Sample records for calandrias

  1. Critical heat flux variations on CANDU calandria tube surface

    Energy Technology Data Exchange (ETDEWEB)

    Behdadi, A.; Luxat, J.C., E-mail:, E-mail: [McMaster Univ., Engineering Physics Dept., Hamilton, Ontario (Canada)


    Heavy water moderator surrounding each fuel channel is one of the important safety features in CANDU reactors since it provides an in-situ passive heat sink for the fuel in situations where other engineered means of heat removal from fuel channels have failed. In a critical break LOCA scenario, fuel cooling becomes severely degraded due to rapid flow reduction in the affected flow pass of the heat transport system. This can result in pressure tubes experiencing significant heat-up during early stages of the accident when coolant pressure is still high, thereby causing uniform thermal creep strain (ballooning) of the pressure tube (PT) into contact with its calandria tube (CT). The contact of the hot PT with the CT causes rapid redistribution of stored heat from the PT to CT and a large heat flux spike from the CT to the moderator fluid. For conditions where subcooling of the moderator fluid is low, this heat flux spike can cause dryout of the CT. This can detrimentally affect channel integrity if the CT post-dryout temperature becomes sufficiently high to result in continued thermal creep strain deformation of both the PT and the CT. The focus of this work is to develop a mechanistic model to predict Critical Heat Flux (CHF) on the CT surface following a contact with its pressure tube. A mechanistic CHF model is applied based on a concept of wall dry patch formation, prevention of rewetting and subsequent dry patch spreading. Results have been compared to an empirical correlation and a good agreement has been obtained. The model has been used to predict the spatial variation of CHF over a cylinder with dimensions of CANDU CT. (author)

  2. The thermal interaction of a buoyant plume from a calandria tube with an oblique jet

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    Rossouw, D.J.; Atkins, M.D.; Beharie, K. [Nuclear Science Division, School of Mechanical & Aeronautical Engineering, University of the Witwatersrand, Johannesburg (South Africa); Kim, T., E-mail: [Nuclear Science Division, School of Mechanical & Aeronautical Engineering, University of the Witwatersrand, Johannesburg (South Africa); Rhee, B.W.; Kim, H.T. [Severe Accident and PHWR Safety Research Division, Korea Atomic Energy Research Institute, Daejun (Korea, Republic of)


    Highlights: • A crucial role of relative orientation between mixed convection modes is observed. • The extent of thermal interaction strongly depends on the relative orientation. • Coolant flow is substantially diffused by a buoyant plume if counter-acting. • Slightly oblique coolant flow to the gravitational axis provides the best cooling. - Abstract: Severe reactor core damage may occur from fuel channel failure as a consequence of excessive heat emitted from calandria tubes (CTs) in a pressurised heavy water (D{sub 2}O) reactor (CANDU). The heating of the CTs is caused by creep deformation of the pressure tubes (PTs), which may be ballooning or sagging depending on the internal pressure of the PTs. The deformation of the pressure tube is due to overheating as a result of a loss-of-coolant accident (LOCA) and emergency core cooling system (ECCS) failure. To prevent the exacerbation of the LOCA, circulating D{sub 2}O in the moderator tank may be utilized by forming a secondary jet that externally cools the individual CTs. The buoyant plume develops around the CTs and interacts with the secondary jet at a certain oblique angle with respect to the gravitational axis, depending on the spatial location of the hot calandria tubes (or the hot reactor core region). This study reports on how the local and overall heat transfer characteristics on a calandria tube where the buoyant plume develops, are altered by the obliqueness of the external secondary jet (from a co-current jet to a counter-current jet) in a simplified configuration at the jet Reynolds number of Re{sub j} = 1500 for the Archimedes number of Ar{sub D} = 0.11 and Rayleigh number of Ra{sub D} = 1.6 × 10{sup 6} (modified Rayleigh number of 3.0 × 10{sup 7}).

  3. Development of video probe system for inspection of feeder pipe support in calandria reactor

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    Cho, Jai Wan; Lee, Nam Ho; Choi, Young Soo


    There are 760 feederpipes, which they are connected to inlet/outlet of the 380 pressure tube channels on the front of the calandria, in CANDU-type Reactor of Wolsung Nuclear Power Plant. As an ISI(In-Service Inspection) and PSI (Post- Service Inspection) requirements, maintenance activities of measuring the thickness of curvilinear part of feederpipe and inspecting the feederpipe support area within calandria are needed to ensure continued reliable operation of nuclear power plant. And untrasonic probe is used to measure the thickness of curvilinear part of feederpipe, however workers are exposed to radioactivity irradiation during the measurement period. But, it is impossible to inspect feederpipe support area thoroughlv because of narrow and confined accessibility, that is, an inspection space between the pressure tube channels is less than 100mm and pipes in feederpipe support area are congested. And also, workers involved in inspecting feederpipe support area are under the jeopardy of high-level radiation exposure. Concerns about sliding home, which make the move of feederpipe connected to pressure tube channel smooth as pressure tube expands and contracts in its axial direction, stuck to feederpipe support and some of the structural components have made necessary the development of video inspection probe system with narrow and confined accessibility to observe and inspect feederpipe support area more close. Using video inspection probe system, it is possible to inspect and repair abnormality of feederpipe support connected to pressure tube channels of the calandria more accurate and quantative than naked eye. Therefore, that will do much for ensuring safety of CANDU-type nuclear power plant.

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

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


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

  5. Mixed convection around calandria tubes in a ¼ scale CANDU-6 moderator circulation tank

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    Atkins, M.D.; Rossouw, D.J.; Boer, M. [Nuclear Science Division, School of Mechanical and Aeronautical Engineering, University of the Witwatersrand, Johannesburg (South Africa); Kim, T., E-mail: [Nuclear Science Division, School of Mechanical and Aeronautical Engineering, University of the Witwatersrand, Johannesburg (South Africa); Rhee, B.W.; Kim, H.T. [Severe Accident and PHWR Safety Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)


    Highlights: • A secondary jet is formed at a stagnation region and is directed towards the center of the MCT. • The secondary jet undergoes the significant dissipation and mixing due to calandria tubes (CTs). • Its cooling effectiveness is reduced on the CTs in the bottom of the MCT. • With forced convection dominance, peak heat transfer is on the upper CT surface. • With natural convection dominance, peak heat transfer is on the lower CT surface. - Abstract: This study experimentally characterizes mixed convection around calandria tubes (CTs) in a ¼ scale CANDU-6 moderator circulation tank (MCT) that uses air as the working fluid. In a full scale CANDU-6 reactor that undergoes a postulated dual failure with a loss-of-coolant accident without the emergency core cooling system available, mixed convection heat transfer occurs around the CTs. The cooling effectiveness of the moderator is diminished as an emergency heat sink if overheating eventually leads to film boiling. To prevent the onset of film boiling, local sub-cooling margins of the moderator needs to be maintained or else the critical heat flux should be increased. Circulating the moderator which interacts with the overheated CTs increases the heat transfer into the moderator which may suppress film boiling. The present experimental results demonstrate that the cooling effectiveness of the circulating moderator, in particular the secondary jet, is attenuated substantially as it is convected away from the inner wall towards the center of the MCT. The momentum of the secondary jet is diffused through the CTs. At a low jet Reynolds number, the secondary jet becomes ineffective so that some overheated CTs positioned in the other half of the MCT are cooled only by natural convection.

