry change option, and the FE limit analyses using the large geometry change option provide plastic collapse loads (using the twice-elastic-slope method) and instability loads. For the bending mode, both closing bending and opening bending are considered, and a wide range of parameters related to the bend geometry is considered. Based on the FE results, closed-form approximations of plastic limit and collapse load solutions for pipe bends under combined pressure and bending are proposed

Gas pressure in bubble attached to tube circular outlet

OpenAIRE

Salonen, A; Gay, Cyprien; Maestro, A; Drenckhan, W; Rio, Emmanuelle

2016-01-01

In the present Supplementary notes to our work ``Arresting bubble coarsening: A two-bubble experiment to investigate grain growth in presence of surface elasticity'' (accepted in EPL), we derive the expression of the gas pressure inside a bubble located above and attached to the circular outlet of a vertical tube.

A review of current knowledge on the effects of hydrogen on the pressure tubes of Ontario Hydro operating reactors

International Nuclear Information System (INIS)

Leger, M.

1982-01-01

Since the occurrence of cracking in Zr-2.5 wt% Nb pressure tubes in Pickering 'A' units 3 and 4 in 1974/75 a great deal of information on the behaviour of hydrogen in pressure tube materials has been generated through research effort by both AECL and Ontario Hydro. In order to use this information effectively and to provide direction and co-ordination for ongoing research, a review of available information and current concerns on hydrogen in pressure tubes was undertaken. The review was divided into two main areas of interest: hydrogen ingress and hydride effects. The uncertainties in the rates of hydrogen ingress into the pressure tubes have been found to be very large. On the basis of current knowledge, predictions of the future behaviour of pressure tubes due to hydride effects are extremely difficult

Plastic loads of pipe bends under combined pressure and out-of-plane bending

International Nuclear Information System (INIS)

Lee, Kuk Hee; Kim, Yun Jae; Park, Chi Yong; Lee, Sung Ho; Kim, Tae Ryong

2007-01-01

Based on three-Dimensional (3-D) FE limit analyses, this paper provides plastic limit and TES(Twice- Elastic-Slope) loads for pipe bends under combined pressure and out-of-plane bending. The plastic limit loads are determined from FE limit analyses based on elastic.perfectly-plastic materials using the small geometry change option, and the FE limit analyses using the large geometry change option provide TES plastic loads. A wide range of parameters related to the bend geometry is considered. Based on the FE results, closed-form approximations of plastic limit and TES plastic load solutions for pipe bends under out-of-plane bending are proposed

In-pile creep strain and failure of CW 316 Ti pressurized tubes

International Nuclear Information System (INIS)

Boutard, J.L.; Maillard, A.; Carteret, Y.; Levy, V.; Meny, L.

1984-06-01

The in-pile creep and failure behavior of CW 316 Ti pressurized tubes irradiated in the same rig at 660-680 0 C and 81.4 dpaF max in Phenix is presented and compared to monitors of the same heat. The in-pile plastic strains are of the same order of what is expected from the monitors and are rather independent of the dose rate in the range 4 to 9 x 10 -3 dpaF/h. Such a behavior supports the assumption that the out-of-pile deformation mechanisms are operative in pile and a certain balance occurs between modification of the microstructure, dynamic hardening and deformation mechanisms due to irradiation. Examinations by fractography and optical micrography, show that the failures are intergranular either in-pile or out-of-pile. In both cases the damage consists in intergranular wedge cracks, and no cavitation can be observed by transmission electron microscopy. Then the in-pile embrittlement which gives lower failure strain and time is to be associated with a decrease of the surface energy of grain-boundaries rather then growth and coalescence of cavities

A calculating method of tube-to-tubesheet joints design for steam generator

International Nuclear Information System (INIS)

Zhang Fuyuan

1993-01-01

A theoretical calculating method of the hydraulically expanded tube-to-tubesheet joints design is described. As a mathematical model, the total expanded process of the joints is divided in four stages. with the elastic and plastic theories, the stress, strain and displacement of the tube or tube and tubesheet are analysed by stages, then expansion pressure, deformation, residual stress and push-out force are evaluated. The method may be used to design the steam generators and steel tubular heat exchangers. The paper points out that the hydraulic-expansion plus local roller expansion (hybrid expansion) is better than the only hydraulic-expansion for the tube-to-tubesheet joints of the nuclear steam generators

The transformation behaviour of the beta phase in Zr-2.5 wt% Nb pressure tubes

International Nuclear Information System (INIS)

Griffiths, M.; Winegar, J.E.

1994-12-01

A temperature-time-transformation (TTT) diagram has been developed for the β-phase in Zr-2.5 wt% Nb pressure tubes. The results show that the morphology and/or physical state of the β-phase in pressure tubes has a significant effect on the transformation behaviour compared with a bulk Zr-19 wt%Nb alloy. (author). 14 refs., 1 tab., 15 figs

The application of ductile-fracture analysis to predictions of pressure-tube failure

International Nuclear Information System (INIS)

Simpson, L.A.

1981-08-01

Progress during the past six years towards establishing a method for predicting critical crack length in a reactor pressure tube, based on data from tests on small fracture-mechanics specimens, is reviewed. The disadvantages of relying on data from burst tests alone are described along with the benefits of a small-specimen method. It is clear from the work reviewed that only an approach that can account for the ability of the presssure tube material to increase its crack-growth resistance during stable crack extension is suitable for the prediction of critical crack length. A method that utilizes crack-growth resistance curves based on crack-opening displacement, or the J integral, is described, along with a large body of experimental data. It is concluded that the resistance curve approach provides a viable method for the analysis of fracture in pressure tubes that can greatly improve our understanding of the material's behaviour

Delayed hydride cracking behavior of Zr-2.5Nb alloy pressure tubes for PHWR700

Energy Technology Data Exchange (ETDEWEB)

Sunil, S.; Bind, A.K.; Khandelwal, H.K.; Singh, R.N., E-mail: rnsingh@barc.gov.in; Chakravartty, J.K.

2015-11-15

In order to attain improved in-reactor performance few prototypes pressure tubes of Zr-2.5Nb alloy were manufactured by employing forging to break the cast structure and to obtain more homogeneous microstructure. Both double forging and single forging were employed. The forged material was further processed by employing hot extrusion, cold pilgering and autoclaving. A detailed characterization in terms of mechanical properties and microstructure of the prototype tubes were carried for qualifying it for intended use as pressure tubes in PHWR700 reactors. In this work, Delayed Hydride Cracking (DHC) behavior of the forged Zr-2.5Nb pressure tube material characterized in terms of DHC velocity and threshold stress intensity factor associated with DHC (K{sub IH}) was compared with that of conventionally manufactured material in the temperature range of 200–283 °C. Activation energy associated with the DHC in this alloy was found to be ∼60 kJ/mol for the forged materials.

Conceptual mechanical design for a pressure-tube type supercritical water-cooled reactor

Energy Technology Data Exchange (ETDEWEB)

Yetisir, M.; Diamond, W.; Leung, L.K.H.; Martin, D.; Duffey, R. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

2011-07-01

This paper presents a conceptual mechanical design for a heavy-water-moderated pressure-tube supercritical water (SCW) reactor, which has evolved from the well-established CANDU nuclear reactor. As in the current designs, the pressure-tube SCW reactor uses a calandria vessel and, as a result, many of today's technologies (such as the shutdown safety systems) can readily be adopted with small changes. Because the proposed concept uses a low-pressure moderator, it does not require a pressure vessel that is subject to the full SCW pressure and temperature conditions. The proposed design uses batch refueling and hence, the reactor core is orientated vertically. Significant simplifications result in the design with the elimination of on line fuelling systems, fuel channel end fittings and fuel channel closure seals and thus utilize the best features of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. The safety goal is based on achieving a passive 'no core melt' configuration for the channels and core, so the mechanical features and systems directly reflect this desired attribute. (author)

Conceptual mechanical design for a pressure-tube type supercritical water-cooled reactor

International Nuclear Information System (INIS)

Yetisir, M.; Diamond, W.; Leung, L.K.H.; Martin, D.; Duffey, R.

2011-01-01

This paper presents a conceptual mechanical design for a heavy-water-moderated pressure-tube supercritical water (SCW) reactor, which has evolved from the well-established CANDU nuclear reactor. As in the current designs, the pressure-tube SCW reactor uses a calandria vessel and, as a result, many of today's technologies (such as the shutdown safety systems) can readily be adopted with small changes. Because the proposed concept uses a low-pressure moderator, it does not require a pressure vessel that is subject to the full SCW pressure and temperature conditions. The proposed design uses batch refueling and hence, the reactor core is orientated vertically. Significant simplifications result in the design with the elimination of on line fuelling systems, fuel channel end fittings and fuel channel closure seals and thus utilize the best features of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. The safety goal is based on achieving a passive 'no core melt' configuration for the channels and core, so the mechanical features and systems directly reflect this desired attribute. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-12-15

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

Deuterium absorption in CANDU Zr-2.5Nb pressure tubes

International Nuclear Information System (INIS)

Ploc, R.A.; McRae, G.A.

1999-12-01

Corrosion of CANDU Zr-2.5%Nb pressure tubes in heavy water results in the formation of an oxide film and the absorption of deuterium by the alloy. If deuterium concentrations are allowed to exceed the terminal solid solubility of the alloy, brittle deuterides can form, thereby limiting the service life of a component. In CANDU pressure tubes, ingress rates are largely determined by the metastable β-Zr that is present as a thin layer encasing the predominant α-Zr grains (approximately 90% by volume). The distribution and continuity of the corroded β-phase in the oxide provides a pervasive web for the development of interconnected porosity from the free surface to the oxide/metal interface. Changing the distribution of the β-phase in the alloy changes the nature of the oxide porosity, a technique that can be used to reduce deuterium ingress rates. (author)

Deuterium absorption in CANDU Zr-2.5Nb pressure tubes

Energy Technology Data Exchange (ETDEWEB)

Ploc, R.A.; McRae, G.A

1999-12-01

Corrosion of CANDU Zr-2.5%Nb pressure tubes in heavy water results in the formation of an oxide film and the absorption of deuterium by the alloy. If deuterium concentrations are allowed to exceed the terminal solid solubility of the alloy, brittle deuterides can form, thereby limiting the service life of a component. In CANDU pressure tubes, ingress rates are largely determined by the metastable {beta}-Zr that is present as a thin layer encasing the predominant {alpha}-Zr grains (approximately 90% by volume). The distribution and continuity of the corroded {beta}-phase in the oxide provides a pervasive web for the development of interconnected porosity from the free surface to the oxide/metal interface. Changing the distribution of the {beta}-phase in the alloy changes the nature of the oxide porosity, a technique that can be used to reduce deuterium ingress rates. (author)

Experimental study on condensation heat transfer enhancement and pressure drop penalty factors in four microfin tubes

Energy Technology Data Exchange (ETDEWEB)

Han, D [Korea University, Seoul (Korea). Institute of Advanced Machinery Design; Lee, Kyu-Jung [Korea University, Seoul (Korea). Dept. of Mechanical Engineering

2005-08-01

Heat transfer and pressure drop characteristics of four microfin tubes were experimentally investigated for condensation of refrigerants R134a, R22, and R410A in four different test sections. The microfin tubes examined during this study consisted of 8.92, 6.46, 5.1, and 4 mm maximum inside diameter. The effect of mass flux, vapor quality, and refrigerants on condensation was investigated in terms of the heat transfer enhancement factor and the pressure drop penalty factor. The pressure drop penalty factor and the heat transfer enhancement factor showed a similar tendency for each tube at given vapor quality and mass flux. Based on the experimental data and the heat-momentum analogy, correlations for the condensation heat transfer coefficients in an annular flow regime and the frictional pressure drops are proposed. (author)

Compressed-tube pressure cell for optical studies at ocean pressures: Application to glucose mutarotation kinetics.

