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Sample records for internal coolant channels

  1. Nuclear reactor coolant channels

    International Nuclear Information System (INIS)

    Macbeth, R.V.

    1978-01-01

    Reference is made to coolant channels for pressurised water and boiling water reactors and the arrangement described aims to improve heat transfer between the fuel rods and the coolant. Baffle means extending axially within the channel are provided and disposed relative to the fuel rods so as to restrict flow oscillations occurring within the coolant from being propagated transversely to the axis of the channel. (UK)

  2. Nuclear reactor coolant channels

    International Nuclear Information System (INIS)

    Macbeth, R.V.

    1978-01-01

    A nuclear reactor coolant channel is described that is suitable for sub-cooled reactors as in pressurised water reactors as well as for bulk boiling, as in boiling water reactors and steam generating nuclear reactors. The arrangement aims to improve heat transfer between the fuel elements and the coolant. Full constructional details are given. See also other similar patents by the author. (U.K.)

  3. Coolant channel module CCM

    International Nuclear Information System (INIS)

    Hoeld, Alois

    2007-01-01

    A complete and detailed description of the theoretical background of an '(1D) thermal-hydraulic drift-flux based mixture-fluid' coolant channel model and its resulting module CCM will be presented. The objective of this module is to simulate as universally as possible the steady state and transient behaviour of the key characteristic parameters of a single- or two-phase fluid flowing within any type of heated or non-heated coolant channel. Due to the possibility that different flow regimes can appear along any channel, such a 'basic (BC)' 1D channel is assumed to be subdivided into a number of corresponding sub-channels (SC-s). Each SC can belong to only two types of flow regime, an SC with just a single-phase fluid, containing exclusively either sub-cooled water or superheated steam, or an SC with a two-phase mixture flow. After an appropriate nodalisation of such a BC (and therefore also its SC-s) a 'modified finite volume method' has been applied for the spatial discretisation of the partial differential equations (PDE-s) which represent the basic conservation equations of thermal-hydraulics. Special attention had to be given to the possibility of variable SC entrance or outlet positions (which describe boiling boundaries or mixture levels) and thus the fact that an SC can even disappear or be created anew. The procedure yields for each SC type (and thus the entire BC), a set of non-linear ordinary 1st order differential equations (ODE-s). To link the resulting mean nodal with the nodal boundary function values, both of which are present in the discretised differential equations, a special quadratic polygon approximation procedure (PAX) had to be constructed. Together with the very thoroughly tested packages for drift-flux, heat transfer and single- and two-phase friction factors this procedure represents the central part of the here presented 'Separate-Region' approach, a theoretical model which provides the basis to the very effective working code package CCM

  4. Internal contamination during enmasse coolant channel replacement (EMCCR) operation at MAPS-2, Kalpakkam

    International Nuclear Information System (INIS)

    Rajaram, S.; Sreedevi, K.R.; Rajendran, V.; Parthasarathi, S.; Kannan, V.; Gurg, R.P.

    2003-01-01

    The operation of Enmasse Coolant Channel Replacement (EMCCR) of MAPS Unit-2, started in the month of March 2002. Since the whole operation was carried out inside the reactor building in the environment of high radiation level in air, it was necessary to whole body count all the personnel involved in the operation. These personnel were whole body monitored for internal contamination due to gamma emitting radionuclides. To measure the internal contamination of the personnel a shadow shield whole body counter having NaI (Tl) detector is used. To study the concentration of airborne radionuclides present during the EMCCR job air filter samples were collected and analysed using HPGe gamma Spectrometer. It was seen that nearly 95% of the total activity present in the air was due to activation products and the remaining 5% is due to fission products. About 2000 personnel (of which 97% contract personnel) involved in the EMCCR job were whole body monitored. Around 57% showed no internal contamination. In the remaining 43% of personnel activation products such as 65 Zn, 95 Zr/ 95 Nb, and 60 Co were observed. It is interesting to note that the radionuclides observed in the internal contamination were also observed in air filter samples collected from the work spot. The observation that all the personnel involved in the EMCCR job received dose less than the Derived Recording Level (DRL) of 0.6 mSv can be attributed to a clean and neat job carried out by the management and personnel involved. (author)

  5. Dynamic Analysis of Coolant Channel and Its Internals of Indian 540 MWe PHWR Reactor

    Directory of Open Access Journals (Sweden)

    A. Rama Rao

    2008-04-01

    Full Text Available The horizontal coolant channel is one of the important parts of primary heat transport system in PHWR type of reactors. There are in all 392 channels in the core of Indian 540 MWe reactor. Each channel houses 13 natural uranium fuel bundles and shielding and sealing plugs one each on either side of the channel. The heavy water coolant flows through the coolant channel and carries the nuclear heat to outside the core for steam generation and power production in the turbo-generator. India has commissioned one 540 MWe PHWR reactor in September 2005 and another similar unit will be going into operation very shortly. For a complete dynamic study of the channel and its internals under the influence of high coolant flow, experimental and modeling studies have been carried out. A good correlation has been achieved between the results of experimental and analytical models. The operating life of a typical coolant channel typically ranges from 10 to 15 full-power years. Towards the end of its operating life, its health monitoring becomes an important activity. Vibration diagnosis plays an important role as a tool for life management of coolant. Through the study of dynamic characteristics of the coolant channel under simulated loading condition, an attempt has been made to develop a diagnostics to monitor the health of the coolant channel over its operating life. A study has been also carried out to characterize the fuel vibration under different flow condition.

  6. Heat transfer in a two-pass internally ribbed turbine blade coolant channel with cylindrical vortex generators

    Energy Technology Data Exchange (ETDEWEB)

    Hibbs, R.; Chen, Y.; Nikitopoulos, D. [Louisiana State Univ., Baton Rouge, LA (United States)] [and others

    1995-10-01

    The effect of vortex generators on the mass (heat) transfer from the ribbed passage of a two pass turbine blade coolant channel is investigated with the intent of optimizing the vortex generator geometry so that significant enhancements in mass/heat transfer can be achieved. In the experimental configuration considered, ribs are mounted on two opposite walls; all four walls along each pass are active and have mass transfer from their surfaces but the ribs are non-participating. Mass transfer measurements, in the form of Sherwood number ratios, are made along the centerline and in selected inter-rib modules. Results are presented for Reynolds number in the range of 5,000 to 40,000, pitch to rib height ratios of 10.5 and 21, and vortex generator-rib spacing to rib height ratios of 0.55, and 1.5. Centerline and spanwise averaged Sherwood number ratios are presented along with contours of the Sherwood number ratios. Results indicate that the vortex generators induce substantial increases in the local mass transfer rates, particularly along the side walls, and modest increases in the average mass transfer rates. The vortex generators have the effect of making the inter-rib profiles along the ribbed walls more uniform. Along the side walls, horse-shoe vortices that characterize the vortex generator wake are associated with significant mass transfer enhancements. The wake effects and the levels of enhancement decrease somewhat with increasing Reynolds number and decreasing pitch.