  6. Analyses on fluid flow and heat transfer inside Calandria vessel of CANDU-6 using CFD

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Manwoong [Korea Institute of Nuclear Safety, 19 Guseong-dong, Yuseong-gu, Daejeon 305-338 (Korea, Republic of)]. E-mail:; Yu, Seon-Oh [Korea Institute of Nuclear Safety, 19 Guseong-dong, Yuseong-gu, Daejeon 305-338 (Korea, Republic of); Kim, Hho-Jung [Korea Institute of Nuclear Safety, 19 Guseong-dong, Yuseong-gu, Daejeon 305-338 (Korea, Republic of)


    In a CANada Deuterium Uranium (CANDU) reactor, fuel channel integrity depends on the coolability of the moderator as an ultimate heat sink under transient conditions such as a loss of coolant accident (LOCA) with a coincidence of a loss of emergency core cooling (LOECC), as well as a normal operating condition. This study presents the assessments of moderator thermal-hydraulic characteristics in the normal operating condition and one transient condition for CANDU-6 reactors, using a general purpose three-dimensional computational fluid dynamics code. This study consists of two steps. First, an optimized calculation scheme is obtained by many-sided comparisons of the predicted results with the related experimental data, and by evaluating the fluid flow and temperature distributions. Then, in the second step, with the optimized scheme, the analyses for real CANDU-6 of normal operating condition and transition condition have been performed. The present model has successfully predicted the experimental results and also reasonably assessed the thermal-hydraulic characteristics of the real CANDU-6 with 380 fuel channels. Flow regime map with major parameters representing the flow pattern inside Calandria vessel has also proposed to be used as operational and/or regulatory guidelines.

  7. Mechanistic modeling of pool film-boiling and quench on a Candu calandria tube following a critical break LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, J.T.; Luxat, J.C. [McMaster University, A315 JHE Building, 1280 Main St.W. Hamilton, ON, L8S 4L7 (Canada)


    Following a postulated critical LBLOCA a pressure tube (PT) can experience creep deformation and balloon uniformly into contact with the calandria tube (CT). The resultant heat flux to CT is high as stored heat is transferred out of the hot PT. This heat flux can cause dryout on the outer surface of the CT and establish film boiling. This paper presents a model of buoyancy-driven natural convection film boiling on the outside of a horizontal tube with diameter relevant to a Candu CT (approximately 13 cm). A second order, non-linear and non-homogeneous ODE for vapour film thickness has been derived. The variation of steady state vapour film thickness prior to quench as a function of subcooling temperature, wall superheat, and incident heat flux is examined. The CT outer surface heatup rate and effective film boiling heat transfer coefficient from the model are in good agreement with available experimental data. (authors)

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

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    Hyoung Tae Kim


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


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    Constantin POPESCU


    Full Text Available The authors contribution to this paper is to present a concept solution of a remote control robot (RCR used for the horizontal fuel channels pressure tube decommissioning in the CANDU nuclear reactor. The authors highlight in this paper, few details of geometry, operations, constraints by kinematics and dynamics of the robot movement outside of the reactor fuel channel. Outside operations performed has as the main steps of dismantling process the followings: positioning front of Calandria structure at the fuel channel to be decommissioned, coupling and locking to the End Fitting (EF, sorting and storage extracted items in the safe container. All steps are performed in automatic mode. The remote control robot (RCR represents a safety system controlled by sensors and has the capability to analyze any error registered and decide next activities or abort the outside decommissioning procedure in case of any risk rise in order to ensure the environmental and workers protection.

  10. Numerical Analysis on the Calandria Tubes in the Moderator of a Heavy Water Reactor Using OpenFOAM and Other Codes (United States)

    Chang, Se-Myong; Kim, Hyoung Tae


    CANDU, a prototype of heavy water reactor is modeled for the moderator system with porous media buoyancy-effect heat-transfer turbulence model. OpenFOAM, a set of C++ classes and libraries developed under the object-oriented concept, is selected as the tool of numerical analysis. The result from this computational code is compared with experiments and other commercial code data through ANSYS-CFX and COMSOL Multi-physics. The three-dimensional code concerning buoyancy force, turbulence, and heat transfer is tested and shown to be successful for the analysis of thermo-hydraulic system of heavy water reactors.


    Directory of Open Access Journals (Sweden)

    Constantin POPESCU


    Full Text Available The authors contribution to this paper is to present a concept solution of a remote control robot (RCR used for decommissioning of the horizontal fuel channels pressure tube in the CANDU nuclear reactor. In this paper the authors highlight few details of geometry, operations, constraints by kinematics and dynamics of the robot movement inside of the reactor fuel channel. Inside operations performed has as the main steps of dismantling process the followings: unblock and extract the channel closure plug (from End Fitting - EF, unblock and extract the channel shield plug (from Lattice Tube - LT, cut the ends of the pressure tube, extract the pressure tube and cut it in small parts, sorting and storage extracted items in the safe robot container. All steps are performed in automatic mode. The remote control robot (RCR represents a safety system controlled by sensors and has the capability to analyze any error registered and decide next activities or abort the inside decommissioning procedure in case of any risk rise in order to ensure the environmental and workers protection.


    Oppenheimer, E.D.; Weisberg, R.A.


    This patent relates to a barrier system for a sodium heavy water reactor capable of insuring absolute separation of the metal and water. Relatively cold D/sub 2/O moderator and reflector is contained in a calandria into which is immersed the fuel containing tubes. The fuel elements are cooled by the sodium which flows within the tubes and surrounds the fuel elements. The fuel containing tubes are surrounded by concentric barrier tubes forming annular spaces through which pass inert gases at substantially atmospheric pressure. Header rooms above and below the calandria are provided for supplying and withdrawing the sodium and inert gases in the calandria region. (AEC)

  13. Computational fluid dynamics analysis of the Canadian deuterium uranium moderator tests at the Stern Laboratories Inc.