Science.gov (United States)

Lamelas, F J

2016-12-01

A self-contained compressed-tube pressure cell is tested to 25 MPa. The cell is very simple to construct and offers stable pressure control with optical access to fluid samples. The physical path length of light through the cell is large enough to measure optical activity. The entire system is relatively small and portable, and it is vibration-free, since a compressor is not used. Operation of the cell is demonstrated by measuring the mutarotation rate of aqueous glucose solutions at 25 °C. A logarithmic plot of the rate constant vs. pressure yields an activation volume for mutarotation of -22 cm 3 /mol, approximately twice the value measured previously at higher pressures.

Fatigue crack initiation at complex flaws in hydrided Zr-2.5%Nb samples from CANDU pressure tubes

International Nuclear Information System (INIS)

Stoica, L.; Radu, V.

2016-01-01

The paper addresses the phenomena which occur at locations where the oxide layer of the inner surface of CANDU tube pressure is damaged by the contact with the fuel element or due to the action of hard particles at the interface between the tube pressure and bearing pad of fuel element. In such situations generate defects, which most often are defects known as ''bearing pad fretting flaws'' or ''debris fretting flaws''. In this paper the experiments are completed in a series of previous works on the mechanical fatigue phenomenon on samples prepared from the pressure tube Zr-2.5% Nb alloy. The phenomenon of variable mechanical stress (or fatigue) may lead to initiation of cracks at the tip of volumetric flaws, according to the accumulation of hydrides, which then fractures and can propagate through the tube wall pressure due to the mechanism of type DHC (Delayed Hydride Cracking). (authors)

Method for studying the plastic buckling of shells. Testing

International Nuclear Information System (INIS)

Alix, M.; Combescure, A.; Hoffmann, A.; Roche, R.

1980-05-01

In this article a description is given of the method selected for studying the elasto-plastic buckling of shells of any shape. The emphasis is mainly on three points: the difficulty in making a strict formulation with respect to plasticity, the model selected (MOTAN model) is presented; the effect of so called 'non conservative' forces; and the effect of great deformations that might precede the buckling. The method is compared to tests: basket handle buckling of bottoms, buckling of elliptical bottoms under internal pressure, of compresses thin tubes, of metal drums, spherical diaphragm, shearing rings [fr

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

International Nuclear Information System (INIS)

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

2012-01-01

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

Non-destructive evaluation of stream generator tubes and pressure tubes from the PHWR reactors, using the rotating magnetic field method

International Nuclear Information System (INIS)

Premel, D.; Placko, D.; Grimberg, R.; Savin, A.

2001-01-01

This work presents a new type of eddy current transducer with a rotating magnetic field devoted to the inspection of steam generator tubes and pressure tubes from the PHWR reactors. A theoretical model has been developed that permits the calculations of the emf induced in the reception coils in the presence of the copper or magnetite deposits, anti-vibration railing and garter springs. (authors)

Replacement of a cracked pressure tube in Bruce GS unit 2

International Nuclear Information System (INIS)

Dunn, J.T.

1982-06-01

In 1982 February, a primary heat transport system leak was detected in the annulus gas system by on-line instrumentation. The source of the leak was found to be a small axial crack in the pressure tube of fuel channel X-14. This fuel channel was removed and replaced by station maintenance staff, and the unit was returned to service five weeks after it had been shut down. The cracked pressure tube was sent to Chalk River Nuclear Laboratories for examination, and the crack was found to be very similar to those found in Pickering GS units 3 and 4 in 1974-75. It was caused by delayed hydride cracking during the period of high residual stress between the time of rolling and the pre-service stress relief

Conceptual design of a pressure tube light water reactor with variable moderator control

International Nuclear Information System (INIS)

Rachamin, R.; Fridman, E.; Galperin, A.

2012-01-01

This paper presents the development of innovative pressure tube light water reactor with variable moderator control. The core layout is derived from a CANDU line of reactors in general, and advanced ACR-1000 design in particular. It should be stressed however, that while some of the ACR-1000 mechanical design features are adopted, the core design basics of the reactor proposed here are completely different. First, the inter fuel channels spacing, surrounded by the calandria tank, contains a low pressure gas instead of heavy water moderator. Second, the fuel channel design features an additional/external tube (designated as moderator tube) connected to a separate moderator management system. The moderator management system is design to vary the moderator tube content from 'dry' (gas) to 'flooded' (light water filled). The dynamic variation of the moderator is a unique and very important feature of the proposed design. The moderator variation allows an implementation of the 'breed and burn' mode of operation. The 'breed and burn' mode of operation is implemented by keeping the moderator tube empty ('dry' filled with gas) during the breed part of the fuel depletion and subsequently introducing the moderator by 'flooding' the moderator tube for the 'burn' part. This paper assesses the conceptual feasibility of the proposed concept from a neutronics point of view. (authors)

Pressure Drop Correlations of Single-Phase and Two-Phase Flow in Rolling Tubes

International Nuclear Information System (INIS)

Xia-xin Cao; Chang-qi Yan; Pu-zhen Gao; Zhong-ning Sun

2006-01-01

A series of experimental studies of frictional pressure drop for single phase and two-phase bubble flow in smooth rolling tubes were carried out. The tube inside diameters were 15 mm, 25 mm and 34.5 mm respectively, the rolling angles of tubes could be set as 10 deg. and 20 deg., and the rolling periods could be set as 5 s, 10 s and 15 s. Combining with the analysis of single-phase water motion, it was found that the traditional correlations for calculating single-phase frictional coefficient were not suitable for the rolling condition. Based on the experimental data, a new correlation for calculating single-phase frictional coefficient under rolling condition was presented, and the calculations not only agreed well with the experimental data, but also could display the periodically dynamic characteristics of frictional coefficients. Applying the new correlation to homogeneous flow model, two-phase frictional pressure drop of bubble flow in rolling tubes could be calculated, the results showed that the relative error between calculation and experimental data was less than ± 25%. (authors)

Simulation of ratcheting in straight pipes using ANSYS with an improved cyclic plasticity model

International Nuclear Information System (INIS)

Hassan, T.; Zhu, Y.; Matzen, V.C.

1996-01-01

Ratcheting has been shown to be a contributing cause of failure in several seismic experiments on piping components and systems. Most commercial finite element codes have been unable to simulate the ratcheting in those tests accurately. The reason for this can be traced to inadequate plasticity constitutive models in the analysis codes. The authors have incorporated an improved cyclic plasticity model, based on an Armstrong-Frederick kinematic hardening rule in conjunction with the Drucker-Palgen plastic modulus equation, into an ANSYS user subroutine. This modified analysis code has been able to simulate quite accurately the ratcheting behavior of a tube subjected to a constant internal pressure and axially strain controlled cycling. This paper describes simulations obtained form this modified ANSYS code for two additional tests: (1) a tube subjected to constant axial stress and prescribed torsional cycling, and (2) a straight pipe subjected to constant internal pressure and quasi-static cyclic bending. The analysis results from the modified ANSYS code are compared to the experimental data, as well as results from ABAQUS and the original ANSYS code. The resulting correlation shows a significant improvement over the original ANSYS and the ABAQUS codes

ELASTIC-PLASTIC AND RESIDUAL STRESS ANALYSIS OF AN ALUMINUM DISC UNDER INTERNAL PRESSURES

Directory of Open Access Journals (Sweden)

Numan Behlül BEKTAŞ

2004-02-01

Full Text Available This paper deals with elastic-plastic stress analysis of a thin aluminum disc under internal pressures. An analytical solution is performed for satisfying elastic-plastic stress-strain relations and boundary conditions for small plastic deformations. The Von-Mises Criterion is used as a yield criterion, and elastic perfectly plastic material is assumed. Elastic-plastic and residual stress distributions are obtained from inner radius to outer radius, and they are presented in tables and figures. All radial stress components, ?r, are compressive, and they are highest at the inner radius. All tangential stress components, ??, are tensile, and they are highest where the plastic deformation begins. Magnitude of the tangential residual stresses is higher than those the radial residual stresses.

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.

Comparison and Analysis of Lithium Bromide-water Absorption Chillers Using Plastic Heat Transfer Tubes and Traditional Lithium Bromide-water Absorption Chillers

OpenAIRE

Xue-dong Zhang

2010-01-01

There are extensive applications of lithium bromide-water absorption chillers in industry, but the heat exchangers corrosion and refrigerating capacity loss are very difficult to be solved. In this paper, an experiment was conducted by using plastic heat transfer tubes instead of copper tubes. As an example, for a lithium bromide-water absorption chiller of refrigerating capacity of 35kW, the correlative performance of the lithium bromide-water absorption chiller using pl...

A study on the critical heat flux for annuli and round tubes under low pressure conditions

International Nuclear Information System (INIS)

Park, Jae Wook

1997-02-01

This study aims to reveal the characteristics of the critical heat flux (CHF) of internally heated concentric annuli and vertical round tubes in low-pressure and low-flow (LPLF) conditions. Although many efforts have been devote to the subject of the CHF during the last forty years, the information on the CHF phenomenon for LPLF conditions is still very limited. The applicable ranges of the CHF correlations for annuli and round tubes are concentrate on the operating conditions of nuclear power plant (NPP), namely high-pressure and high-flow (HPHF) conditions. these facts promoted to collect the reliable CHF data for LPLF conditions for both annuli and round tubes. The critical heat flux data for vertical flow boiling of water in annuli and round tubes at low pressures and low mass fluxes show the following trends: The observed CHF mechanism for annuli was changed in the order of flooding, churn-to-annular flow transition, and local dryout under a large bubble in churn flow as the flow rate was increased from zero to higher values. The observed parametric trends for annuli are consistent with the previous understanding except that the CHF for downward flow is considerably lower (up to 40%) than that for upward flow. The critical quality is much lower than that for round tubes at the same inlet conditions. The observed parametric trends for round tubes are generally consistent with the previous understanding except for system pressure an tube diameter effect. For the system pressure effect, it is observed that the pressure effect is complicated but not so large, whereas the existing CHF correlations do not present the parametric trend exactly. For tube diameter effect, the decreasing trends of CHF with respect to tube diameter was the general understanding so far, but in this region the CHF show a increasing trend of tube diameter. The prediction and the parametric trend analyses are performed by two view points, I.e., for fixed inlet conditions and for local

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.