  7. Freeform Deposition Method for Coolant Channel Closeout

    Science.gov (United States)

    Gradl, Paul R. (Inventor); Reynolds, David Christopher (Inventor); Walker, Bryant H. (Inventor)

    2017-01-01

    A method is provided for fabricating a coolant channel closeout jacket on a structure having coolant channels formed in an outer surface thereof. A line of tangency relative to the outer surface is defined for each point on the outer surface. Linear rows of a metal feedstock are directed towards and deposited on the outer surface of the structure as a beam of weld energy is directed to the metal feedstock so-deposited. A first angle between the metal feedstock so-directed and the line of tangency is maintained in a range of 20-90.degree.. The beam is directed towards a portion of the linear rows such that less than 30% of the cross-sectional area of the beam impinges on a currently-deposited one of the linear rows. A second angle between the beam and the line of tangency is maintained in a range of 5-65 degrees.

  8. CFD analyses of coolant channel flowfields

    Science.gov (United States)

    Yagley, Jennifer A.; Feng, Jinzhang; Merkle, Charles L.

    1993-01-01

    The flowfield characteristics in rocket engine coolant channels are analyzed by means of a numerical model. The channels are characterized by large length to diameter ratios, high Reynolds numbers, and asymmetrical heating. At representative flow conditions, the channel length is approximately twice the hydraulic entrance length so that fully developed conditions would be reached for a constant property fluid. For the supercritical hydrogen that is used as the coolant, the strong property variations create significant secondary flows in the cross-plane which have a major influence on the flow and the resulting heat transfer. Comparison of constant and variable property solutions show substantial differences. In addition, the property variations prevent fully developed flow. The density variation accelerates the fluid in the channels increasing the pressure drop without an accompanying increase in heat flux. Analyses of the inlet configuration suggest that side entry from a manifold can affect the development of the velocity profile because of vortices generated as the flow enters the channel. Current work is focused on studying the effects of channel bifurcation on the flow field and the heat transfer characteristics.

  9. Power module assemblies with staggered coolant channels

    Science.gov (United States)

    Herron, Nicholas Hayden; Mann, Brooks S; Korich, Mark D

    2013-07-16

    A manifold is provided for supporting a power module assembly with a plurality of power modules. The manifold includes a first manifold section. The first face of the first manifold section is configured to receive the first power module, and the second face of the first manifold section defines a first cavity with a first baseplate thermally coupled to the first power module. The first face of the second manifold section is configured to receive the second power module, and the second face of the second manifold section defines a second cavity with a second baseplate thermally coupled to the second power module. The second face of the first manifold section and the second face of the second manifold section are coupled together such that the first cavity and the second cavity form a coolant channel. The first cavity is at least partially staggered with respect to second cavity.

  10. Channel type reactors with supercritical water coolant. Russian experience

    International Nuclear Information System (INIS)

    Kuznetsov, Y.N.; Gabaraev, B.A.

    2003-01-01

    Transition to coolant of supercritical parameters allows for principle engineering-andeconomic characteristics of light-water nuclear power reactors to be substantially enhanced. Russian experience in development of channel-type reactors with supercritical water coolant has demonstrated advantages and practical feasibility of such reactors. (author)

  11. Vision system for precision alignment of coolant channels

    International Nuclear Information System (INIS)

    Kar, S.; Rao, Y.V.; Valli Kumar; Joshi, D.G.; Chadda, V.K.; Nigam, R.K.; Kayal, J.N.; Panwar, S.; Sinha, R.K.

    1997-01-01

    This paper describes a vision system which has been developed for precision alignment of Coolant Channel Replacement Machine (CCRM) with respect to the front face of the coolant channel under repair/replacement. It has provisions for automatic as well as semi-automatic alignment. A special lighting scheme has been developed for providing illumination to the front face of the channel opening. This facilitates automatic segmentation of the digitized image. The segmented image is analysed to obtain the centre of the front face of the channel opening and thus the extent of misalignment i.e. offset of the camera with respect to the front face of the channel opening. The offset information is then communicated to the PLC to generate an output signal to drive the DC servo motors for precise positioning of the co-ordinate table. 2 refs., 5 figs

  12. Analysis of radiation exposure during creep adjustment to the coolant channels at Madras Atomic Power Station

    International Nuclear Information System (INIS)

    Varadhan, R.S.; Venkataramana, K.; Kannan, R.K.; Sreekumaran Nair, B.; Chudalayandi, K.

    1994-01-01

    In pressurised heavy water reactors the coolant channels made of zircaloy-2 undergo creep deformation used intense neutron irradiation in the reactor core. In order to measure and provide for the changes in the dimensions, base line data of internal diameters, sag and length of the 306 coolant channels are measured as pre service inspection (PSI) before the reactor is loaded with fuel prior to criticality. Subsequently as part of in service inspection (ISI), axial creep of every channel is measured in every annual shutdown of the reactor and creep adjustment is done on those channels where creep expansion margin for the next one year operation is low. A study was carried out to assess the radiological impact of the job at Madras Atomic Power Station (MAPS). Various measures adopted for reducing the individual and collective doses on the job are discussed in this report. (author). 3 refs., 2 tabs

  13. Rapid thermal transient in a reactor coolant channel

    International Nuclear Information System (INIS)

    Cherubini, A.