    Directory of Open Access Journals (Sweden)

    Hyoung Tae Kim


    Full Text Available A numerical calculation with the commercial computational fluid dynamics code CFX-14.0 was conducted for a test facility simulating the Canadian deuterium uranium moderator thermal–hydraulics. Two kinds of moderator thermal–hydraulic tests at Stern Laboratories Inc. were performed in the full geometric configuration of the Canadian deuterium uranium moderator circulating vessel, which is called a calandria tank, housing a matrix of horizontal rod bundles simulating calandria tubes. The first of these tests is the pressure drop measurement of a cross flow in the horizontal rod bundles. The other is the local temperature measurement on the cross section of the horizontal cylinder vessel simulating the calandria system. In the present study, the full geometric details of the calandria tank are incorporated in the grid generation of the computational domain to which the boundary conditions for each experiment are applied. The numerical solutions are reviewed and compared with the available test data.

  14. Computational fluid dynamics analysis of the Canadian deuterium uranium moderator tests at the Stern Laboratories Inc.


    Kim, Hyoung Tae; Chang, Se-Myong


    A numerical calculation with the commercial computational fluid dynamics code CFX-14.0 was conducted for a test facility simulating the Canadian deuterium uranium moderator thermal–hydraulics. Two kinds of moderator thermal–hydraulic tests at Stern Laboratories Inc. were performed in the full geometric configuration of the Canadian deuterium uranium moderator circulating vessel, which is called a calandria tank, housing a matrix of horizontal rod bundles simulating calandria tubes. The first ...

  15. OpenFOAM Analysis of CANDU-6 Moderator Flow

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


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

  16. Development of computer code for level 2 PSA of CANDU plant

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    Kim, Dong Ha; Jin, Young Ho; Park, Soo Yong; Song, Yong Mann


    The objective of this study is to develop a computer code, ISAAC (Integrated Severe Accident Analysis Code for CANDU Plants), which could handle the severe accident phenomena at Wolsong plants. The MAAP 4 computer code, which is a fast running code with state-of-the-art models, has been used as a basis. The new features in ISAAC are : (1) core module describing the configuration of core (horizontal fuel channel, calandria tubes, pressure tubes, fuel bundles, etc.) and core behavior such as channel heatup, oxidation, sagging and relocation of molten corium into the calandria, (2) the PHTS module modeling water/gas pressure and temperature, water level, break flow rates in the two separated figure-of-eight primary loop, (3) calandria transfer rates between calandria and water in the calandria bottom, (4) special safety systems including emergency core cooling system (ECCS), dousing sprays, moderator cooling system, shield cooling system, and over-pressure protection system which mitigate the severe accidents. The code has been tested for the various representative severe accidents. The results were reasonable from the view point of accident progression and its phenomena. Therefore, ISAAC can be said to have the basic capability to simulate the severe accidents in the Wolsong plants. (author). 28 refs., 7 tabs., 140 figs.

  17. Numerical analysis of debris melting phenomena during late phase CANDU 6 severe accident

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    Nicolici, Stefan, E-mail: [University Politehnica of Bucharest, Power Plant Engineering Faculty, Splaiul Independentei Street, No. 313, RO 060042, Sector 6, Bucharest (Romania); Dupleac, Daniel; Prisecaru, Ilie [University Politehnica of Bucharest, Power Plant Engineering Faculty, Splaiul Independentei Street, No. 313, RO 060042, Sector 6, Bucharest (Romania)


    Highlights: Black-Right-Pointing-Pointer We used Ansys Fluent to simulate debris melt phenomena for a CANDU reactor. Black-Right-Pointing-Pointer Two models were employed: Large Eddy Simulation and PECM tool. Black-Right-Pointing-Pointer The PECM approach is comparable within 10% with LES model. Black-Right-Pointing-Pointer The debris can be well contained inside the calandria during a severe accident. - Abstract: The objective of this paper was to study the phase change of the debris formed on the CANDU 6 calandria bottom in a postulated accident sequence. The molten pool and crust formation were studied employing the Ansys Fluent code. Two models using Large Eddy Simulation and Phase Change Effective Convectivity Model (PECM) predict the conjugate, radiative and convective heat transfer inside and from the corium pool. LES require a very fine grid to capture the crust formation and the free convection flow. This aspect (fine mesh requirement) correlated with the long transient has imposed the use of a slice from the 3D calandria geometry in order not to exceed the computing resources. From the safety point of view it is very important to maintain a heat flux through the wall below the critical heat flux assuring the integrity of the calandria vessel, thus the paper results include temperature profiles and heat fluxes through calandria wall. The results predicted by the PECM approach are comparable within 10% with those obtained by the computational fluid dynamics method, proving that the PECM can be used to perform further sensitivity studies.

  18. Salida de campo a Fuensaldaña (Valladolid) el 20 de julio de 1956


    Valverde Gómez, José Antonio, 1926-2003


    Salida de campo a Fuensaldaña (Valladolid) el 20 de julio de 1956, de la que se anotaron observaciones sobre las siguientes aves: Melanocorypha calandra (Calandria), Pterocles alchata (Ganga ibérica) y Pterocles orientalis (Ganga ortega). Se incluyen dos pequeños dibujos a rotulador. Field trip to Fuensaldaña (Valladolid) the 20th of July of 1956, of which there were noted observations about the following birds: Melanocorypha calandra (Calandria Lark), Pterocles alchata (Pin-tailed Sandgro...

  19. Welding procedures used in the fabrication of fuel elements for the DON Reactor exponential experiment; La soldadura en la fabricacion de elementos combustibles destinados a una experiencia exponencial

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    Diaz Beltran, A.; Jaraiz Franco, E.; Rivas Diaz, M. de las


    This exponential experiment required 74 units (37 loaded with UO{sub 2} and 37 with UC) to simulate the Reactor fuel channels. Each unit was enclosed in a tube similar to the calandria ones. It contained the pressure tube, the shroud and the 19 rods cluster. Within the pressure tube, in touch with the elements, was the organic liquid. (Author)

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

    Energy Technology Data Exchange (ETDEWEB)

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


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

  1. Structural safety of coolant channel components under excessively high pressure tube diametral expansion rate at garter spring location

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    Aravind, M. [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India); Sinha, S.K., E-mail: [Reactor Engineering Division, Bhabha Atomic Research Centre, Trombay, Mumbai 400 085 (India)


    Structural safety of coolant channel assembly in the event of high diametral expansion of pressure tube in a 220 MWe pressurised heavy water reactor was investigated using axisymmetric and 3-D finite element models. The axisymmetric analyses were performed and stresses were evaluated for pressure tube, girdle wire and calandria tube at different point of time for diametral expansion rates of 0.2%, 0.25% and 0.3% per year of the pressure tube inside diameter. The results of this study indicated that for the case of 0.3% per year of diametral expansion rate (worst case scenario), occurrence of complete circumferential interference of garter spring with calandria tube at the location of maximum expansion would take place much earlier at around 14 years or 4.2% of the total expansion of pressure tube as opposed to its anticipated design life (30 years). This fact was further corroborated by 3-D finite element analysis performed for the actual assembly configuration under actual loadings. The latter analysis revealed that net section yielding of calandria tube occurs in just 1 year after the occurrence of total circumferential interference between calandria tube and garter spring spacer. It has also been observed that the maximum stress intensity in girdle wire does not increase beyond the ultimate tensile strength even when maximum stress intensity in calandria tube reaches its yield strength. These analyses also revealed that the structural as well as functional integrity of pressure tube and the garter spring is not affected as result of this interference.