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

International Nuclear Information System (INIS)

Tsiklauri, G.; Schmitt, B.

1995-09-01

Thermally induced two-phase instabilities in non-uniformly heated boiling charmers 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

The mechanism of coking pressure generation I: Effect of high volatile matter coking coal, semi-anthracite and coke breeze on coking pressure and plastic coal layer permeability

Energy Technology Data Exchange (ETDEWEB)

Seiji Nomura; Merrick Mahoney; Koichi Fukuda; Kenji Kato; Anthony Le Bas; Sid McGuire [Nippon Steel Corporation, Chiba (Japan). Environment and Process Technology Center

2010-07-15

One of the most important aspects of the cokemaking process is to control and restrain the coking pressure since excessive coking pressure tends to lead to operational problems and oven wall damage. Therefore, in order to understand the mechanism of coking pressure generation, the permeability of the plastic coal layer and the coking pressure for the same single coal and the same blended coal were measured and the relationship between them was investigated. Then the 'inert' (pressure modifier) effect of organic additives such as high volatile matter coking coal, semi-anthracite and coke breeze was studied. The coking pressure peak for box charging with more uniform bulk density distribution was higher than that for top charging. It was found that the coking pressure peaks measured at different institutions (NSC and BHPBilliton) by box charging are nearly the same. The addition of high volatile matter coking coal, semi-anthracite and coke breeze to a low volatile matter, high coking pressure coal greatly increased the plastic layer permeability in laboratory experiments and correspondingly decreased the coking pressure. It was found that, high volatile matter coking coal decreases the coking pressure more than semi-anthracite at the same plastic coal layer permeability, which indicates that the coking pressure depends not only on plastic coal layer permeability but also on other factors. Coking pressure is also affected by the contraction behavior of the coke layer near the oven walls and a large contraction decreases the coal bulk density in the oven center and hence the internal gas pressure in the plastic layer. The effect of contraction on coking pressure needs to be investigated further. 33 refs., 18 figs., 5 tabs.

Delayed hydride cracking and elastic properties of Excel, a candidate CANDU-SCWR pressure tube material

International Nuclear Information System (INIS)

Pan, Z.L.

2010-01-01

Excel, a Zr alloy which contains 3.5%Sn, 0.8%Nb and 0.8%Mo, shows high strength, good corrosion resistance, excellent creep-resistance and dimension stability and thus is selected as a candidate pressure tube material for CANDU-SCWR. In the present work, the delayed hydride cracking properties (K IH and the DHC growth rates), the hydrogen solubility and elastic modulus were measured in the irradiated and unirradiated Excel pressure tube material. (author)

Tube to tube excursive instability - sensitivities and transients

International Nuclear Information System (INIS)

Brown, M.; Layland, M.W.

1980-01-01

A simple basic analysis of excursive instability in a boiler tube shows how it depends upon operating conditions and physical properties. A detailed mathematical model of an AGR boiler is used to conduct a steady state parameter sensitivity survey. It is possible from this basis to anticipate the effects of changes in operating conditions and changes in design parameters upon tube to tube stability. Dynamic responses of tubes operating near the stability threshold are examined using a mathematical model. Simulated excursions are triggered by imparting small abrupt pressure changes on the boiler inlet pressure. The influences of the magnitude of the pressure change, waterside friction factor and gas side coupling between tubes are examined. (author)

Thermal performance and pressure drop of spiral-tube ground heat exchangers for ground-source heat pump

International Nuclear Information System (INIS)

Jalaluddin; Miyara, Akio

2015-01-01

Thermal performance and pressure drop of the spiral-tube GHE were evaluated in this present work. A numerical simulation tool was used to carry out this research. The heat exchange rates per meter borehole depth of the spiral-tube GHE with various pitches and their pressure drops were compared with that of the U-tube GHE. Furthermore, a comparative analysis between a spiral pipe and straight pipe was performed. In comparison with the straight pipe, using the spiral pipe in the borehole increased the heat exchange rate to the ground per meter borehole depth. However, the pressure drop of water flow also increased due to increasing the length of pipe per meter borehole depth and its spiral geometry. The accuracy of the numerical model was verified for its pressure drop with some pressure drop correlations. The heat exchange rate and pressure drop of the GHEs are presented. As an example, the heat exchange rate per meter borehole depth of spiral pipe with 0.05 m pitch in the turbulent flow increased of 1.5 times. Its pressure drop also increased of 6 times. However, from the view point of energy efficiency, using the spiral pipe in the ground-source heat pump system gives a better performance than using the straight pipe. The heat exchange rate and pressure drop are important parameter in design of the ground-source heat pump (GSHP) system. - Highlights: • Thermal performance and pressure drop of spiral-tube GHE are presented. • Effects of spiral pitch on thermal performance and pressure drop are analyzed. • Using a spiral pipe increases heat exchange rate per meter borehole depth of GHE. • Pressure drop per meter borehole depth also increases in the spiral pipe.

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.

An experimental study on two-phase pressure drop in small diameter horizontal, downward inclined and vertical tubes

Directory of Open Access Journals (Sweden)

Autee Arun

2015-01-01

Full Text Available An experimental study of two-phase pressure drop in small diameter tubes orientated horizontally, vertically and at two other downward inclinations of θ= 300 and θ = 600 is described in this paper. Acrylic transparent tubes of internal diameters 4.0, 6.0, and 8.0 mm with lengths of 400 mm were used as the test section. Air-water mixture was used as the working fluid. Two-phase pressure drop was measured and compared with the existing correlations. These correlations are commonly used for calculation of pressure drop in macro and mini-microchannels. It is observed that the existing correlations are inadequate in predicting the two-phase pressure drop in small diameter tubes. Based on the experimental data, a new correlation has been proposed for predicting the two-phase pressure drop. This correlation is developed by modification of Chisholm parameter C by incorporating different parameters. It was found that the proposed correlation predicted two-phase pressure drop at satisfactory level.

A comparison of R-22, R-134a, R-410a, and R-407c condensation performance in smooth and enhanced tubes: Part 2, Pressure drop

Energy Technology Data Exchange (ETDEWEB)

Eckels, S J; Tesene, B A

1999-07-01

This paper reports pressure drops during condensation for R-22, R-134a, R-410a, and R-407c in three enhanced tubes and one smooth tube. The test tubes were a 3/8 inch outer diameter smooth tube, a 3/8 inch outer diameter microfin tube, a 5/16 inch outer diameter microfin tube, and a 5/8 inch outer diameter microfin tube. Pressure drops are reported at four mass fluxes, at two saturation temperatures, and over a range of average qualities in the test tubes. The pressure drops for R-410a were approximately 40% lower than those of R-22 in both tubes. R-407c had 10% to 20% lower pressure drops than R-22, while 134-a had slightly larger pressure drops than R-22. The microfin tube pressure drops were, on average, 40% to 80% higher than those for the smooth tube for all refrigerants. The pressure drop penalty of the microfin tube was shown to decrease with increased quality.

Enhancement of leak rate estimation model for corroded cracked thin tubes

International Nuclear Information System (INIS)

Chang, Y.S.; Jeong, J.U.; Kim, Y.J.; Hwang, S.S.; Kim, H.P.

2010-01-01

During the last couple of decades, lots of researches on structural integrity assessment and leak rate estimation have been carried out to prevent unanticipated catastrophic failures of pressure retaining nuclear components. However, from the standpoint of leakage integrity, there are still some arguments for predicting the leak rate of cracked components due primarily to uncertainties attached to various parameters in flow models. The purpose of present work is to suggest a leak rate estimation method for thin tubes with artificial cracks. In this context, 23 leak rate tests are carried out for laboratory generated stress corrosion cracked tube specimens subjected to internal pressure. Engineering equations to calculate crack opening displacements are developed from detailed three-dimensional elastic-plastic finite element analyses and then a simplified practical model is proposed based on the equations as well as test data. Verification of the proposed method is done through comparing leak rates and it will enable more reliable design and/or operation of thin tubes.

Multiaxial ratcheting behavior of zirconium alloy tubes under combined cyclic axial load and internal pressure

Energy Technology Data Exchange (ETDEWEB)

Chen, G.; Zhang, X. [School of Chemical Engineering and Technology, Tianjin University, Tianjin 300072 (China); Xu, D.K. [Environmental Corrosion Center, Institute of Metal Research, Chinese Academy of Sciences, Shenyang 110016 (China); Li, D.H. [Hunan Taohuajiang Nuclear Power Co., Ltd, Yiyang, 413000 (China); Chen, X. [School of Chemical Engineering and Technology, Tianjin University, Tianjin 300072 (China); Zhang, Z., E-mail: zhe.zhang@tju.edu.cn [School of Chemical Engineering and Technology, Tianjin University, Tianjin 300072 (China)

2017-06-15

In this study, a series of uniaxial and multiaxial ratcheting tests were conducted at room temperature on zirconium alloy tubes. The experimental results showed that for uniaxial symmetrical cyclic test, the axial ratcheting strain ɛ{sub x} did not accumulate obviously in initial stage, but gradually increased up to 1% with increasing stress amplitude σ{sub xa}. For multiaxial ratcheting tests, the zirconium alloy tube was highly sensitive to both the axial stress amplitude σ{sub xa} and the internal pressure p{sub i}. The hoop ratcheting strain ɛ{sub θ} increased continuously with the increase of axial stress amplitude, whereas the evolution of axial ratcheting strain ɛ{sub x} was related to the axial stress amplitude. The internal pressure restricted the ratcheting accumulation in the axial direction, but promoted the hoop ratcheting strain on the contrary. The prior loading history greatly restrained the ratcheting behavior of subsequent cycling with a small internal pressure. - Highlights: •Uniaxial and multiaxial ratcheting behavior of the zirconium alloy tubes are investigated at room temperature. •The ratcheting depends greatly on the stress amplitude or internal pressure. •The interaction between the axial and hoop ratcheting mechanisms is greatly dependent on the internal pressure level. •The ratcheting is influenced significantly by the loading history of internal pressure.

Multiaxial ratcheting behavior of zirconium alloy tubes under combined cyclic axial load and internal pressure

International Nuclear Information System (INIS)

Chen, G.; Zhang, X.; Xu, D.K.; Li, D.H.; Chen, X.; Zhang, Z.