    1986-01-01

    This report deals with the problem of one-dimensional thermo-fluid-dynamics in a reactor coolant channel, with the aim of determining the evolution in time of the coolant (H*L2O), in one-and/or two-phase regimes, subjected to a great and rapid increase in heat flux (accident conditions). To this aim, the following are set out: a) the physical model used; b) the equations inherent in the above model; c) the numerical methods employed to solve them by means of a computer programme called CABO (CAnale BOllente). Next a typical problem of rapid thermal transient resolved by CABO is reported. The results obtained, expressed in form of graphs, are fully discussed. Finally comments on possible developments of CABO follow

  14. Consideration of hot channel factors in design for providing operating margins on coolant channel outlet temperature

    International Nuclear Information System (INIS)

    Sharma, V.K.; Surendar, C.; Bapat, C.N.

    1994-01-01

    The Indian Pressurized Heavy Water Reactors (IPHWR) are horizontal pressure tube reactors using natural uranium oxide fuel in the form of short (495 mm) clusters. The fuel clusters in the Zr-Nb pressure tubes are cooled by high pressure, high temperature and subcooled circulating heavy water. Coolant flow distribution to individual channels is designed to match the power distribution so as to obtain uniform coolant outlet temperature. However, during operation, the coolant outlet temperature in individual channels deviate from their nominal value due to: tolerances in process design; effects of grid frequency on the pump speed; deviation in channel powers from the nominal values due to on-power fuelling and movement of reactivity devices, and so on. Thus an operating margin, between the highest permissible and nominal coolant outlet temperatures, is required taking into account various hot channel factors that contribute to higher coolant outlet temperatures. The paper discusses the methodology adopted to assess various hot channel factors which would provide optimum operating margins while ensuring sub-cooling. (author)

  15. Void fraction calculation in a channel containing boiling coolant

    International Nuclear Information System (INIS)

    Norelli, F.

    1978-01-01

    The problem of void fraction calculation was studied for a channel containing boiling coolant, when a slip ratio correlation is used. Use of fitting (e.g. polinomial or rational algebraic) for slip ratio correlation and the characteristic method are proposed in this work. In this way we are reduced to some elementary quadrature problem. Another problem discussed in the present work concerns what we must consider as ''initial condition'' in any initial value problem, in order to take into account different error distributions in steady state and in successive time-dependent calculations

  16. On line monitoring of temperatures of coolant channels by thermal imaging in a laboratory set-up fabricated for the detection of leakage of coolants

    Energy Technology Data Exchange (ETDEWEB)

    Mukherjee, S; Ghosh, J K [Bhabha Atomic Research Centre, Bombay (India). Radiometallurgy Div.; Patel, R J [Bhabha Atomic Research Centre, Mumbai (India). Refuelling Technology Division

    1994-12-31

    Leakage from coolant channels in Pressurised Heavy Water Reactors (PHWR) increases the temperatures of the faulty channels. Measurement of temperatures of the coolant channels is, therefore, one way to detect the leaking channel. Thermal imaging technique offers a unique means for this detection providing a fast, non-contact, on-line measurement. An experiment was carried out for the detection of leakage of coolants through the seal plugs of the coolant channels in PHWR using an experimental setup under the simulated conditions of temperature and pressure of the coolant channels inside the reactor and using an infrared imaging system. The experimental details and the observations have been presented. 7 figs.

  17. On line monitoring of temperatures of coolant channels by thermal imaging in a laboratory set-up fabricated for the detection of leakage of coolants

    International Nuclear Information System (INIS)

    Mukherjee, S.; Ghosh, J.K.; Patel, R.J.

    1994-01-01

    Leakage from coolant channels in Pressurised Heavy Water Reactors (PHWR) increases the temperatures of the faulty channels. Measurement of temperatures of the coolant channels is, therefore, one way to detect the leaking channel. Thermal imaging technique offers a unique means for this detection providing a fast, non-contact, on-line measurement. An experiment was carried out for the detection of leakage of coolants through the seal plugs of the coolant channels in PHWR using an experimental setup under the simulated conditions of temperature and pressure of the coolant channels inside the reactor and using an infrared imaging system. The experimental details and the observations have been presented. 7 figs

  18. Estimation of the mechanical behavior of irradiated coolant channels at a nuclear plant for its decomissing

    International Nuclear Information System (INIS)

    Piquin, Ruben; Zanni, Pablo

    2003-01-01

    The widespread replacement of reactor internals generates a substantial volume of active material.It is essential to work with these components at least in a partial way before the next planned stop.Due to the fact that the reactor internals pool and the storage pools for irradiated nuclear fuel have limited capacities, it has been proposed to compact an experimental shift of 50 irradiated coolant channels, that are currently placed in storage pools.Basically the processed waste will be put in baskets at the bottom of the storage pools.The alternative choice proposes to divide an irradiation coolant channel tube into different parts: stainless steel section, zircaloy-4 section and stainless steel section with hardened zones with cobalt alloys named Estelite-6.Having planned the constructive and operative solutions, the mechanical characterization of the different parts of the channel tube remains to be done.In the present paper, the necessary compacted strength of the irradiation coolant channel tube will be estimated for the stainless steel section and for the zircaloy-4 section, starting from experiment with unirradiated material and considering effects of radiation damage and hydrides on the ductility.These results will be used to design the necessary compacted tools for the semi-industrial installation

  19. Upper internals of PWR with coolant flow separator

    International Nuclear Information System (INIS)

    Chevereau, G.; Heuze, A.

    1989-01-01

    The upper internals for a PWR has a collecting volume for the coolant merging from the core and an apparatus for separating the flow of coolant. This apparatus has a guide for the control rods, a lower plate perforated to allow the coolant through from the core, an upper plate also perforated to allow the coolant through to the collecting volume and a peripheral binding ring joining the two plates. Each guide comprises an envelope without holes and joined perceptibly tight to the plates [fr

  20. Fatigue cycles evaluation of 500 MWe PHWR coolant channel sealdisc

    International Nuclear Information System (INIS)

    Chawla, D.S.; Vaze, K.K.; Kushwaha, H.S.; Gupta, K.S.; Bhambra, H.S.