  2. CANDU in-reactor quantitative visual-based inspection techniques (United States)

    Rochefort, P. A.


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

  3. Uncertainty Analysis of the Potential Hazard of MCCI during Severe Accidents for the CANDU6 Plant

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    Sooyong Park


    Full Text Available This paper illustrates the application of a severe accident analysis computer program to the uncertainty analysis of molten corium-concrete interaction (MCCI phenomena in cases of severe accidents in CANDU6 type plant. The potential hazard of MCCI is a failure of the reactor building owing to the possibility of a calandria vault floor melt-through even though the containment filtered vent system is operated. Meanwhile, the MCCI still has large uncertainties in several phenomena such as a melt spreading area and the extent of water ingression into a continuous debris layer. The purpose of this study is to evaluate the MCCI in the calandria vault floor via an uncertainty analysis using the ISAAC program for the CANDU6.

  4. Fission-product releases from a PHWR terminal debris bed

    Energy Technology Data Exchange (ETDEWEB)

    Brown, M.J.; Bailey, D.G., E-mail: [Canadian Nuclear Laboratories, Chalk River, Ontario (Canada)


    During an unmitigated severe accident in a pressurized heavy water reactor (PHWR) with horizontal fuel channels, the core may disassemble and relocate to the bottom of the calandria vessel. The resulting heterogeneous in-vessel terminal debris bed (TDB) would likely be quenched by any remaining moderator, and some of the decay heat would be conducted through the calandria vessel shell to the surrounding reactor vault or shield tank water. As the moderator boiled off, the solid debris bed would transform into a more homogeneous molten corium pool located between top and bottom crusts. Until recently, the severe accident code MAAP-CANDU assumed that unreleased volatile and semi-volatile fission products remained in the TDB until after calandria vessel failure, due to low diffusivity through the top crust and the lack of gases or steam to flush released fission products from the debris. However, national and international experimental results indicate this assumption is unlikely; instead, high- and medium-volatility fission products would be released from a molten debris pool, and their volatility and transport should be taken into account in TDB modelling. The resulting change in the distribution of fission products within the reactor and containment, and the associated decay heat, can have significant effects upon the progression of the accident and fission-product releases to the environment. This article describes a postulated PHWR severe accident progression to generate a TDB and the effects of fission-product releases from the terminal debris, using the simple release model in the MAAP-CANDU severe accident code. It also provides insights from various experimental programs related to fission-product releases from core debris, and their applicability to the MAAP-CANDU TDB model. (author)

  5. Post irradiation examination of tight fit garter springs from Indian PHWR

    Energy Technology Data Exchange (ETDEWEB)

    Dubey, J.S., E-mail:; Shriwastaw, R.S.; Kumar, Ashwini; Shah, Priti Kotak; Rath, B.N.; Kumar, Sunil; Mishra, Prerna; Alur, V.D.; Mallik, G.K.; Anantharaman, S.


    Garter springs play an important role in maintaining the annulus gap between hot pressure tubes and cold calandria tubes of PHWRs. Post irradiation examination (PIE) was carried out on the garter springs removed from Indian PHWR after around 8 and 15 Hot Operating Years (HOY). PIE studies included visual examination, dimensional measurements, metallographic examination and relevant mechanical tests. The girdle wires of these garter springs were also examined and subjected to the tension and bend tests. This paper gives the results of the PIE investigations and discusses its relevance for continued performance of garter springs in PHWRs.

  6. Fuel elements assembling for the DON project exponential experience; Montaje de los elementos combustibles para la experiencia exponencial del proyecto DON

    Energy Technology Data Exchange (ETDEWEB)

    Anca Abati, R. de


    It is described the fuel unit used in the DON exponential experience, the manufacturing installments and tools as well as the stages in the fabrication.These 74 elements contain each 19 cartridges loaded with synterized urania, uranium carbide and indium, gold, and manganese probes. They were arranged in calandria-like tubes and the process-tube. This last one containing a cooling liquid simulating the reactor organic. Besides being used in the DON reactor exponential experience they were used in critic essays by the substitution method in the French reactor AQUILON II. (Author) 6 refs.

  7. Biomass and Soil Carbon Stocks in Wet Montane Forest, Monteverde Region, Costa Rica: Assessments and Challenges for Quantifying Accumulation Rates

    Directory of Open Access Journals (Sweden)

    Lawrence H. Tanner


    Full Text Available We measured carbon stocks at two forest reserves in the cloud forest region of Monteverde, comparing cleared land, experimental secondary forest plots, and mature forest at each location to assess the effectiveness of reforestation in sequestering biomass and soil carbon. The biomass carbon stock measured in the mature forest at the Monteverde Institute is similar to other measurements of mature tropical montane forest biomass carbon in Costa Rica. Local historical records and the distribution of large trees suggest a mature forest age of greater than 80 years. The forest at La Calandria lacks historical documentation, and dendrochronological dating is not applicable. However, based on the differences in tree size, above-ground biomass carbon, and soil carbon between the Monteverde Institute and La Calandria sites, we estimate an age difference of at least 30 years of the mature forests. Experimental secondary forest plots at both sites have accumulated biomass at lower than expected rates, suggesting local limiting factors, such as nutrient limitation. We find that soil carbon content is primarily a function of time and that altitudinal differences between the study sites do not play a role.