2017-01-01

In this study, a series of uniaxial and multiaxial ratcheting tests were conducted at room temperature on zirconium alloy tubes. The experimental results showed that for uniaxial symmetrical cyclic test, the axial ratcheting strain ɛ x did not accumulate obviously in initial stage, but gradually increased up to 1% with increasing stress amplitude σ xa . For multiaxial ratcheting tests, the zirconium alloy tube was highly sensitive to both the axial stress amplitude σ xa and the internal pressure p i . The hoop ratcheting strain ɛ θ increased continuously with the increase of axial stress amplitude, whereas the evolution of axial ratcheting strain ɛ x was related to the axial stress amplitude. The internal pressure restricted the ratcheting accumulation in the axial direction, but promoted the hoop ratcheting strain on the contrary. The prior loading history greatly restrained the ratcheting behavior of subsequent cycling with a small internal pressure. - Highlights: •Uniaxial and multiaxial ratcheting behavior of the zirconium alloy tubes are investigated at room temperature. •The ratcheting depends greatly on the stress amplitude or internal pressure. •The interaction between the axial and hoop ratcheting mechanisms is greatly dependent on the internal pressure level. •The ratcheting is influenced significantly by the loading history of internal pressure.

Life management of Zr 2.5% Nb pressure tube through estimation of fracture properties by cyclic ball indentation technique

International Nuclear Information System (INIS)

Chatterjee, S.; Madhusoodanan, K.; Rama Rao, A.

2015-01-01

In Pressurised Heavy Water Reactors (PHWRs) fuel bundles are located inside horizontal pressure tubes. Pressure tubes made up of Zr 2.5 wt% Nb undergo degradation during in-service environmental conditions. Measurement of mechanical properties of degraded pressure tubes is important for assessing its fitness for further service in the reactor. The only way to accomplish this important objective is to develop a system based on insitu measurement technique. Considering the importance of such measurement, an In-situ Property Measurement System (IProMS) based on cyclic ball indentation technique has been designed and developed indigenously. The remotely operable system is capable of carrying out indentation trial on the inside surface of the pressure tube and to estimate important mechanical properties like yield strength, ultimate tensile strength, hardness etc. It is known that fracture toughness is one of the important life limiting parameters of the pressure tube. Hence, five spool pieces of Zr 2.5 wt% Nb pressure tube of different mechanical properties have been used for estimation of fracture toughness by ball indentation method. Curved Compact Tension (CCT) specimens were also prepared from the five spool pieces for measurement of fracture toughness from conventional tests. The conventional fracture toughness values were used as reference data. A methodology has been developed to estimate the fracture properties of Zr 2.5 wt% Nb pressure tube material from the analysis of the ball indentation test data. This paper highlights the comparison between tensile properties measured from conventional tests and IProMS trials and relates the fracture toughness parameters measured from conventional tests with the IProMS estimated fracture properties like Indentation Energy to Fracture. (author)

Pressure Measurements on a Deforming Surface in Response to an Underwater Explosion in a Water-Filled Aluminum Tube

Directory of Open Access Journals (Sweden)

G. Chambers

2001-01-01

Full Text Available Experiments have been conducted to benchmark DYSMAS computer code calculations for the dynamic interaction of water with cylindrical structures. Small explosive charges were suspended using hypodermic needle tubing inside Al tubes filled with distilled water. Pressures were measured during shock loading by tourmaline crystal, carbon resistor and ytterbium foil gages bonded to the tube using a variety of adhesives. Comparable calculated and measured pressures were obtained for the explosive charges used, with some gages surviving long enough to record results after cavitation with the tube wall.

Water-storage-tube systems. Final report

Energy Technology Data Exchange (ETDEWEB)

Hemker, P.

1981-12-24

Passive solar collection/storage/distribution systems were surveyed, designed, fabricated, and mechanically and thermally tested. The types studied were clear and opaque fiberglass tubes, metal tubes with plastic liners, and thermosyphoning tubes. (MHR)

Heat exchanging tube behaviour in steam generators of pressurized water reactors

International Nuclear Information System (INIS)

Pastor, D.; Oertel, K.

1979-01-01

Based on a comprehensive failure statistics, materials corrosion chemistry and thermohydraulics problems of the tubings of steam generators are considered. A historical review of failures in the tubings of steam generators in pressurized water reactors reflects the often successless measures by designers, manufacturers and operating organizations for preventing failures, especially with regard to materials selection and water regime. It is stated that laboratory tests could not give sufficient information about safe and stable operation of nuclear steam generators unless real constructive, hydrodynamic, thermodynamical and chemical conditions of operation had been taken into account. (author)

Hydraulic pressure pulses with elastic and plastic structural flexibility: test and analysis (LWBR Development Program)

International Nuclear Information System (INIS)

Schwirian, R.E.

1978-03-01

Pressure pulse tests were conducted with a flexible test section in a test vessel filled with room temperature water. The pressure pulses were generated with a drop hammer and piston pulse generator and were of a sufficient magnitude to cause plastic deformation of the test section. Because of the strong pressure relief effect of the deforming test section, pressure pulse magnitudes were below 265 psig in magnitude and had durations of 50 to 55 msecs. Calculations performed with the FLASH-35 bi-linear hysteresis model of structural deformation show good agreement with experiment. In particular, FLASH 35 adequately predicts the decrease in peak pressure and the increase in pulse duration due to elastic and plastic deformation of the test section. Predictions of flexible member motion are good, but are less satisfactory than the pressure pulse results due to uncertainties in the values of yield point and beyond yield stiffness used to model the various flexible members. Coupled with this is a strong sensitivity of the FLASH 35 predictions to the values of yield point and beyond yield stiffness chosen for the various flexible members. The test data versus calculation comparisons presented here provide preliminary qualification for FLASH 35 calculations of transient hydraulic pressures and pressure differentials in the presence of flexible structural members which deform both elastically and plastically

Free Piston Double Diaphragm Shock Tube

OpenAIRE

OGURA, Eiji; FUNABIKI, Katsushi; SATO, Shunichi; ABE, Takashi; 小倉, 栄二; 船曳, 勝之; 佐藤, 俊逸; 安部, 隆士

1997-01-01

A free piston double diaphragm shock tube was newly developed for generation of high Mach number shock wave. Its characteristics was investigated for various operation parameters; such as a strength of the diaphragm at the end of the comparession tube, an initial pressure of low pressure tube, an initial pressure of medium pressure tube and the volume of compression tube. Under the restriction of fixed pressures for the driver high pressure tube (32×10^5Pa) and the low pressure tube (40Pa) in...

MEASUREMENT OF ENDOTRACHEAL TUBE CUFF PRESSURE IN MECHANICALLYVENTILATED PATIENTS ON ARRIVAL TO INTENSIVE CARE UNIT - A CROSS-SECTIONAL STUDY

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Arun Kumar Ajjappa

2017-04-01

Full Text Available BACKGROUND The monitoring of Endotracheal Tube (ETT cuff pressure in intubated patients on arrival to intensive care unit is very essential. The cuff pressure must be within an optimal range of 20-30cm H2O ensuring ventilation with no complications related to cuff overinflation and underinflation. This can be measured with a cuff pressure manometer. The aim of the study is to measure the endotracheal tube cuff pressure in patients on arrival to intensive care unit and to identify prevalence of endotracheal cuff underinflation and overinflation. MATERIALS AND METHODS A cross-sectional study was done on mechanically-ventilated patients who were intubated in casualty (emergency department on arrival to intensive care unit in S.S. Institute of Medical Sciences and Research Centre, Davangere. About 50 critically-ill patients intubated with a high volume, low pressure endotracheal tube were included in the study. An analogue manometer was used to measure the endotracheal tube cuff pressure. It was compared with the recommended level. The settings of mechanical ventilation, endotracheal tube size and peak airway pressure were recorded. RESULTS It was found that the mean cuff pressure was 64.10 cm of H2O with a standard deviation of 32.049. Of the measured cuff pressures, only 2% had pressures within an optimal range (20-30cm of H2O. 88% had cuff pressures more than 30cm of H2O. The mean peak airway pressure found to be 20.50cm of H2O with a Standard Deviation (SD of 5.064. CONCLUSION This study is done to emphasise the importance of cuff pressure measurement in all mechanically-ventilated patients as cuff pressure is found to be high in most of the patients admitted to intensive care unit. Complications of overinflation and underinflation can only be prevented if the acceptable cuff pressures are achieved.

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

International Nuclear Information System (INIS)

Guest, J.C.; Hutchings, J.A.

1975-01-01

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 550 0 C. A flow stress concept developed for simple bursting tests was shown to apply to creep conditions at 550 0 C. 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)

Nuclear power - replacement of pressure tubes in CANDU reactors

International Nuclear Information System (INIS)

Anon.

1992-01-01

The CANDU pressure tube reactor is an effective electricity generator. While most units have been built in Canada, units are successfully operated in Argentina and Korea as well as India and Pakistan, which have early versions of the same concept. Units are also under construction in Korea and Romania. The main constructional components of a CANDU core are the calandria vessel, the fuel channels and the reactivity control mechanisms. The fuel channel, in particular the pressure tubes, see an environment comprising high flux, high temperature water at high pressures, which induces changes in the properties and dimensions of the channel components. From the first, fuel channels were designed to be replaced because of the difficulty in predicting the behaviour of zirconium alloys in such service over a long period of time. In fact some phenomena, that were not known at the time of the earliest designs, have led to unacceptable changes in the properties of the channels and these early reactors have had to be retubed at half their intended life. These deficiencies have been corrected in the latest designs and fuel channels in reactors that have commenced operation over the last 10 years, are predicted to reach the intended 30 years life before replacement is necessary. The changing of fuel channels, the details and experience of which are explained, has been shown to be an effective way of refurbishing the CANDU reactor, extending its lifetime a further 25-30 years. (author)

Estimation of fracture toughness of Zr 2.5% Nb pressure tube of Pressurised Heavy Water Reactor using cyclic ball indentation technique

Energy Technology Data Exchange (ETDEWEB)

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

2016-08-15

Highlights: • Measurement of fracture toughness 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 situ Property Measurement System (IProMS) has been designed in house. • Conventional and IProMS tests conducted on pressure tube spool pieces having different mechanical properties. • Correlation has been established between the conventional and IProMS estimated fracture properties. - Abstract: In Pressurised Heavy Water Reactors (PHWRs) fuel bundles are located inside horizontal pressure tubes made up of Zr 2.5 wt% Nb alloy. Pressure tubes undergo degradation during its service life due to high pressure, high temperature and radiation environment. Measurement of mechanical properties of degraded pressure tubes is important for assessing their fitness for further operation. Presently as per safety guidelines imposed by the regulatory body, a few pre-decided pressure tubes are removed from the reactor core at regular intervals during the planned reactor shut down to carry out post irradiation examination (PIE) in a laboratory which consumes lots of man-rem and imposes economic penalties. Hence a system is indeed felt necessary which can carry out experimental trials for measurement of mechanical properties of pressure tubes under in situ conditions. The only way to accomplish this important objective is to develop a system based on an in situ measurement technique. 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 an indentation test either on the 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 conditions. Considering the importance of such measurements, an In situ Property

Precision tubes for high-pressure diesel injection lines; Praezisrohre fuer Hochdruck-Dieseleinspritzleitungen

Energy Technology Data Exchange (ETDEWEB)

Hagedorn, M.; Lechtenfeld, U.; Zaremba, A. [Mannesmann Praezisrohr GmbH, Hamm (Germany)

2008-03-15

The requirements on diesel injection lines raise because of increasing customers demands and more rigid environmental laws. In this context higher injection pressures effect both aspects positively. One important condition for increasing pressure levels is the economical provision of suitable injection lines. To reach this aim, Mannesmann Praezisrohr GmbH developed precision tubes for injection lines, which are fulfilling these increasing requirements. (orig.)