    1998-07-01

    At each end of coolant channel there is one sealing plug assembly. The sealdisc is a part of sealing plug assembly. The sealdisc is used to avoid leakage of heavy water. The importance of sealdisc can be understood by the fact that there are 784 sealdiscs in one 500 MWe PHWR unit. During the life time of reactor the sealdisc will be subjected to cyclic loads due to reactor startup, shutdown, power setback and also due to refuelling operations. Excessive reversal of stresses may lead to fatigue failure. The sealdisc failure may cause loss of coolant accidents. Since sealdisc is safety class 1 component, it has to be qualified according to ASME Section III Division 1 NB. For cyclic loads, the fatigue analysis is essential to assess the allowable number of cycles and also to check the total usage factor due to different cyclic loads. To evaluate the allowable fatigue cycles, the analysis is carried out using finite element method. The present report deals with the fatigue cycles evaluation of 500 MWe PHWR sealdisc. The finite element model having eight noded axisymmetric elements is used for the analysis. The various loads considered in the analysis are mechanical loads arising due to refuelling operations and number of temperature-pressure transients. During refuelling, the sealdisc is removed and reinstalled back by use of fuelling machine ram which applies load at centre as well as at rocker point of sealdisc. The stress analysis is carried out for each stage of loading during refuelling and fatigue cycles are evaluated. For temperature transient, decoupled thermal analysis is carried out. At various instants of time, the stresses are computed using temperatures calculated in thermal analysis. The pressure variation is also considered along with temperature variation. The fatigue cycles are evaluated for each transient using maximum alternating stress intensities. The usage factors are calculated for various temperature/pressure transients and refuelling loads

  1. Knock-limited performance of several internal coolants

    Science.gov (United States)

    Bellman, Donald R; Evvard, John C

    1945-01-01

    The effect of internal cooling on the knock-limited performance of an-f-28 fuel was investigated in a CFR engine, and the following internal coolants were used: (1) water, (2), methyl alcohol-water mixture, (3) ammonia-methyl alcohol-water mixture, (4) monomethylamine-water mixture, (5) dimethylamine-water mixture, and (6) trimethylamine-water mixture. Tests were run at inlet-air temperatures of 150 degrees and 250 degrees F. to indicate the temperature sensitivity of the internal-coolant solutions.

  2. Ageing of coolant channels in nuclear reactors (PHWRs)

    International Nuclear Information System (INIS)

    Mitra, T.L.; Chowdhury, M.K.; Gupta, R.K.; Pandarinathan, P.R.; Seth, V.K.

    1994-01-01

    In PHWRs, ageing of various components takes place due to factors like fast neutron flux, temperature, stress, environment etc. In coolant channel, the most severely affected component due to ageing is pressure tube, though other components like end fitting, calandria tube, garter spring spacer also show ageing to a limited extent. Ageing effects in pressure tube are seen in the form of diametral and axial creep, corrosion, delayed hydrogen cracking and irradiation hardening. In calandria tube and garter spring spacer, creep and hardening are seen though these are not of concern in PHWRs. In end fitting, irradiation embrittlement and abrasion of sealing faces are the areas of concern. Ageing process in these components are the areas of concern. Ageing process in these components are effectively retarded by taking measures like selection of proper material, manufacturing process, control of environmental chemistry, and design modifications. Experience and information gained in various Indian and foreign reactors have been used to improve upon the design in 220 MWe reactors and have formed the basis of design for 500 MWe reactors. (author). 3 refs., 5 figs

  3. Theoretical studying the stability of steady-state regime of a channel with a coolant condensation

    International Nuclear Information System (INIS)

    Savikhin, O.G.

    1987-01-01

    Based on the boiling channel stability theory, the channel steady-state stability with the coolant condensation is studied. Condensable coolants are used in the NPP steam-separator superheaters as well as in cryogenic technique. Under certain conditions the coolant flow rate and temperature fluctuations may be excited in the parallel channel system with coolant condensation, which produce a sufficient effect on the heat exchange equipment operation reliability. To describe unsteady processes of heat and mass transfer in the channel, a homogeneous two-phase flow one dimensional model is used. The results obtained allow one to make a conclusion concerning the effect of some parameters on condensing channel steady-state regime stability: reduction of inlet and outlet unheated communication length, pressure drop increase at the outlet plate and its reduction at the inlet one lead to the increase of stability margin

  4. Methodologies and technologies for life assessment and management of coolant channels of Indian pressurised heavy water reactors

    International Nuclear Information System (INIS)

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

    2002-01-01

    coolant channels and various research publications in international journals formed the bases for evolution of the above safety methodologies. Today, the analytical models together with the safety evaluation methodologies have become an important tool for assessing fitness for service of individual pressure tube in different reactors. Apart from developing the above analytical tools and methodologies for safety evaluation, several systems and tools have been designed and developed to cater for the activities like remote inspection, online integrity monitoring and life extension etc., of our life management programme of coolant channel. Amongst the systems developed for remote inspection of the coolant channel such as Dry Visual Inspection System (DRYVIS), BARC Inspection System(BARCIS) and Hydraulic Remote Inside Diameter Measurement System (HYRIM) etc., BARCIS is more versatile in terms of its capability to carry out different type of measurements. Inspection of the coolant channel by BARCIS has been incorporated as an important activity in our life management programme of the coolant channel. A non destructive diagnostic technique based on the principle of vibration measurement has been developed to identify the contact between the pressure tube and calandria tube. This technique called 'Non Intrusive Vibration Diagnostic Technique (NIVDT)' is being used as a tool to screen the vulnerable pressure tubes so as reduce the reduce the inspection load. Tools and techniques developed for online integrity monitoring includes Sliver Sample Scraping Tool (SSST). At present, scraping operation can be carried out in both the dry and the wet channel of operating reactor. The development of SSST is intended for taking nondestructively sliver samples from the operating coolant channel. These sliver samples are then analysed to get hydrogen content in them. Measured hydrogen content helps not only in predicting the operating life of a coolant channel but also in improving the

  5. Design and development of remotely operated coolant channel cutting machine

    International Nuclear Information System (INIS)

    Suthar, R.L.; Sinha, A.K.; Srikrishnamurty, G.

    1994-01-01

    One of the coolant tubes of Narora Atomic Power Station (NAPS) reactor needs to be removed. To remove a coolant tube, four cutting operations, (liner tube cutting, end-fitting cutting, machining of seal weld of bellow ring and finally coolant tube cutting) are required to be carried out. A remotely operated cutting machine to carry out all these operations has been designed and developed by Central Workshops. This machine is able to cut at the exact location because of numerically controlled axial and radial travel of tool. Only by changing the tool head and tool holder, same machine can be used for various types of cutting/machining operations. This report details the design, manufacture, assembly and testing work done on the machine. (author). 4 figs

  6. In-Service Inspection system for coolant channels of Indian PHWRS - evolution and experience

    International Nuclear Information System (INIS)

    Puri, R.K.; Singh, M.