    Directory of Open Access Journals (Sweden)



    Full Text Available The scope of this paper is to achieve the device functioning steps for the commissioning of the horizontal fuel channels of calandria vessel. The dismantling of the fuel channel is performed by one device which shall provide radiation protection during the stages of decommissioning, ensuring radiation protection of the workers. For the decommissioning operation design shall be taken to ensure all aspects of security, environmental protection during decommissioning operation steps and creating and implementing work procedures resulting from developed decommissioning plan. The fuel channel decommissioning device is designed for dismantling and extraction of the fuel channel and its components. The decommissioning operation consists of following major steps: platform with device positioning to the fuel channel to be dismantled; coupling and locking the device at the fuel channel; unblock, extract and store the channel closure plug; unblock, extract and store the channel shield plug; block and cut the middle and the end of the pressure tube; block, extract and store the end fitting; block, extract and store the half of pressure tube; mounting of the extended closing plug. The operations steps are performed by the Cutting and Extraction Device and by the extraction actuator from the device handling elements assembly. After each step of dismantling is necessary the confirmation its finalization in order to perform the next operation step. The dismantling operation steps of the fuel channel components are repeated for all the 380 channels of the reactor, from the front of calandria side (plane R as well as the rear side (plane R'.

  9. Simulating the behaviour of zirconium-alloy components in nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Coleman, C.E


    To prevent failure in nuclear components one needs to understand the interactions between adjacent materials and the changes in their physical properties during all phases of reactor operation. Three examples from CANDU reactors are described to illustrate the use of simulations that imitate complicated reactor situations. These are: swelling tests that led to a method for increasing the tolerance or Zircaloy fuel cladding to power ramps; observations of the behaviour of leaking cracks in Zr-2.5Nb pressure tubes that provide confidence in the use of leak-before-break as part of the defence against flaw development; and contact boiling tests on modifications to the surfaces of Zircaloy calandria tubes that enhance the ability of the heavy water moderator to act as a heat sink after a postulated loss-of-coolant accident. (author)

  10. Remote field eddy current testing

    Energy Technology Data Exchange (ETDEWEB)

    Cheong, Y. M.; Jung, H. K.; Huh, H.; Lee, Y. S.; Shim, C. M


    The state-of-art technology of the remote field eddy current, which is actively developed as an electromagnetic non-destructive testing tool for ferromagnetic tubes, is described. The historical background and recent R and D activities of remote-field eddy current technology are explained including the theoretical development of remote field eddy current, such as analytical and numerical approach, and the results of finite element analysis. The influencing factors for actual applications, such as the effect of frequency, magnetic permeability, receiving sensitivity, and difficulties of detection and classification of defects are also described. Finally, two examples of actual application, 1) the gap measurement between pressure tubes and calandria tube in CANDU reactor and, 2) the detection of defects in the ferromagnetic heat exchanger tubes, are described. The future research efforts are also included.

  11. Thermodynamic Modelling of an Ejector with Compressible Flow by a One-Dimensional Approach

    Directory of Open Access Journals (Sweden)

    Claude Chacoux


    Full Text Available The purpose of this study is the dimensioning of the cylindrical mixing chamber of a compressible fluid ejector used in particular in sugar refineries for degraded vapor re‑compression at the calandria exit, during the evaporation phase. The method used, known as the “integral” or “thermodynamic model”, is based on the model of the one‑dimensional isentropic flow of perfect gases with the addition of a model of losses. Characteristic curves and envelope curves are plotted. The latter are an interesting tool from which the characteristic dimensions of the ejector can be rapidly obtained for preliminary dimensioning (for an initial contact with a customer for example. These ejectors, which were specifically designed for the process rather than selected from a catalog of standard devices, will promote energy saving.

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

    Energy Technology Data Exchange (ETDEWEB)

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


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

  13. Generic assessment of tight-fitting annulus spacer mobility

    Energy Technology Data Exchange (ETDEWEB)

    Robertson, J. [Nuclear Safety Solutions Limited, Toronto, Ontario (Canada)]. E-mail:; Micuda, L. [Bruce Power, Toronto, Ontario (Canada)]. E-mail:; Van Den Brekel, N. [Ontario Power Generation, Pickering, Ontario (Canada)]. E-mail:


    This paper provides a generic assessment of the mobility of tight-fitting fuel channel annulus spacers in OPG and Bruce Power nuclear units. This assessment is applicable to all tight-fitting annulus spacers, including those used in the original fuel channel installation (Darlington Units 1-4, retubed Pickering Units 1-4, and Bruce Unit 8) and as a result of single fuel channel replacements (SFCR) (Pickering Units 5-8, Bruce Units 3-7). Tight-fitting annulus spacers were designed not to move. Pressure tube to calandria tube contact analyses, and the associated blister susceptibility assessments, have assumed that these tight-fitting spacers remain at the pre-service installed locations. Given the importance of this assumption, the technical basis for the expectation that tight-fitting annulus spacers do not move significantly from their pre-service locations, relative to the pressure tube, was reviewed in detail. The review also assessed the inspection data, comparing spacer locations from in-service and pre-service inspections. The review has concluded that tight-fitting spacers do not move sufficiently to necessitate a postulated spacer movement in fuel channel contact analyses. The paper describes the background of this issue, briefly reviews the experimental programs used to qualify the positional stability of the tight-fitting spacer design, and evaluates the current database of in-service spacer location inspection information to demonstrate that no significant movement relative to the pressure tube has been observed. (author)

  14. Development of thermal-hydraulic models for the safety evaluation of CANDU reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jae Young; Jung, Yun Sik; Hwang, Gi Suk; Kim, Nam Seok [Handong Univ., Pohang (Korea, Republic of); No, Hee Cheon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)


    The objective of the present research is to evaluate the safety analysis for CANDU and to improve the Horizontal Stratification Entrainment Model (HSEM) of RELAP5/MOD3.3. This report includes two items the one is the experimental study of entrainment at horizontal pipe with {+-} 36 .deg. C , {+-} 72 .deg. C branch pies, the other is the model improvement of the moderator heat sink in the Calandria. The off-take experiments on onset of entrainment and branch quality were investigated by using water and air as working fluid, and the experimental data were compared by the previous correlations. The previous correlations could not expect experimental results, thus the weak points of the previous correlations were investigated. The improvement of the previous model continues as the next year research. The thermal hydraulic scaling analysis of SPEL, STERN and ideal linear scaling analysis have been studied. As a result, a new scaling method were needed to design a new experimental facility (HGU). A new scaling method with 1/8 length scale was applied. From these results, the thermal hydraulic model for CFD code simulation was designed and test apparatus has been made. The moderator temperature distribution experiments and CFD code simulation will be continued in next year.