Heat transfer and pressure drop in a tube bank inclined with respect to the flow

Energy Technology Data Exchange (ETDEWEB)

Yanez Moreno, A.A.

1985-01-01

This research is intended to lend understanding and to quantify the heat-transfer and fluid-flow characteristics for yawed tube banks in both staggered and in-line arrays. The investigated range of yaw angle was from 90 (crossflow) to 45/sup 0/, while the freestream Reynolds number (based on the tube diameter) ranged between 7000 and 45,000. The transverse and longitudinal center-to-center distances between the tubes were S/sub T//D = S/sub L//D = 2, respectively. The heat-transfer experiments were carried out on a row-by-row basis. Pressure drop measurements were made not only upstream and downstream of the tube bank but also within it. The patterns of fluid flow adjacent to the tubes were visualized using the oil-lampblack technique. A detailed study was carried out to determine the heat-transfer characteristics of a yawed single cylinder. The yaw angle range was between 90 and 30/sup 0/, and flow visualization was also performed. The pressure measurements showed that the overall dimensionless pressure drop for the staggered array is higher than that for the in-line array for a given Reynolds number or yaw. The flow-visualization patterns showed that the boundary layer separation depends on the yaw angle. For the single cylinder, the Nusselt number varied with the yaw angle in an undulating manner and did not correlate with the Independence Principle.

Delayed hydride cracking in Zr-2.5% wt Nb pressure tubes

International Nuclear Information System (INIS)

Cirimello, Pablo; Haddad, Roberto; Domizzi, Gladys

2003-01-01

During service, pressure tubes of CANDU nuclear power reactor are prone to suffer crack growth by delayed hydride cracking (DHC). For a given H 2 plus D 2 concentration there is a critical temperature (T c ) below which DHC may occur. In this work, T c was measured for CCT specimens cut from Zr-2.5 Wt % Nb pressure tubes. Hydrogen was added to the specimens to get concentrations of 40, 59 and 72 ppm. It was found that T c is higher than the corresponding precipitation temperature. The axial crack velocity (V p ) was also measured. Decreasing temperature from T c makes V p increase until a maximum is attained at a temperature close to precipitation temperature. At lower temperatures, in the presence of precipitated hydrides, decreasing temperature implies lower velocities, following an Arrhenius law: Vp=Aexp(-Q/RT), with an activation energy Q= 66 KJ/mol K. (author)

Hot hardness studies on zircaloy 2 pressure tube along three orientations

International Nuclear Information System (INIS)

Kutty, T.R.G.; Ravi, K.; Jarvis, T.; Sengupta, A.K.; Majumdar, S.; Tewari, R.; Shrivastava, D.; Dey, G.K.

2002-01-01

Zirconium based alloys are the natural choice for both the fuel element cans and in-core structural components in water cooled nuclear reactors. In this paper, the hot hardness behaviour of zircaloy 2 pressure tubes has been examined from room temperature to 400 degC using a hot hardness tester. For the purpose of comparison, the hardness of the as cast and room temperature rolled specimens has also been carried out. For this, the samples were cut along three orientations and hardness was measured in each of these directions using Vickers diamond pyramid indenter. The variation in hardness of the pressure tube samples show that the hardness was highest along circumferential direction and least along the axial direction. The room temperature rolled samples showed highest hardness along the rolling planes. These variations in hardness could be explained in terms of development of texture during working on the material. (author)

Analytical solution for stress, strain and plastic instability of pressurized pipes with volumetric flaws

International Nuclear Information System (INIS)

Cunha, Sérgio B.; Netto, Theodoro A.

2012-01-01

The mechanical behavior of internally pressurized pipes with volumetric flaws is analyzed. The two possible modes of circumferentially straining the pipe wall are identified and associated to hypothesized geometries. The radial deformation that takes place by bending the pipe wall is studied by means of axisymmetric flaws and the membrane strain developed by unequal hoop deformation is analyzed with the help of narrow axial flaws. Linear elastic shell solutions for stress and strain are developed, the plastic behavior is studied and the maximum hoop stress at the flaw is related to the undamaged pipe hoop stress by means of stress concentration factors. The stress concentration factors are employed to obtain equations predicting the pressure at which the pipe fails by plastic instability for both types of flaw. These analytical solutions are validated by comparison with burst tests on 3″ diameter pipes and finite element simulations. Forty-one burst tests were carried out and two materials with very dissimilar plastic behavior, carbon steel and austenitic stainless steel, were used in the experiments. Both the analytical and the numerical predictions showed good correlation with the experimentally observed burst pressures. - Highlights: ► An analytical model for the burst of a pipe with a volumetric flaw is developed. ► Deformation, strain and stress are modeled in the elastic and plastic domains. ► The model is comprehensively validated by experiments and numerical simulations. ► The burst pressure model’s accuracy is equivalent to finite element simulations.

Computation of the effect of pipe plasticity on pressure-pulse propagation in a fluid system

International Nuclear Information System (INIS)

Youngdahl, C.K.; Kot, C.A.

1975-04-01

A simple computational model is developed for incorporating the effect of elastic-plastic deformation of piping on pressure-transient propagation in a fluid system. A computer program (PLWV) is described that incorporates this structural interaction model into a one-dimensional method-of-characteristics procedure for fluid-hammer analysis. Computed results are shown to be in good agreement with available experimental data. The most significant effect of plastic deformation is to limit the peak pressure of a pulse leaving a pipe to approximately the yield pressure of the pipe, if the pipe is sufficiently long. 7 references. (U.S.)

Anisotropic thermal creep of internally pressurized Zr-2.5Nb tubes

International Nuclear Information System (INIS)

Li, W.; Holt, R.A.

2010-01-01

The anisotropy of creep of internally pressurized cold-worked Zr-2.5Nb tubes with different crystallographic textures is reported. The stress exponent n was determined to be about three at transverse stresses from 100 to 250 MPa with an activation energy of ∼99.54 kJ/mol in the temperature range 300-400 o C. The stress exponent increased to ∼6 for transverse stresses from 250 to 325 MPa. From this data an experimental regime of 350 o C and 300 MPa was established in which dislocation glide is the likely strain-producing mechanism. Creep tests were carried out under these conditions on internally pressurized Zr-2.5Nb tubes with 18 different textures. Creep strain and creep anisotropy (ratio of axial to transverse steady-state creep rate, ε . A /ε . T ) exhibited strong dependence on crystallographic textures of the Zr-2.5Nb tubes. It was found that the values of (ε . A /ε . T ) increased as the difference between the resolved faction of basal plane normals in the transverse and radial directions (f T - f R ) increases. The tubes with the strongest radial texture showed a negative axial creep strain and a negative creep rate ratio (ε . A /ε . T ) and tubes with a strong transverse texture exhibited the positive values of steady-state creep rate ratio (ε . A /ε . T ) and good creep resistance in the transverse direction. These behaviors are qualitatively similar to those observed during irradiation creep, and also to the predictions of polycrystalline models for creep in which glide is the strain-producing mechanism and prismatic slip is the dominant system. A detailed analysis of the results using polycrystalline models may assist in understanding the anisotropy of irradiation creep.

Fracture toughness of irradiated Zr-2.5Nb pressure tube from KAPS-2 evaluated using disk compact tension specimens

International Nuclear Information System (INIS)

Shah, Priti Kotak; Dubey, J.S.; Balakrishnan, K.S.; Shriwastaw, R.S.; Dhotre, M.P.; Bhandekar, A.; Pandit, K.M.; Anantharaman, S.

2013-12-01

The report gives the results of the fracture toughness tests carried out over the range of temperatures on specimens prepared from the irradiated S-07Zr-2.5Nb pressure tube removed from Kakrapar Atomic Power Station-2 (KAPS-2) as a part of materials surveillance programme. The pressure tube had experienced ∼ 8 effective full power years (EFPY) of reactor operation and had hydrogen equivalent (H eq ) content less than 20 ppm along the tube length. The fracture toughness tests have been carried out using 30 mm Disk Compact Tension (DCT) specimens, that were punched out of the irradiated pressure tube. The disk punching was carried out using specially made shielded enclosure and hydraulic press. Fatigue pre-cracking and fracture toughness tests were performed using servo-hydraulic universal testing machine with Direct Current Potential Drop (DCPD) equipment to monitor the crack length. The tests were carried out at different test temperature from ambient to 300℃. The fracture toughness values have been used to estimate the critical pressure for the tube. The fracture properties indicate that such tubes have sufficient toughness to satisfy the Leak-Before-Break (LBB) criterion for in-reactor operation. (author)

A perturbation model for the oscillatory flow of a Bingham plastic in rigid and periodically displaced tubes.

Science.gov (United States)

De Chant, L J

1999-10-01

An approximate analytical model for the pulsatile flow of an ideal Bingham plastic fluid in both a rigid and a periodically displaced tube has been developed using regular perturbation methods. Relationships are derived for the velocity field and dimensionless flow rate. The solution compares adequately with available experimentally measured oscillatory non-Newtonian fluid flow data. These solutions provide useful analytical models supporting experimental and computation studies of arterial blood flow.