    2006-01-01

    In-Service Inspection (ISI) is the most important of all periodic monitoring and surveillance activities for assuring the structural integrity of coolant channels in the life extension and management of pressurized heavy water reactors (PHWR-CANDU). Indian PHWRs (220 MWe) are characterized by consists by 306 coolant channels in each unit. These channels have to be inspected for various parameters over the operating life of the reactor. ISI of coolant channels necessitated the indigenous development of an inspection system called BARCIS (BARC Channel Inspection System) at Bhabha Atomic Research Center. BARCIS consists of mainly three parts; drive and control unit, special sealing plug and an inspection head carrying various NDT sensors. Five such systems have been built and deployed at various power plants. The paper deals with the development of the BARCIS system for meeting the ISI requirements of coolant channels, development cycle of this system from its conception to evolution to the present state, challenges, data generated and experience gained (ISI of nearly 900 coolant channels has been completed). Prior to BARCIS, pressure tube gauging equipment for pre-service inspection of coolant tubes was developed in 1980. Moreover a tool for ISI of coolant channels in dry condition was developed in 1990. The paper also describes evolution of various contingency procedures and devices developed over the last one decade. Future plans taking into account technological advancement, changes in the scope of inspection due to design and operating experiences and plant layout will also be covered. The paper describes the efforts put in to develop drive and control mechanism to suit the different vault layouts. The drive mechanism is responsible for linear and rotary movement of the inspection head to carry out 100% volumetric inspection. Special emphasis has been laid on the safety devices required during the inspection activity. Special measures for heavy water retention in

  7. Management of large scale coolant channel replacement programme for Indian PHWRs

    International Nuclear Information System (INIS)

    Bhatnagar, V.K.; Chadda, S.K.; Arya, R.C.

    1994-01-01

    Coolant channel assemblies form most important core components of pressurised heavy water reactors. Zirconium alloy pressure tube which form part of coolant channel assemblies are subjected to environment of high neutron flux, high pressure and temperature. Under those operating environmental conditions, the pressure tubes material undergoes degradation of metallurgical and mechanical properties in addition to dimensional changes. The coolant channels are subjected to an in-service inspection (ISI) programme for monitoring the health particularly of the pressure tubes. The en-mass replacement of pressure tubes is needed after most of the pressure tubes show unacceptable conditions for an assured safe and reliable operation. An overview of various issues pertaining to this aspect is presented. (author). 4 figs

  8. Effect of ribbed and smooth coolant cross-flow channel on film cooling

    International Nuclear Information System (INIS)

    Peng, Wei; Sun, Xiaokai; Jiang, Peixue; Wang, Jie

    2017-01-01

    Highlights: • Little different for plenum model and the cross-flow model at M = 0.5. • Crossflow model is much better than plenum model at M = 1.0, especially with ribs. • Coolant flow channel with V-shaped ribs has the best adiabatic film cooling. • Film cooling with the plenum model is better at M = 0.5 than at M = 1.0. • Crossflow model is better at M = 0.5 near film hole and at M = 1.0 for downstream. - Abstract: The influence of ribbed and unribbed coolant cross-flow channel on film cooling was investigated with the coolant supply being either a plenum-coolant feed or a coolant cross-flow feed. Validation experiments were conducted with comparison to numerical results using different RANS turbulence models showed that the RNG k–ε turbulence model and the RSM model gave closer predictions to the experimental data than the other RANS models. The results indicate that at a low blowing ratio of M = 0.5, the coolant supply channel structure has little effect on the film cooling. However, at a high blowing ratio of M = 1.0, the adiabatic wall film cooling effectiveness is significantly lower with the plenum feed than with the cross-flow feed, especially for the cases with ribs. The film cooling with the plenum model is better at M = 0.5 than at M = 1.0. The film cooling with the cross-flow model is better at a blowing ratio of M = 0.5 in the near hole region, while further downstream, it is better at M = 1.0. The results also show that the coolant cross-flow channel with V-shaped ribs has the best adiabatic film cooling effectiveness.

  9. Effect of ribbed and smooth coolant cross-flow channel on film cooling

    Energy Technology Data Exchange (ETDEWEB)

    Peng, Wei; Sun, Xiaokai [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Tsinghua University, Beijing 100084 (China); Jiang, Peixue, E-mail: jiangpx@tsinghua.edu.cn [Key Laboratory for Thermal Science and Power Engineering of Ministry of Educations, Department of Thermal Engineering, Tsinghua University, Beijing 100084 (China); Wang, Jie [Institute of Nuclear and New Energy Technology, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Tsinghua University, Beijing 100084 (China)

    2017-05-15

    Highlights: • Little different for plenum model and the cross-flow model at M = 0.5. • Crossflow model is much better than plenum model at M = 1.0, especially with ribs. • Coolant flow channel with V-shaped ribs has the best adiabatic film cooling. • Film cooling with the plenum model is better at M = 0.5 than at M = 1.0. • Crossflow model is better at M = 0.5 near film hole and at M = 1.0 for downstream. - Abstract: The influence of ribbed and unribbed coolant cross-flow channel on film cooling was investigated with the coolant supply being either a plenum-coolant feed or a coolant cross-flow feed. Validation experiments were conducted with comparison to numerical results using different RANS turbulence models showed that the RNG k–ε turbulence model and the RSM model gave closer predictions to the experimental data than the other RANS models. The results indicate that at a low blowing ratio of M = 0.5, the coolant supply channel structure has little effect on the film cooling. However, at a high blowing ratio of M = 1.0, the adiabatic wall film cooling effectiveness is significantly lower with the plenum feed than with the cross-flow feed, especially for the cases with ribs. The film cooling with the plenum model is better at M = 0.5 than at M = 1.0. The film cooling with the cross-flow model is better at a blowing ratio of M = 0.5 in the near hole region, while further downstream, it is better at M = 1.0. The results also show that the coolant cross-flow channel with V-shaped ribs has the best adiabatic film cooling effectiveness.