  15. Detection and repositioning of tight fighting annulus spacers in CANDU® fuel channels

    Energy Technology Data Exchange (ETDEWEB)

    Hersak, G.; Kittmer, A. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Goszczynski, J. [Candu Energy Inc., Mississauga, Ontario (Canada); Kazimer, D. [Bruce Power, Tiverton, Ontario (Canada)


    The latest generation of CANDU® reactors has been constructed with tight-fitting annulus spacers to maintain the annular gap between the inner pressure tubes and the outer calandria tubes. These spacers cannot be detected and repositioned with the existing Spacer Location and Repositioning (SLAR) process, which is designed to work with loose-fitting annulus spacers. There is currently no established technology to detect and reposition tight-fitting annulus spacers. Atomic Energy of Canada Limited has been performing research and development to locate and move tight-fitting annulus spacers using Modal Detection and Repositioning (MODAR) technology since 2005 and is currently working in collaboration with Candu Energy and Bruce Power on a production system to be deployed for an In-reactor demonstration in the next year. The MODAR technology uses controlled vibrations on a short, isolated length of pressure tube to locate and reposition tight-fitting annulus spacers. MODAR technology will allow the utilities to demonstrate fuel channel integrity to the regulator and obtain approval for additional years of reactor operation. This paper briefly describes the technology and provides an overview of the tool testing and development. (author)

  16. Implementation of Moderator Circulation Test Temperature Measurement System

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Yeong Muk; Hong, Seok Boong; Kim, Min Seok; Choi, Hwa Rim [KAERI, Daejeon (Korea, Republic of); Kim, Hyung Shin [Chungnam University, Daejeon (Korea, Republic of)


    Moderator Circulation Test(MCT) facility is 1/4 scale facility designed to reproduce the important characteristics of moderator circulation in a CANDU6 calandria under a range of operating conditions. MCT is an equipment with 380 acrylic pipes instead of the heater rods and a preliminary measurement of velocity field using PIV(Particle Image Velocimetry) is performed under the iso-thermal test conditions. The Korea Atomic Energy Research Institute (KAERI) started implementation of MCT Temperature Measurement System (TMS) using multiple infrared sensors. To control multiple infrared sensors, MCT TMS is implemented using National Instruments (NI) LabVIEW programming language. The MCT TMS is implemented to measure sensor data of multiple infrared sensors using the LabVIEW. The 35 sensor pipes of MCT TMS are divided into 2 ports to meet the minimum measurement time of 0.2 seconds. The software of MCT TMS is designed using collection function and processing function. The MCT TMS has the function of monitoring the states of multiple infrared sensors. The GUI screen of MCT TMS is composed of sensor pipe categories for user.

  17. Nuclear Archeology for CANDU Power Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Broadhead, Bryan L [ORNL


    The goal of this work is the development of so-called 'nuclear archeology' techniques to predict the irradiation history of both fuel-related and non-fuel-related materials irradiated in the CANDU (CANada Deuterium Uranium) family of nuclear reactors. In this application to CANDU-type reactors, two different scenarios for the collection of the appropriate data for use in these procedures will be assumed: the first scenario is the removal of the pressure tubes, calandria tubes, or fuel cladding and destructive analysis of the activation products contained in these structural materials; the second scenario is the nondestructive analysis (NDA) of the same hardware items via high-resolution gamma ray scans. There are obvious advantages and disadvantages for each approach; however, the NDA approach is the central focus of this work because of its simplicity and lack of invasiveness. The use of these techniques along with a previously developed inverse capability is expected to allow for the prediction of average flux levels and irradiation time, and the total fluence for samples where the values of selected isotopes can be measured.

  18. Utilization of noise analysis technique for mechanical vibrations estimation in the ATUCHA{sub 1} and Embalse Argentine NPP; Uso de la tecnica de analisis de ruido para la estimacion de vibraciones mecanicas en las centrales nucleares argentinas Atucha I y Embalse

    Energy Technology Data Exchange (ETDEWEB)

    Lescano, V.H.; Wentzeis, L.M. [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Centro Atomico Constituyentes; Guevara, M.; Moreno, C. [Nucleoelectrica Argentina S.A., Cordoba (Argentina). Central Nuclear Embalse; Pineyro, J. [Nucleoelectrica Argentina S.A., Buenos Aires (Argentina). Central Nuclear Atucha I


    In Argentine, comprehensive noise measurements have been performed with the reactor instrumentation of the PHWR power plant Atucha I and Embalse. The Embalse reactor is a CANDU-600 (600 Mwe) type pressurized heavy water reactor. It's a heavy water moderator and heavy water cooled natural uranium fueled pressure tube system. Signal of vanadium and platinum type in core-self power neutron detectors of ex-core ion chambers and of a moderator pressure sensor have been recorded and analysed. The vibration of reactor internals as vertical and horizontal in-core neutron flux detectors units and the coolant channels systems, consisting of calandria and pressure tubes with fuel bundles, have been identified and monitored during normal reactor operation. Atucha I, is a PHWR reactor natural uranium fueled, and heavy water moderated and cooled. Neutron noise techniques using of ex-core ionization chambers and in-core Vanadium SPND's were implemented, among others, in order to produce early detection of anomalous vibrations in the reactor internals. Noise analysis was successfully performed to identify normal and peculiar vibrations in particular reactor internals. (author)

  19. Gadolinium depletion event in a CANDU® moderator - causes and recovery

    Energy Technology Data Exchange (ETDEWEB)

    Evans, D.W.; Price, J.; Swami, D.; Fracalanza, E.; Brett, M.E.; Puzzuoli, F.V.; Garg, A. [Ontario Power Generation, Pickering, Ontario (Canada); Herrmann, O.; Rudolph, A. [Kinectrics Inc., Toronto, Ontario (Canada); Stuart, C.; Glowa, G. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Smee, J. [Niagara Technical Consultants, St. Catharines, Ontario (Canada)


    Gadolinium nitrate is added to the moderator of CANDU units to maintain the reactor in a guaranteed shutdown state (GSS). In April 2008, after being in stable GSS for over 30 hours, one of Ontario Power Generation's Pickering-B units showed a gradual depletion of the dissolved gadolinium, despite purification being isolated. Further additions of gadolinium stabilized the moderator gadolinium concentration, however, since the root cause of the depletion was not immediately identified, the unit was placed in the drained shutdown state, per established procedures. The cumulative gadolinium depletion amounted to about 3200 grams, the equivalent of about 12 ppm. Analysis showed the presence of oxalate in the moderator water. It is well-known that gadolinium forms a very insoluble oxalate (log K{sub sp} = -29.1). Although sub-micron filtration of water samples did not show the presence of gadolinium particulate, the measured levels of oxalate, 1.2 to 2 ppm, were sufficient to react with 1.4 to 2.4 ppm of gadolinium. The source of oxalate was traced to radiolysis of dissolved CO{sub 2} species. This unit had been experiencing chronic low-level ingress of CO{sub 2} from the Annulus Gas System. Free oxalate ion is normally susceptible to radiolytic breakdown back to CO{sub 2}, but Gd{sup 3+} provides a stable sink for radiogenic oxalate, 2 Gd{sup 3+} + 3 C{sub 2}O{sub 4}{sup 2-} → Gd{sub 2}(C{sub 2}O{sub 4}){sub 3}. Subsequent testing confirmed that gadolinium oxalate is quite stable with respect to gamma irradiation. Inspections showed well-crystallized gadolinium oxalate deposited on moderator system surfaces. Estimates indicated that about 1200 grams of gadolinium could have deposited on in-core surfaces, including the outside of the calandria tubes. That amount of negative reactivity was a concern, since it would prevent re-start of the unit. OPG, with support from AECL-Chalk River and Kinectrics, embarked on a two-pronged chemistry recovery program aimed at 1