Pressure sores–a constant problem for plegic patients and a permanent challenge for plastic surgery

Science.gov (United States)

Marinescu, S; Florescu, IP; Jecan, C

2010-01-01

Pressure sores–a constant problem for plegic patients and a permanent challenge for plastic surgery Pressure sores can be defined as lesions caused by unrelieved pressure resulting in damage of the underlying tissue. They represent a common problem in the pathology of plegic patients and, plastic surgery has a significant role in their treatment. Pressure sores occur over bony prominences and so, they are most commonly seen at the sacrum and trochanters in paralyzed patients and at ischium for the patients who sit in a wheelchair for a long time. For these patients, surgical treatment is very important because on one hand, it stops the loss of nutrients and proteins at the site of the pressure sore, and on the other hand, it permits the initiation of neuromuscular recuperation treatment much faster. PMID:20968200

Domain sizing diffraction and deformation degree of Zr-2.5%Nb pressure pipes of national production

International Nuclear Information System (INIS)

Banchik, A D; Buoli, D C; Flores, A; Vizcaino, P

2012-01-01

The Life Extension Program for the PHWR Embalse power reactor requires the replacement of the original Zr-2, 5% Nb pressure tubes by a new set of Zr-2,5%Nb pressure tubes. It was also decided to make the new set de pressure tubes locally from imported extruded tubes. As the mechanical properties of pressure tubes in operation depends on the microstructural deformation it is necessary to add to the normal macroscopic quality control same microstructural studies, specially .during the development stage. In the present work x-ray diffraction techniques are applied to determine the magnitude of the micro-structural deformation of three pressure tubes obtained Zr-2,5% Nb extruded tubes that were processes following designer specifications. The pressure tubes were obtained by two step of plastic deformation in a HPTR 60-120 tube rolling machine without intermediate annealing and a final thermal treatment. The line width at half height (FWHM) diffraction lines diagrams are drawn in a Williamson - Hall plot to determine the size of the coherent diffraction domains and estimate the degree of micro-structural deformation. To evaluate the possible effects of texture and/or anisotropy the X-ray measurement were made at the three principal directions of the processed tubes. In summary, the three pressure tubes obtained follow the same average Williamson-Hall line and similar dispersion with respect to that line. That result implies an acceptable homogeneity between the micro-deformation of these pressure tubes. As then have been chosen at random from a lot of preceded pressure tubes, it is possible to predict similar conclusions for the entire batch (author)

Electromechanical phase transition of a dielectric elastomer tube under internal pressure of constant mass

Directory of Open Access Journals (Sweden)

Song Che

2017-05-01

Full Text Available The electromechanical phase transition for a dielectric elastomer (DE tube has been demonstrated in recent experiments, where it is found that the unbulged phase gradually changed into bulged phase. Previous theoretical works only studied the transition process under pressure control condition, which is not consistent with the real experimental condition. This paper focuses on more complex features of the electromechanical phase transition under internal pressure of constant mass. We derive the equilibrium equations and the condition for coexistent states for a DE tube under an internal pressure, a voltage through the thickness and an axial force. We find that under mass control condition the voltage needed to maintain the phase transition increases as the process proceeds. We analyze the entire process of electromechanical phase transition and find that the evolution of configurations is also different from that for pressure control condition.

Mechanistic modeling of heat transfer process governing pressure tube-to-calandria tube contact and fuel channel failure

International Nuclear Information System (INIS)

Luxat, J.C.

2002-01-01

Heat transfer behaviour and phenomena associated with ballooning deformation of a pressure tube into contact with a calandria tube have been analyzed and mechanistic models have been developed to describe the heat transfer and thermal-mechanical processes. These mechanistic models are applied to analyze experiments performed in various COG funded Contact Boiling Test series. Particular attention is given in the modeling to characterization of the conditions for which fuel channel failure may occur. Mechanistic models describing the governing heat transfer and thermal-mechanical processes are presented. The technical basis for characterizing parameters of the models from the general heat transfer literature is described. The validity of the models is demonstrated by comparison with experimental data. Fuel channel integrity criteria are proposed which are based upon three necessary and sequential mechanisms: Onset of CHF and local drypatch formation at contact; sustained film boiling in the post-contact period; and creep strain to failure of the calandria tube while in sustained film boiling. (author)

DHC velocity comparison of CANDU Zr-2.5Nb pressure tube for the difference of test environments

International Nuclear Information System (INIS)

Cho, S. Y.; Kim, Y. S.; Oh, D. Z.; Yim, K. S.; Yim, Y. W.

2001-01-01

Zr-2.5Nb Pressure tube was used in the distilled water under high temperature and pressure. However, the evaluation of DHCV for pressure tube was limited in the air until now. Therefore, it was necessary for DHCV both in the air and in the distilled water under high temperature and pressure to evaluate. In advance, new DHC equipment simulating the real operating condition in the distilled water under high temperature and pressure was developed and DHCV test was conducted by this equipment. The test was carried out under simulated condition using distilled water of 250 .deg. C, 86bar and this result was compared with of DHCV in the air of 250 .deg. C. DHCV of the distilled water was ranged from 8.42x10 -8 to 9.92x 10 -8 m/s and the average value was 9.01x 10 -8 m/s. As compared with the air condition, it was found that characteristics of DHCV was not affected by the distilled water of high temperature and pressure. At the same temperature, DHCV of the irradiated pressure tube was faster than that of test result

INVESTIGATION OF MATERIAL RESISTANCE TO PLASTIC DEFORMATION AT PROCESSING METALS BY PRESSURE WITH IMPOSING ULTRASONIC OSCILLATIONS

Directory of Open Access Journals (Sweden)

V. V. Klubovich

2007-01-01

Full Text Available The paper contains substantiation for application of experimental technique in order to investigate material resistance to plastic deformation at processing metals by pressure with imposing ultrasonic oscillations while proceeding from laws of similarity. It is shown that at modeling any metal processing by pressure with imposing ultrasonic oscillations it is possible to consider that actual elastic and plastic metal properties remain constant during processing under ultrasound action. The second aspect that requires a special attention at modeling is pulse or vibration-shock deformation at processing metals by pressure with imposing ultrasonic oscillations.

Processing biobased polymers using plasticizers: Numerical simulations versus experiments

Science.gov (United States)

Desplentere, Frederik; Cardon, Ludwig; Six, Wim; Erkoç, Mustafa

2016-03-01

In polymer processing, the use of biobased products shows lots of possibilities. Considering biobased materials, biodegradability is in most cases the most important issue. Next to this, bio based materials aimed at durable applications, are gaining interest. Within this research, the influence of plasticizers on the processing of the bio based material is investigated. This work is done for an extrusion grade of PLA, Natureworks PLA 2003D. Extrusion through a slit die equipped with pressure sensors is used to compare the experimental pressure values to numerical simulation results. Additional experimental data (temperature and pressure data along the extrusion screw and die are recorded) is generated on a dr. Collin Lab extruder producing a 25mm diameter tube. All these experimental data is used to indicate the appropriate functioning of the numerical simulation tool Virtual Extrusion Laboratory 6.7 for the simulation of both the industrial available extrusion grade PLA and the compound in which 15% of plasticizer is added. Adding the applied plasticizer, resulted in a 40% lower pressure drop over the extrusion die. The combination of different experiments allowed to fit the numerical simulation results closely to the experimental values. Based on this experience, it is shown that numerical simulations also can be used for modified bio based materials if appropriate material and process data are taken into account.

Natural convection in vertical tubes with variable properties and prescribed pressure at the end of the tube

International Nuclear Information System (INIS)

Almeida Rego, O.A. de; Fernandes, E.C.

1983-01-01

The analysis of free convection flow development in a heated vertical open tube was established in the present work for the air, with Prandtl number equal to 0.7 and for water with Prandtl numbers equal to 1.0, 2.5 and 5.0 with variable properties and prescribed pressure conditions at the end of the tube. It is considered that the flow is incompressible, laminar and stable and can be described by the continuity, momentum and energy equations with the usual boundary-layer assumptions. The equations were solved by finite difference method and from the velocity and temperature distributions many quantities such as dimensioless flow and heat rates and Nusselt numbers can be determined. (Author) [pt

Evaluation of a leaking crack in an irradiated CANDU pressure tube

International Nuclear Information System (INIS)

Coleman, C.E.; Simpson, L.A.

1988-06-01

Leak-before-break is used in CANDU reactors as part of the defence against rupture of the pressure tubes. Two important features of this technique are the action time available for detection of a leaking crack and the size of the leak allowing crack location. Support for continued reliance on leak-before-break is being obtained from experiments, on irradiated Zr-2.5 Nb pressure tubes attached to their end fittings, that simulate the behaviour of a leaking crack in a reactor. At reactor operating temperatures leaking cracks grow more slowly than dry cracks in the laboratory because they are cooled when pressurised water flashes to steam on their surface. These cracks remain stable till they are at least 70 mm long. From the results of these experiments the action time is at least 100 h. The leak rate increases rapidly when a through-wall crack extends a small amount, thus greatly assisting with crack location

Eccentric pressurized tube for measuring creep rupture

International Nuclear Information System (INIS)

Schwab, P.R.

1981-01-01

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

Decreases in blood ethanol concentrations during storage at 4 °C for 12 months were the same for specimens kept in glass or plastic tubes

Directory of Open Access Journals (Sweden)

A.W. Jones

2016-04-01

Full Text Available Background: The stability of ethanol was investigated in blood specimens in glass or plastic evacuated tubes after storage in a refrigerator at 4 °C for up to 12 months. Methods: Sterile blood, from a local hospital, was divided into 50 mL portions and spiked with aqueous ethanol (10% w/v to give target concentrations of 0.20, 1.00, 2.00 and 3.00 g/L. Ethanol was determined in blood by headspace gas chromatography (HS-GC with an analytical imprecision of <3% (coefficient of variation, CV%. Aliquots of blood were re-analysed after 2, 7, 14, 28, 91, 182 and 364 days of storage at 4 °C. Results: The standard deviation (SD of analysis by HS-GC was 0.0059 g/L at 0.20 g/L and 0.0342 g/L at 3.00 g/L, corresponding to CVs of 2.9% and 1.1%, respectively. The decreases in blood ethanol content were analytically significant after 14–28 days of storage for both glass and plastic tubes The mean (lowest and highest loss of ethanol after 12 months storage was 0.111 g/L (0.084–0.129 g/L for glass tubes and 0.112 g/L (0.088–0.140 g/L for plastic tubes. The corresponding percentage losses of ethanol were 43–45% at a starting concentration of 0.20 g/L and 3.9–4.1% at 3.00 g/L. Conclusion: The concentration of ethanol in blood gradually decreases during storage at 4 °C. After 12 months storage the absolute decrease in concentration was ~0.11 g/L when the starting concentration ranged from 0.20 to 3.0 g/L. Decreases in ethanol content were the same for specimens kept in glass or plastic evacuated tubes. Keywords: Alcohol, Analysis, Blood, Ethanol stability, Plastic vs glass tubes, Storage conditions

Analytical solution for stress, strain and plastic instability of pressurized pipes with volumetric flaws

Energy Technology Data Exchange (ETDEWEB)

Cunha, Sergio B., E-mail: sbcunha@petrobras.com.br [PETROBRAS/TRANSPETRO, Av. Pres. Vargas 328 - 7th floor, Rio de Janeiro, RJ 20091-060 (Brazil); Netto, Theodoro A., E-mail: tanetto@lts.coppe.ufrj.br [COPPE, Federal University ot Rio de Janeiro, Ocean Engineering Department, PO BOX 68508, Rio de Janeiro - RJ (Brazil)

2012-01-15

The mechanical behavior of internally pressurized pipes with volumetric flaws is analyzed. The two possible modes of circumferentially straining the pipe wall are identified and associated to hypothesized geometries. The radial deformation that takes place by bending the pipe wall is studied by means of axisymmetric flaws and the membrane strain developed by unequal hoop deformation is analyzed with the help of narrow axial flaws. Linear elastic shell solutions for stress and strain are developed, the plastic behavior is studied and the maximum hoop stress at the flaw is related to the undamaged pipe hoop stress by means of stress concentration factors. The stress concentration factors are employed to obtain equations predicting the pressure at which the pipe fails by plastic instability for both types of flaw. These analytical solutions are validated by comparison with burst tests on 3 Double-Prime diameter pipes and finite element simulations. Forty-one burst tests were carried out and two materials with very dissimilar plastic behavior, carbon steel and austenitic stainless steel, were used in the experiments. Both the analytical and the numerical predictions showed good correlation with the experimentally observed burst pressures. - Highlights: Black-Right-Pointing-Pointer An analytical model for the burst of a pipe with a volumetric flaw is developed. Black-Right-Pointing-Pointer Deformation, strain and stress are modeled in the elastic and plastic domains. Black-Right-Pointing-Pointer The model is comprehensively validated by experiments and numerical simulations. Black-Right-Pointing-Pointer The burst pressure model's accuracy is equivalent to finite element simulations.