  10. Identification of flow patterns by neutron noise analysis during actual coolant boiling in thin rectangular channels

    International Nuclear Information System (INIS)

    Kozma, R.; van Dam, H.; Hoogenboom, J.E.

    1992-01-01

    The primary objective of this paper is to introduce results of coolant boiling experiments in a simulated materials test reactor-type fuel assembly with plate fuel in an actual reactor environment. The experiments have been performed in the Hoger Onderwijs Reactor (HOR) research reactor at the Interfaculty Reactor Institute, Delft, The Netherlands. In the analysis, noise signals of self-powered neutron detectors located in the neighborhood of the boiling region and thermocouple in the channel wall and in the coolant are used. Flow patterns in the boiling coolant have been identified by means of analysis of probability density functions and power spectral densities of neutron noise. It is shown that boiling has an oscillating character due to partial channel blockage caused by steam slugs generated periodically between the plates. The observed phenomenon can serve as a basis for a boiling detection method in reactors with plate-type fuels

  11. On Line Neutron Flux Mapping in Fuel Coolant Channels of a Research Reactor

    International Nuclear Information System (INIS)

    Barbot, Loic; Domergue, Christophe; Villard, Jean-Francois; Destouches, Christophe; Braoudakis, George; Wassink, David; Sinclair, Bradley; Osborn, John-C.; Wu, Huayou; Blandin, C.; Thevenin, Mathieu; Corre, Gwenole; Normand, Stephane

    2013-06-01

    This work deals with the on-line neutron flux mapping of the OPAL research reactor. A specific irradiation device has been set up to investigate fuel coolant channels using subminiature fission chambers to get thermal neutron flux profiles. Experimental results are compared to first neutronic calculations and show good agreement (C/E ∼0.97). (authors)

  12. Analysis of molten fuel behavior in coolant channel during severe accidents in KALIMER

    International Nuclear Information System (INIS)

    Suk, Soo Dong; Lee, Yong Bum; Hahn, Do Hee

    2004-11-01

    Preliminary safety analyses of the KALIMER-600 design have shown that the design has inherent safety characteristics and is capable of accommodating double fault initiators such as ATWS events without boiling coolant or melting fuel. For the future design of liquid metal reactor, however, the evaluation of the safety performance and the determination of containment requirements may require consideration of tripe-fault accident sequences of extremely low probability of occurrence that leads to fuel melting. For any postulated accident sequence which leads to core melting, in-vessel retention of the core debris will required as a design requirement for the future design of LMR. For sodium-cooled core designs with metallic fuel, one of the major phenomenological modeling uncertainties to be resolved is the potential for freezing and plugging of molten metallic fuel in above- and below-core structures and possibly in inter-subassembly spaces. In this study, scoping analyses were carried out to evaluate the penetration depths in the coolant channels by molten fuel mixture during the unprotected loss-of-flow accidents in the core of the KALIMER-600. It is assumed in the analyses that a solid fuel crust would start to form upon contact with the coolant channel structure temperature of which is below the fuel solidus. The analysis results predict that the coolant channels would be plugged by the freezing molten fuel in the inlet lower shield as well as in the outlet, fission-gas-plenum region for the KALIMER-600 design

  13. Stresses imposed by coolant channel end shield interaction in 200 MWe PHWR

    International Nuclear Information System (INIS)

    Mehra, V.K.; Singh, R.K.; Soni, R.S.; Kushwaha, H.S.; Kakodkar, A.

    1983-01-01

    End shield of 200 MWe Pressurised Heavy Water Reactor (PHWR) is a composite tube sheet structure consisting of two circular tube sheets joined together by lattice tubes. Each lattice tube houses a coolant channel assembly which is connected to the end shield through shock absorber device. End shield assembly is suspended in the vault by hanger rods and its horizontal position is controlled by a set of pre-compressed springs. Coolant channel assemblies elongate due to their exposure to fast neutron flux in the reactor. This permanent elongation is monitored periodically. When growth of the channel exceeds a present value, it is prevented from further elongation by the shock absorbing device. Resultant force exerted on the end shield makes it move. This paper describes a numerical method used for evaluating these forces and movement of the end shield. Stresses produced by these forces are calculated by using finite element method. Typical stress values are verified by strain gauge measurements. (orig.)

  14. A thermal analysis computer programme package for the estimation of KANUPP coolant channel flows and outlet header temperature distribution

    International Nuclear Information System (INIS)

    Siddiqui, M.S.

    1992-06-01

    COFTAN is a computer code for actual estimation of flows and temperatures in the coolant channels of a pressure tube heavy water reactor. The code is being used for Candu type reactor with coolant flowing 208 channels. The simulation model first performs the detailed calculation of flux and power distribution based on two groups diffusion theory treatment on a three dimensional mesh and then channel powers, resulting from the summation of eleven bundle powers in each of the 208 channels, are employed to make actual estimation of coolant flows using channel powers and channel outlet temperature monitored by digital computers. The code by using the design flows in individual channels and applying a correction factor based on control room monitored flows in eight selected channels, can also provide a reserve computational tool of estimating individual channel outlet temperatures, thus providing an alternate arrangements for checking Rads performance. 42 figs. (Orig./A.B.)

  15. Bandwidth of reactor internals vibration resonance with coolant pressure oscillations

    International Nuclear Information System (INIS)

    Proskuryakov, K.N.; Novikov, K.S.; Galivec, E.Yu.

    2009-01-01

    In a few decades a significant increase in a part of an electricity development on the NPP will require NPP to be operated in non full capacity modes and increase in operation time in transitive modes. Operating in such conditions as compared to the operation on a constant mode will lead to the increase in cyclic dynamical loading. In water cooled water moderated reactors these loading are realized as low-cyclic and high-cyclic loadings. High-cyclic loadings increases are caused by a raised vibration in non stationary modes of operation. It is known, that in some modes of a non full capacity reactor high-cyclic dynamic loadings can increase. It is obvious, that the development of management technologies is necessary for the life time management operation. In the context of this problem one of the main tasks are revealing and the prevention of the conditions of the occurrence of the operation leading to the resonant interaction of the coolant fluctuations and the equipment, reactor vessel (RV), fuel assemblies (FA) and reactor internals (RI) vibration. To prevent the appearance of the conditions for resonance interaction between the fluid flow and the equipments, it is necessary to provide the different frequencies for the self oscillations in the separated elements of the circulating system and also in the parts of the system formed by the comprising of these elements. While solving these problems it is necessary to have a theoretical and settlement substantiation of an oscillation frequency band of coolant outside of which there is no resonant interaction. The presented work is devoted to finding the solution of this problem. There are results of theoretical an estimation of width of such band as well as the examples of a preliminary quantitative estimation of Q - factors of coolant acoustic oscillatory circuit formed by the equipment of the NPP. The accordance of results had been calculated with had been measured are satisfied for practical purposes. These

  16. The module CCM for the simulation of the thermal-hydraulic situation within a coolant channel

    International Nuclear Information System (INIS)

    Hoeld, A.