  20. Experimental investigation of quench and re-wetting temperatures of hot horizontal tubes well above the limiting temperature for solid–liquid contact

    Energy Technology Data Exchange (ETDEWEB)

    Takrouri, Kifah, E-mail: [Department of Engineering Physics, McMaster University, 1280 Main Street West, Hamilton, Ontario L8S 4L7 (Canada); Luxat, John, E-mail: [Department of Engineering Physics, McMaster University, 1280 Main Street West, Hamilton, Ontario L8S 4L7 (Canada); Hamed, Mohamed [Thermal Processing Laboratory (TPL), Department of Mechanical Engineering, McMaster University, 1280 Main Street West, Hamilton, Ontario L8S 4L7 (Canada)


    Highlights: • Quench and re-wetting temperatures were measured upon jet quenching of hot cylindrical tubes. • Correlations have been developed and provided good fit of data. • Quench and re-wetting temperatures were found to greatly depend on water subcooling. • Stagnation point showed higher quench and re-wetting temperatures than other locations. • Quench temperature decreased by increasing surface curvature and tube conductivity. • Re-wetting temperature is a weak function of both variables. - Abstract: Quench cooling of a hot dry surface involves the rapid decrease in surface temperature resulting from bringing the hot surface into sudden contact with a coolant at a lower temperature. Quench temperature is the onset of the rapid decrease in surface temperature and corresponds to the onset of destabilization of a vapor film that exists between the hot surface and the coolant. Situations involving quench cooling are encountered in a number of postulated accidents in Canada Deuterium Uranium CANDU reactors, such as the quench of a hot calandria tube in certain Loss of Coolant Accidents LOCA. If the calandria tube temperature is not reduced by initiation of quench heat transfer, then this may lead to subsequent fuel channel failure and for this accident knowledge of quench heat transfer characteristics is of great importance. In this study, a Water Quench Facility WQF has been designed and built at the Thermal Processing Laboratory TPL at McMaster University and a series of experimental tests were carried out to investigate the quench of hot horizontal tubes using a vertical rectangular water multi-jet system. The tubes were heated to a temperature between 380 and 780 °C then cooled to the jet temperature. The temperature variation with time in tube circumferential and axial directions was measured. The two-phase flow behavior and the propagation of the re-wetting front around and along the tubes were simultaneously observed using a high-speed camera

  1. Prediction of the moderator temperature field in a heavy water reactor based on a cellular neural network

    Directory of Open Access Journals (Sweden)

    S.O. Starkov


    Full Text Available Reactors with heavy water coolants and moderators have been used extensively in today's power industry. Monitoring of the moderator condition plays an important role in ensuring normal operation of a power plant. A cellular neural network, the architecture of which has been adapted for hardware implementation, is proposed for use in a system for prediction of the heavy water moderator temperature. A reactor model composed in accordance with the CANDU Darlington heavy water reactor design was used to form the training sample collection and to control correct operation of the neural network structure. The sample components for the adjustment and configuration of the network topology include key parameters that characterize the energy generation process in the core. The paper considers the feasibility of the temperature prediction only for the calandria's central cross-section. To solve this problem, the cellular neural network architecture has been designed, and major parts of the digital computational element and methods for their implementation based on an FPLD have also been developed. The method is described for organizing an optical coupling between individual neural modules within the network, which enables not only the restructuring of the topology in the training process, but also the assignment of priorities for the propagation of the information signals of neurons depending on the activity in a situation analysis at the neural network structure inlet. Asynchronous activation of cells was used based on an oscillating fractal network, the basis for which was a modified ring oscillator. The efficiency of training the proposed architecture using stochastic diffusion search algorithms is evaluated. A comparative analysis of the model behavior and the results of the neural network operation have shown that the use of the neural network approach is effective in safety systems of power plants.

  2. 2-D CFD time-dependent thermal-hydraulic simulations of CANDU-6 moderator flows

    Energy Technology Data Exchange (ETDEWEB)

    Mehdi Zadeh, Foad [Department of Engineering Physics/Polytechnique Montréal, Montréal, QC (Canada); Étienne, Stéphane [Department of Mechanical Engineering/Polytechnique Montréal, Montréal, QC (Canada); Teyssedou, Alberto, E-mail: [Department of Engineering Physics/Polytechnique Montréal, Montréal, QC (Canada)


    Highlights: • 2-D time-dependent CFD simulations of CANDU-6 moderator flows are presented. • A thermal-hydraulic code using thermal physical fluid properties is used. • The numerical approach and convergence is validated against available data. • Flow configurations are correlated using Richardson’s number. • Frequency components indicate moderator flow oscillations vs. Richardson numbers. - Abstract: The distribution of the fluid temperature and mass density of the moderator flow in CANDU-6 nuclear power reactors may affect the reactivity coefficient. For this reason, any possible moderator flow configuration and consequently the corresponding temperature distributions must be studied. In particular, the variations of the reactivity may result in major safety issues. For instance, excessive temperature excursions in the vicinity of the calandria tubes nearby local flow stagnation zones, may bring about partial boiling. Moreover, steady-state simulations have shown that for operating condition, intense buoyancy forces may be dominant, which can trigger a thermal stratification. Therefore, the numerical study of the time-dependent flow transition to such a condition, is of fundamental safety concern. Within this framework, this paper presents detailed time-dependent numerical simulations of CANDU-6 moderator flow for a wide range of flow conditions. To get a better insight of the thermal-hydraulic phenomena, the simulations were performed by covering long physical-time periods using an open-source code (Code-Saturne V3) developed by Électricité de France. The results show not only a region where the flow is characterized by coherent structures of flow fluctuations but also the existence of two limit cases where fluid oscillations disappear almost completely.