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

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)

Chatterjee, S.; Panwar, Sanjay; Madhusoodanan, K.

2015-01-01

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

Optimization of drift gases for accuracy in pressurized drift tubes

CERN Document Server

Kirchner, J J; Dinner, A R; Fidkowski, K J; Wyatt, J H

2001-01-01

Modern detectors such as ATLAS use pressurized drift tubes to minimize diffusion and achieve high coordinate accuracy. However, the coordinate accuracy depends on the exact knowledge of converting measured times into coordinates. Linear space-time relationships are best for reconstruction, but difficult to achieve in the $E \\propto \\frac{1}{r}$ field. Previous mixtures, which contained methane or other organic quenchers, are disfavored because of ageing problems. From our studies of nitrogen and carbon dioxide, two mixtures with only small deviations from linearity were determined and measured. Scaling laws for different pressures and magnetic fields are also given.

Optimization of drift gases for accuracy in pressurized drift tubes

International Nuclear Information System (INIS)

Kirchner, J.J.; Becker, U.J.; Dinner, R.B.; Fidkowski, K.J.; Wyatt, J.H.

2001-01-01

Modern detectors such as ATLAS use pressurized drift tubes to minimize diffusion and achieve high coordinate accuracy. However, the coordinate accuracy depends on the exact knowledge of converting measured times into coordinates. Linear space-time relationships are best for reconstruction, but difficult to achieve in the E∝1/r field. Previous mixtures, which contained methane or other organic quenchers, are disfavored because of ageing problems. From our studies of nitrogen and carbon dioxide, two mixtures with only small deviations from linearity were determined and measured. Scaling laws for different pressures and magnetic fields are also given

Study of the consequences of the rupture of a pressure tube in the tank of a gas-cooled, heavy-water moderated reactor

International Nuclear Information System (INIS)

Hareux, F.; Roche, R.; Vrillon, B.

1964-01-01

Bursting of a pressure tube in the tank of a heavy water moderated-gas cooled reactor is an accident which has been studied experimentally about EL-4. A first test (scale 1) having shown that the burst of a tube does not cause the rupture of adjacent tubes, tests on the tank resistance have been undertaken with a very reduced scale model (1 to 10). It has been found that the tank can endure many bursts of tube without any important deformation. Transient pressure in the tank is an oscillatory weakened wave, the maximum of which (pressure peak) has been the object of a particular experimental study. It appears that the most important parameters which affect the pressure peak are; the pressure of the gas included in the bursting pressure tube, the volume of this gas, the mass of air included in the tank and the nature of the gas. A general method to calculate the pressure peak value in reactor tanks has been elaborated by direct application of experimental data. (authors) [fr

Characteristics of CANDU fuel bundles that caused pressure tube fretting at the bundle midplane

Energy Technology Data Exchange (ETDEWEB)

Dennier, D; Manzer, A M [Atomic Energy of Canada Ltd., Mississauga, ON (Canada); Koehn, E [Ontario Hydro, Toronto, ON (Canada)

1996-12-31

Detailed measurements on new bundles, and those that caused fretting during in- and out-reactor tests, have given insight into the factors responsible for fretting at the midplane of the inlet bundle. Bottom fuel elements that were attached near radial endplate spokes and had inboard bearing pads in the rolled joint cavity produced a significant portion of the observed fret marks. These elements are influenced by several driving forces that deflect the centre bearing pads towards the pressure tube surface. The evidence suggests that slight changes in bundle design may be possible to reduce pressure tube fretting. (author). 4 refs., 3 tabs., 8 figs.

Effects on Vocal Fold Collision and Phonation Threshold Pressure of Resonance Tube Phonation with Tube End in Water

Science.gov (United States)

Enflo, Laura; Sundberg, Johan; Romedahl, Camilla; McAllister, Anita

2013-01-01

Purpose: Resonance tube phonation in water (RTPW) or in air is a voice therapy method successfully used for treatment of several voice pathologies. Its effect on the voice has not been thoroughly studied. This investigation analyzes the effects of RTPW on collision and phonation threshold pressures (CTP and PTP), the lowest subglottal pressure…

The plastic work required to induce the limit pressure of a plain cylinder

International Nuclear Information System (INIS)

Lee, K.S.; Moreton, D.N.; Moffat, D.G.

2005-01-01

There has recently been a revival of interest in the definition of the limit and plastic loads of a structure, particularly piping components. Recent proposals have suggested definitions based upon plastic work. In order to normalise such quantities, the plastic work associated with the limit pressure of a plain cylinder may be required. Whilst no unique value for this quantity can be obtained through analysis, this article demonstrates that classical methods allow solutions to be obtained through an empirical route and through an analysis using several approximations. This latter analysis is shown to be adequate for the applications envisaged here

Characterization of local plastic instability during deformation of D9 clad tubes for fast breeder nuclear fuel by infrared thermography

International Nuclear Information System (INIS)

Kapoor, K.; Somasekhar Reddy, K.; Ramana Rao, A.V.

2009-01-01

Plastic instability is a condition occurring ahead of ductile fracture in materials subjected to deformation. This condition is a warning before the final failure occurring by ductile fracture. This condition manifests during necking in tensile deformation and crack propagation during plastic deformation. Early detection of such a condition can prevent permanent damage. In the present case, a novel approach by use of infrared thermography is proposed for characterization of local plastic instability during deformation of D9 clad tube material. Infrared thermal imaging was used for acquiring data during deformation by plug drawing, pilgering and uniaxial tension tests. The local plastic instability due to propagation of defects was monitored by use of Taylor-Quinney coefficient (β). Where, the Taylor-Quinney coefficient is defined as the ratio of plastic work converted into heat, which was evaluated by using thermal imaging data. In our study, an analysis of the heat generated during the process of plastic deformation was measured thereby correlating the abnormal temperature variations with the defects appearing in the material. In our study, an analysis of the heat generated during the process of plastic deformation was measured thereby correlating the abnormal temperature variations with the defects appearing in the material

Heating tubes of cross-linked polyethylene

International Nuclear Information System (INIS)

Knoeppler, H.; Hoffmann, M.

1981-01-01

Oxygen permeability of plastic tubes for floor heating systems was measured as a function of the reduced oxygen content of water in plastic tubes at a flow rate of 0.5 m/s and a temperature of 30 0 C and as a function of oxygen uptake of low-oxygen water in floor heating tubes. Pipes of VEP, periodically cross-linked polyethylene (Engels process), polypropylene copolymeride, and polybutene were compared. The permeability of periodically cross-linked polyethylene is twice as high as that of VEP. Measurements, results, and consequences for floor heating systems are discussed. (KH) [de

Laser-Doppler vibrating tube densimeter for measurements at high temperatures and pressures

International Nuclear Information System (INIS)

Aida, Tsutomu; Yamazaki, Ai; Akutsu, Makoto; Ono, Takumi; Kanno, Akihiro; Hoshina, Taka-aki; Ota, Masaki; Watanabe, Masaru; Sato, Yoshiyuki; Smith, Richard L. Jr.; Inomata, Hiroshi

2007-01-01

A laser-Doppler vibrometer was used to measure the vibration of a vibrating tube densimeter for measuring P-V-T data at high temperatures and pressures. The apparatus developed allowed the control of the residence time of the sample so that decomposition at high temperatures could be minimized. A function generator and piezoelectric crystal was used to excite the U-shaped tube in one of its normal modes of vibration. Densities of methanol-water mixtures are reported for at 673 K and 40 MPa with an uncertainty of 0.009 g/cm 3

Evaporation of R134a in a horizontal herringbone microfin tube: heat transfer and pressure drop

Energy Technology Data Exchange (ETDEWEB)

Wellsandt, S; Vamling, L [Chalmers University of Technology, Gothenburg (Sweden). Department of Chemical Engineering and Environmental Science, Heat and Power Technology

2005-09-01

An experimental investigation of in-tube evaporation of R134a has been carried out for a 4 m long herringbone microfin tube with an outer diameter of 9.53 mm. Measured local heat transfer coefficients and pressure losses are reported for evaporation temperatures between -0.7 and 10.1 {sup o}C and mass flow rates between 162 and 366 kg m{sup -2} s{sup -1}. Results from this work are compared to experimental results from literature as well as predicted values from some available helical microfin correlations. Differences in heat transfer mechanisms between helical and herringbone microfin tubes are discussed, as heat transfer coefficients in the investigated herringbone tube tend to peak at lower vapour qualities compared to helical microfins. Correlations developed for helical microfin tubes generally predict experimental values within {+-}30% for vapour qualities below 50%. However, at higher qualities none of the correlations are able to reflect the early peak of heat transfer coefficients. Predicted pressure gradients reproduce measured values in general within {+-}20%. (author)

Some engineering aspects of the investigation into the cracking of pressure tubes in the Pickering reactors

International Nuclear Information System (INIS)

Ross-Ross, P.A.; Towgood, G.R.; Hunter, T.A.

1976-01-01

In August 1974, Pickering Unit 3 (514 MWe) was shutdown for a period of 8 months because of cracks in 17 of the 390 pressure tubes. The cracks were a result of incorrect installation procedures during construction. Improper positioning of the rolling tool used to join the Zr-2.5 wt% Nb pressure tube to the end fitting produced very high residual tensile stresses. High stresses in combination with periods with the tubes cold caused the cracking. Crack propagation was by fracture of hydrides which are brittle when cold. Subsequent investigation confirmed that properly rolled joints are not susceptible to such cracking. The resources of Canadian industry, Ontario Hydro and Atomic Energy of Canada were coordinated to find engineering solutions to the crack program. The defective tubes were removed from reactor, thoroughly examined to identify the cause of the cracks, and thoroughly tested to prove safety. Non-destructive techniques were quickly adopted for inspection of tubes in Pickering. Tools and procedures for retubing the 17 channels were prepared and Pickering Unit 3 was returned to service at the end of March 1975. (author)

Analytical tools and methodologies for evaluation of residual life of contacting pressure tubes in the early generation of Indian PHWRs

International Nuclear Information System (INIS)

Sinha, S.K.; Madhusoodanan, K.; Rupani, B.B.; Sinha, R.K.