    2000-01-01

    A coolant channel module (Cc) will be presented which aim is to simulate, in a very general way, the thermal-hydraulic behaviour of single- and two-phase fluids flowing along a heated (or cooled) vertical, inclined or horizontal coolant channel. It is based on a theoretical drift-flux supported 3-equation mixture-fluid model describing the steady state and transient behaviour of characteristic thermal-hydraulic parameters of a single- and two-phase flow within such a channel. The module can be applied as an element within an overall theoretical model for large and complex plant assemblies (PWR and BWR core channels, parallel channels in 3D cores, primary and secondary sides of different steam generators types etc.). The model refers to a general (basic) coolant channel (BC) which can consists of different flow regimes. The BC has thus to be subdivided accordingly into a number of subchannels (SC-s). All of them can belong, however, to only two types of SC-s (single-phase fluid with subcooled water or superheated steam or a two-phase flow regime). For both of them the possibility of variable entrance or outlet positions has to be considered. For discretization purposes the BC (and thus also the SC-s) have to be subdivided into a number of (BC and SC) nodes, discretizing thus the conservation equations for mass, energy and momentum along these nodes by applying a very general spatial procedure, namely a 'modified finite volume method'. A special quadratic polygon approximation method (PAX procedure) helps then to establish a connection between nodal boundary and mean nodal parameters. Considering their constitutive equations (among them an adequate drift-flux correlation package) yields finally a set of non-linear algebraic and non-linear ordinary differential equations for the characteristic parameters of each of these SC nodes (mass flow, pressure drop, coolant temperature and/or void fraction). Based on this theory a code package (CCM) could be established

  17. ISS Internal Active Thermal Control System (IATCS) Coolant Remediation Project

    Science.gov (United States)

    Morrison, Russell H.; Holt, Mike

    2005-01-01

    The IATCS coolant has experienced a number of anomalies in the time since the US Lab was first activated on Flight 5A in February 2001. These have included: 1) a decrease in coolant pH, 2) increases in inorganic carbon, 3) a reduction in phosphate buffer concentration, 4) an increase in dissolved nickel and precipitation of nickel salts, and 5) increases in microbial concentration. These anomalies represent some risk to the system, have been implicated in some hardware failures and are suspect in others. The ISS program has conducted extensive investigations of the causes and effects of these anomalies and has developed a comprehensive program to remediate the coolant chemistry of the on-orbit system as well as provide a robust and compatible coolant solution for the hardware yet to be delivered. The remediation steps include changes in the coolant chemistry specification, development of a suite of new antimicrobial additives, and development of devices for the removal of nickel and phosphate ions from the coolant. This paper presents an overview of the anomalies, their known and suspected system effects, their causes, and the actions being taken to remediate the coolant.

  18. Mechanical and Radiological Characterization of Different parts of an Irradiation Coolant Channel Tube from Atucha I Nuclear Plant

    International Nuclear Information System (INIS)

    Piquin, Ruben

    2001-01-01

    The widespread replacement of reactor internals has generated a substantial volume of active material. It is essential to work with these components at least in a partial way before the next planned stop, which will take place during the second semester of the year 2002. Due to the fact that the reactor internals pool and the storage pool for irradiated nuclear fuel have limited capacities, it has been proposed to compact an experimental shift of 50 irradiated coolant channels, that are currently placed in storage pools. Basically the processed waste will be put in baskets at the bottom pools.The alternative choice proposes to divide an irradiation coolant channel tube into different parts: stainless steel section, zircaloy-4 section and stainless steel section with hardened zones with cobalt alloys named Estelite-6. The person in charge has already planned the constructive and operative solutions but the mechanical characterization of the different parts of the channel tube is necessary in order to dimension the compaction tool needed for the semi-industrial installation.In the present special report, two well-differentiated actions will be described. The necessary compacted strength of the irradiation coolant channel tube will be estimated for the stainless steel section and the zircaloy-4 section starting from experiment with unirradiated material and considering effects of radiation damage and hydrides on the ductility.These results will be used to design the necessary compacted tools for the semi-industrial installation. The necessary equipment for the radiological characterization of the different material sections already specified will be described and the most important emitting particles of radiation that could be detected will be mentioned. Also the decontamination process to use including the radiological characterization of every stage of the process will be described in order to establish the decontamination factor. Finally the most important

  19. Thermal-hydraulic analysis and design improvement for coolant channel of ITER shield block

    International Nuclear Information System (INIS)

    Zhao Ling; Li Huaqi; Zheng Jiantao; Yi Jingwei; Kang Weishan; Chen Jiming

    2013-01-01

    As an important part for ITER, shield block is used to shield the neutron heat. The structure design of shield block, especially the inner coolant channel design will influence its cooling effect and safety significantly. In this study, the thermal-hydraulic analysis for shield block has been performed by the computational fluid dynamics software, some optimization suggestions have been proposed and thermal-hydraulic characteristics of the improved model has been analyzed again. The analysis results for improved model show that pressure drop through flow path near the inlet and outlet region of the shield block has been reduced, and the total pressure drop in cooling path has been reduced too; the uniformity of the mass flowrate distribution and the velocity distribution have been improved in main cooling branches; the local highest temperature of solid domain reduced considerably, which could avoid thermal stress becoming too large because of coolant effect unevenly. (authors)

  20. Selection of an Alternate Biocide for the ISS Internal Thermal Control System Coolant, Phase 2

    Science.gov (United States)

    Wilson, Mark E.; Cole, Harold; Weir, Natalee; Oehler, Bill; Steele, John; Varsik, Jerry; Lukens, Clark