  3. Propiedades estructurales de ejes huecos y sólidos con una grieta plana//Structural proprieties of hollow and whole axles, with a flat crack

    Directory of Open Access Journals (Sweden)

    Jon‐Ariza De‐Miguel


    Full Text Available En este trabajo se estudiaron las diversas ventajas constructivas que los ejes huecos presentan respecto a los sólidos, como la reducción de masa, mejor control de calidad, menores defectos en producción y facilidad de acoplamiento. De ahí su profusión en aplicaciones industriales, como ejes de ferrocarril, de reductores o en calandrias. Dichas ventajas se ven alteradas cuando el eje trabaja a fatiga en flexión rotativa y aparece una grieta sobre el mismo. Se comprobaron dichas ventajas y sus variaciones sobre un modelo inicial de una grieta plana superficial, y se ha observado que los ejes huecos resultan más perjudicados por estas grietas para ciertos valores del diámetro interior, mientras que para otros son ventajosos. Se concluye del presente estudio que puede establecerse un rango de valores para el diámetro interior de ejes huecos en el cual su aprovechamiento seríaóptimo.Palabras claves: ejes huecos y sólidos, propiedades estructurales, grieta plana.______________________________________________________________________________AbstractThis work studies the several advantages hollow axles show when compared to whole ones, such as lightness, better quality control, less manufacturing defects, coupling simplicity. Thence their being profusely employed in industry, in applications like railway axles, gearboxes or calenders. Theseadvantages are altered when the axle works under bending with rotation under fatigue and a crack appears on its surface. An analytical research upon a flat surface crack model has been made of these advantages and their variations and it has been observed that hollow axles suffer more from these cracks than whole ones for certain inner diameter values, whereas for others even less than whole ones. It can be deduced that a gap for the inner diameter can be ascertained, where the axle´s rendering would be optimal.Key words: hollowand whole axles, structural properties, flat crack.

  4. La lumière et le temps sur la scène baroque : Poetique & Pratique

    Directory of Open Access Journals (Sweden)

    Françoise Siguret


    Full Text Available Time: Aristotle, in the Poetics, recommends the playwright to confine his tragedy within «two revolutions of the sun»; the concept refers to the light perception, to the fact that greek drama is acted in the open air. The messengers and the chorus represented on the stage, in the present time, what happened outside of it. In the age of the French classical theatre, the chronological sequence of the action had to conform to the laws of the reason: the so-called rule of the twenty-four hours became an indisputable rule of the action. A time exactly measured, substituting the time of the light, cyclical and mythical. In Italy, pastorals, mythological melodramas and all that belonged to the court entertainments (ballets, operas, tournaments conformed to a cyclical time in which the four seasons constituted the scenery, linking life to the four liturgical seasons and to the four parts of the day, from noon to midnight (cfr. Endymion and the Ballet de la Nuit. Light: Need to light up the indoor playhouse for practical and moral issues. Italian craftsmen implement the technical tools; a certain difference between primary light (intended to light up the stage and the auditorium and the lumi (the supplementary lighting related to a specific performance. Buontalenti’s lighting devices (sun, moon, rainbows, divine and princely splendour will enchant the spectators. France will discover these stagecraft effects with the Calandria (1548, without subsequent developments. Afterwards, Corneille will be fascinated by the ‘baroque’ charm (Médée, 1639 and Andromède, 1650. In the second half of the XVIIth century, while machinery invades opera and tragedy in music, Racine refuses anything intended to deceive the eye, though creates a lighting that may be «listened» (Britannicus. The Allegories (the «other discourse» convey meanings on the baroque stage through the perpetual slow motion of the gods and Time, till the final glory of the Prince: Cosimo

  5. Fundamentos ecológicos y biogeográficos de la rareza de la avifauna madrileña: Una propuesta de modificación del catálogo regional de especies amenazadas

    Directory of Open Access Journals (Sweden)

    Carrascal, L. M.


    ático – ‘Sensible a la Alteración de su Hábitat’; Calandria – descatalogarla; Mirlo Acuático - ‘Sensible a la Alteración de su Hábitat’; Colirrojo Real - ‘Sensible a la Alteración de su Hábitat’ o ‘Vulnerable’; Papamoscas Gris - ‘De Interés Especial’; Alcaudón Real Meridional – descatalogarla; Picogordo – ‘De Interés Especial’.

  6. Modelling flow and work hardening behaviour of cold worked Zr–2.5Nb pressure tube material in the temperature range of 30–600 {sup o}C

    Energy Technology Data Exchange (ETDEWEB)

    Dureja, A.K., E-mail: [Reactor Design and Development Group, Bhabha Atomic Research Centre, Trombay, Mumbai (India); Sinha, S.K. [Reactor Design and Development Group, Bhabha Atomic Research Centre, Trombay, Mumbai (India); Pawaskar, D.N.; Seshu, P. [Department of Mechanical Engineering, Indian Institute of Technology Bombay, Mumbai (India); Chakravartty, J.K. [Mechanical Metallurgy Division, Bhabha Atomic Research Centre, Trombay, Mumbai (India); Sinha, R.K. [Department of Atomic Energy, Anushakti-Bhavan, Near Gateway of India, Mumbai (India)


    Under a postulated accident scenario of loss of cooling medium in an Indian Pressurised Heavy Water Reactor (IPHWR), temperature of the pressure tubes can rise and lead to large deformations. In order to investigate the modes of deformation of pressure tube – calandria tube assembly, material property data defining the flow behaviour over a temperature range from room temperature (RT) to 800 {sup o}C are needed. It is of practical importance to formulate mathematical equations to describe the stress–strain relationships of a material for a variety of reasons, such as the analysis of forming operations and the assessment of component's performance in service. A number of constitutive relations of empirical nature have been proposed and they have been found very suitable to describe the behaviour of a material. Although these relations are of empirical nature, various metallurgical factors appear to decide applicability of each of these relations. For example, grain size influences mainly the friction stress while the strain hardening is governed by dislocation density. In a recent work, tensile deformation behaviour of pressure tube material of IPHWR has been carried out over a range of temperature and strain rates (Dureja et al., 2011). It has been found that the strength parameters (yield and ultimate tensile strength) vary along the length of the tube with higher strength at the trailing end as compared to the leading end. This stems from cooling of the billet during the extrusion process which results in the variation of microstructure, texture and dislocation density from the leading to the trailing end. In addition, the variation in metallurgical parameters is also expected to influence the work hardening behaviour, which is known to control the plastic instability (related to uniform strain). In the present investigation, the tensile flow and work-hardening behaviour of a cold worked Zr–2.5Nb pressure tube material of IPHWRs has been studied over