2002-01-01

In-service life of a contacting Zircaloy-2 pressure tube (PT) in the earlier generation of Indian PHWRs, is limited mainly due to the accelerated hydrogen pick-up and nucleation and growth of hydride blister(s) at the cold spot(s) formed on outside surface of pressure tube as a result of its contact with the calandria tube (CT). The activities involving development of the analytical models for simulating the degradation mechanisms leading to PT-CT contact and the methodologies for the revaluation of their safe life under such condition form the important part of our extensive programme for the life management of contacting pressure tubes. Since after the PT-CT contact, rate of hydrogen pick-up and nucleation and growth of hydride blisters govern the safe residual life of the pressure tube, two analytical models (a) hydrogen pick-up model ('HYCON') and (b) model for the nucleation and growth of hydride blister at the contact spot ('BLIST -2D') have been developed in-house to estimate the extent of degradation caused by them. Along with them, a methodology for evaluation of safe residual life has also been formulated for evaluating the safe residual life of the contacting channels. This paper gives the brief description of the models and the methodologies relevant for the contacting Zircaloy-2 pressure tubes. (author)

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.

Properties of an irradiated heat-treated Zr-2.5Nb pressure tube removed from the NPD reactor

Energy Technology Data Exchange (ETDEWEB)

Chow, C.K. [Atomic Energy of Canada Limited, Pinawa, Manitoba (Canada); Coleman, C.E. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Koike, M.H. [Power Reactor and Nuclear Fuel Development Corp., O-Arai Engineering Centre, O-Arai (Japan); Causey, A.R.; Ells, C.E.; Hosbons, R.R.; Sagat, S.; Urbanic, V.F.; Rodgers, D.K

1997-07-01

Some pressure tubes in reactors moderated by heavy water have been made from heat-treated (HT) Zr-2.5Nb. One such tube was removed from the NPD nuclear reactor after 20 years of operation. An extensive program was carried out jointly by AECL and PNC to evaluate the condition and properties of this pressure tube. The investigations include irradiation creep, tensile, corrosion, delayed hydride cracking (DHC), fatigue, and fracture properties. Results show that: (I) the in-reactor elongation rate is much lower and the transverse strain rates are slightly larger than in cold-worked (CW) Zr-2.5Nb tubes; (2) the tensile properties, hydrogen pickup, threshold stress intensity factor for DHC initiation, DHC velocity, and fatigue crack growth rates were similar to those of the CW Zr-2.5Nb material; (3) the fracture toughness of this tube, as measured by curved compact toughness specimens and burst tests, is slightly higher than the CW tubes. The results were also compared with other heat-treated Zr-2.5Nb materials irradiated in the Fugen reactor. The tube was in excellent condition when removed from the reactor and would have been satisfactory for further service. (author)

Numerical study of pressure drop and heat transfer from circular and cam-shaped tube bank in cross-flow of nanofluid

International Nuclear Information System (INIS)

Mirabdolah Lavasani, Arash; Bayat, Hamidreza

2016-01-01

Highlights: • Flow around non-circular and circular shaped tube bank is studied. • Effect of using Al_2O_3-water nanofluid on flow and heat transfer is discussed. • Tubes are with in-line and staggered arrangement. • Pressure drop of non-circular tube is noticeably lower that circular tube. - Abstract: Flow and heat transfer of nanofluid inside circular and cam-shaped tube bank is studied numerically. Reynolds number for cam-shaped tube bank is defined based on equivalent diameter of circular tube and varies in range of 100 ⩽ Re_D ⩽ 400. Nanofluid is made by adding Al_2O_3 nanoparticle with volume fraction of 1–7% to pure water. Results show using nanofluid results in higher heat transfer rate for both circular tube bank and cam-shaped tube bank. Also, staggered arrangement has higher heat transfer for both circular and cam-shaped tube bank. Pressure drop from cam-shaped tube bank is substantially lower than circular tube bank for all range of Reynolds number and volume fraction.

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

International Nuclear Information System (INIS)

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

Study the Effect of Nasal Obstruction Surgery (Septoplasty on Eustachian Tube Function and Middle Ear Pressure

Directory of Open Access Journals (Sweden)

Rahim Davari

2014-12-01

Full Text Available Background & objectives: Deviation of the nasal septum is a common cause of unilateral or bilateral nasal airway obstruction and may follow nasal and midface trauma. Patient complaints of airway obstruction that are consistent with intranasal physical findings often lead to septoplasty and turbinate surgery. Severe nasal septal deviation leads to complete nasal obstruction and disturbs air passage from nostrils, however the effect of septal deviation and nasal obstruction surgery (septoplasty on Eustachian tube function and middle ear pressure is controversial and isn’t clear. Whereas many of surgeons do not believe the considerable effect of septoplasty on Eustachian tube function and middle ear pressure, so they do not recommend this procedure before middle ear surgery. On the other hand, some have an idea that septoplasty has significant effect on middle ear pressure suggesting this procedure before ear surgery like tympanoplasty.   Methods: This prospective analytical-descriptive study was conducted on seventy patients from 18 to 65 years of age who underwent septoplasty due to severe septal deviation leading to nasal obstruction in Beesat Hospital during one year (2012-2013. Middle ear pressure and Eustachian tube function on the septal deviated side and contralateral side before and after septoplasty (3 to 6 months later were measured through tympanometry and Eustachian tube function test (Toynbee test. The comparison between pre- and postoperative ETF tests and middle ear pressures was assessed using Paired –T test and p-value of less than 0.05 was considered as statistically significant.   Results : This study revealed that comparison between mean values of pressure in the deviated and contralateral side has no significant statistical difference before and after septoplasty. Also comparison between Eustachian function in the deviated side and contralateral side showed no significant difference before and after septoplasty. �

Detection of steam generator tube leaks in pressurized water reactors

International Nuclear Information System (INIS)

Roach, W.H.

1984-11-01

This report addresses the early detection of small steam generator tube leaks in pressurized water reactors. It identifies physical parameters, establishes instrumentation performance goals, and specifies sensor types and locations. It presents a simple algorithm that yields the leak rate as a function of known or measurable quantities. Leak rates of less than one-tenth gram per second should be detectable with existing instrumentation

Some observations of the pressure distribution in a tube bank for conditions of self generated acoustic resonance

International Nuclear Information System (INIS)

Fitzpatrick, J.A.; Donaldson, I.S.; McKnight, W.

1979-01-01

The results for mean and fluctuating pressure distributions around tubes in an in-line tube bank are presented for both non-resonant and self-excited acoustic standing wave resonant flow regimes. It is readily deduced that the nature of the flow in the bank is dramatically altered with the onset of acoustic resonance. The velocity gradients which appear across the bank with the onset of resonance would suggest regions of flow recirculation in the bank although no evidence of this was found. The spectra of fluctuating pressure on the duct roof in the bank and on tubes deep in the bank exhibited coherent peaks only during resonance. (author)

Thermally induced processes in mixtures of aluminum with organic acids after plastic deformations under high pressure

Science.gov (United States)

Zhorin, V. A.; Kiselev, M. R.; Roldugin, V. I.

2017-11-01

DSC is used to measure the thermal effects of processes in mixtures of solid organic dibasic acids with powdered aluminum, subjected to plastic deformation under pressures in the range of 0.5-4.0 GPa using an anvil-type high-pressure setup. Analysis of thermograms obtained for the samples after plastic deformation suggests a correlation between the exothermal peaks observed around the temperatures of degradation of the acids and the thermally induced chemical reactions between products of acid degradation and freshly formed surfaces of aluminum particles. The release of heat in the mixtures begins at 30-40°C. The thermal effects in the mixtures of different acids change according to the order of acid reactivity in solutions. The extreme baric dependences of enthalpies of thermal effects are associated with the rearrangement of the electron subsystem of aluminum upon plastic deformation at high pressures.

Elastic-plastic behaviour of thick-walled containers considering plastic compressibility

International Nuclear Information System (INIS)

Betten, J.; Frosch, H.G.

1983-01-01

In this paper the elastic-plastic behaviour of thick-walled pressure vessels with internal and external pressure is studied. To describe the mechanical behaviour of isotropic, plastic compressible materials we use a plastic potential which is a single-valued function of the principle stresses. For cylinders and spheres an analytic expression for the computation of stresses and residual stresses is specified. Afterwards the strains are calculated by using the finite difference method. Some examples will high-light the influence of the plastic compressibility on the behaviour of pressure vessels. (orig.) [de

Boiling heat transfer and dryout in helically coiled tubes under different pressure conditions

International Nuclear Information System (INIS)

Chung, Young-Jong; Bae, Kyoo-Hwan; Kim, Keung Koo; Lee, Won-Jae

2014-01-01

Highlights: • Heat transfer characteristics and dryout for helically coiled tube are performed. • A boiling heat transfer tends to increase with a pressure increase. • Dryout occurs at high quality test conditions investigated. • Steiner–Taborek’s correlation is predicted well based on the experimental results. - Abstract: A helically coiled once-through steam generator has been used widely during the past several decades for small nuclear power reactors. The heat transfer characteristics and dryout conditions are important to optimal design a helically coiled steam generator. Various experiments with the helically coiled tubes are performed to investigate the heat transfer characteristics and occurrence condition of a dryout. For the investigated experimental conditions, Steiner–Taborek’s correlation is predicted reasonably based on the experimental results. The pressure effect is important for the boiling heat transfer correlation. A boiling heat transfer tends to increase with a pressure increase. However, it is not affected by the pressure change at a low power and low mass flow rate. Dryout occurs at high quality test conditions investigated because a liquid film on the wall exists owing to a centrifugal force of the helical coil

Pre and post garter spring repositioning ultrasonic inspection of pressure tubes

International Nuclear Information System (INIS)

Desimone, C.; Katchadjian, P.; Tacchia, Mauricio

1997-01-01

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) [es

Condensation heat transfer and pressure drop of R-410A in flat aluminum multi-port tubes

Science.gov (United States)

Kim, Nae-Hyun

2018-02-01

Brazed heat exchangers with aluminum flat multi-port tubes are being used as condensers of residential air-conditioners. In this study, R-410A condensation tests were conducted in four multi-port tubes having a range of hydraulic diameter (0.78 ≤ Dh ≤ 0.95 mm). The test range covered the mass flux from 100 to 400 kg/m2 s and the heat flux at 3 kW/m2, which are typical operating conditions of residential air conditioners. Results showed that both the heat transfer coefficient and the pressure drop increased as the hydraulic diameter decreased. The effect of hydraulic diameter on condensation heat transfer was much larger than the predictions of existing correlations for the range of investigation. Comparison of the data with the correlations showed that some macro-channel tube correlations and mini-channel tube correlations reasonably predicted the heat transfer coefficient. However, macro-channel correlations highly overpredicted the pressure drop data.