    2004-01-01

    The ISS (International Space Station) ITCS (Internal Thermal Control System) includes two internal coolant loops that utilize an aqueous based coolant for heat transfer. A silver salt biocide had previously been utilized as an additive in the coolant formulation to control the growth and proliferation of microorganisms within the coolant loops. Ground-based and in-flight testing demonstrated that the silver salt was rapidly depleted, and did not act as an effective long-term biocide. Efforts to select an optimal alternate biocide for the ITCS coolant application have been underway and are now in the final stages. An extensive evaluation of biocides was conducted to down-select to several candidates for test trials and was reported on previously. Criteria for that down-select included: the need for safe, non-intrusive implementation and operation in a functioning system; the ability to control existing planktonic and biofilm residing microorganisms; a negligible impact on system-wetted materials of construction; and a negligible reactivity with existing coolant additives. Candidate testing to provide data for the selection of an optimal alternate biocide is now in the final stages. That testing has included rapid biocide effectiveness screening using Biolog MT2 plates to determine minimum inhibitory concentration (amount that will inhibit visible growth of microorganisms), time kill studies to determine the exposure time required to completely eliminate organism growth, materials compatibility exposure evaluations, coolant compatibility studies, and bench-top simulated coolant testing. This paper reports the current status of the effort to select an alternate biocide for the ISS ITCS coolant. The results of various test results to select the optimal candidate are presented.

  1. Development of in-situ laser cutting technique for removal of single selected coolant channel from pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Vishwakarma, S.C.; Upadhyaya, B.N.

    2016-01-01

    We report on the development of a pulsed Nd:YAG laser based cutting technique for removal of single coolant channel from pressurized heavy water reactor (PHWR). It includes development of special tools/manipulators and optimization of laser cutting process parameters for cutting of liner tube, end fitting, bellow lip weld joint, and pressure tube stubs. For each cutting operation, a special tool with precision motion control is utilized. These manipulators/tools hold and move the laser cutting nozzle in the required manner and are fixed on the same coolant channel, which has to be removed. This laser cutting technique has been successfully deployed for removal of selected coolant channels Q-16, Q-15 and N-6 of KAPS-2 reactor with minimum radiation dose consumption and in short time. (author)

  2. Using the coolant temperature noise for measuring the flow rate in the RBMK technological channels

    International Nuclear Information System (INIS)

    Selivanov, V.M.; Karlov, N.P.; Martynov, A.D.; Prostyakov, V.V.; Lysikov, B.V.; Kuznetsov, B.A.; Pallagi, D.; Khorani, Sh.; Khargitai, T.; Tezher, Sh.

    1983-01-01

    The problems are considered connected with the possibility of using thermometric correlation method to measure the coolant flow rate in the RBMK reactor technological channels. The main attention is paid to the study of the physical nature of the coolant temperature pulsations and to estimation of the effect of parameters of the primary thermaelectrical converter (TEC) on the results of measurements. In the process of reactor inspections made using the thermometric correlation flowmeter of a special design, the temperature noise distribution in the points of flow rate measurement is studied, the noise intensity and physical nature are determined, as well as the effect of different TEC parameters (TEC inertia and base distance between them) on the measurement accuracy. On the basis of the analysis of the effect on the results of the TEC thermal inertia measured value divergence, tausub(α) and transport time, tau sub(T), a conclusion is made on the necessity of choosing the base distance between TEC with tausub(T)>tausub(d)

  3. Numerical Simulation of Non-Rotating and Rotating Coolant Channel Flow Fields. Part 1

    Science.gov (United States)

    Rigby, David L.

    2000-01-01

    Future generations of ultra high bypass-ratio jet engines will require far higher pressure ratios and operating temperatures than those of current engines. For the foreseeable future, engine materials will not be able to withstand the high temperatures without some form of cooling. In particular the turbine blades, which are under high thermal as well as mechanical loads, must be cooled. Cooling of turbine blades is achieved by bleeding air from the compressor stage of the engine through complicated internal passages in the turbine blades (internal cooling, including jet-impingement cooling) and by bleeding small amounts of air into the boundary layer of the external flow through small discrete holes on the surface of the blade (film cooling and transpiration cooling). The cooling must be done using a minimum amount of air or any increases in efficiency gained through higher operating temperature will be lost due to added load on the compressor stage. Turbine cooling schemes have traditionally been based on extensive empirical data bases, quasi-one-dimensional computational fluid dynamics (CFD) analysis, and trial and error. With improved capabilities of CFD, these traditional methods can be augmented by full three-dimensional simulations of the coolant flow to predict in detail the heat transfer and metal temperatures. Several aspects of turbine coolant flows make such application of CFD difficult, thus a highly effective CFD methodology must be used. First, high resolution of the flow field is required to attain the needed accuracy for heat transfer predictions, making highly efficient flow solvers essential for such computations. Second, the geometries of the flow passages are complicated but must be modeled accurately in order to capture all important details of the flow. This makes grid generation and grid quality important issues. Finally, since coolant flows are turbulent and separated the effects of turbulence must be modeled with a low Reynolds number

  4. Eddy current monitoring of spacers in coolant channel assemblies of nuclear reactor

    International Nuclear Information System (INIS)

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

    1993-01-01

    An eddy current testing method has been standardised for monitoring spacer springs which are used in coolant channel assemblies of pressurised heavy water nuclear reactors (PHWRs). The standard bobbin coil probe used for monitoring the spacer spring detects only the location but does not monitor the tilt orientation and tilt angle of a tilted spacer spring. The knowledge of location along with the tilt orientation of the spacer spring greatly improves the performance of repositioning methods. A modified probe with angular windings has been developed in laboratory tests for monitoring the location as well as the tilt orientation of the spacer springs. Experimental results are presented showing excellent performance of the modified probe in monitoring the exact location as well as tilt orientation of a spacer spring. The modified probe has also been used successfully in the field during repositioning of spacer springs in PHWRs before commissioning. (Author)

  5. The relationship between dynamic and average flow rates of the coolant in the channels of complex shape

    Science.gov (United States)

    Fedoseev, V. N.; Pisarevsky, M. I.; Balberkina, Y. N.

    2018-01-01

    This paper presents interconnection of dynamic and average flow rates of the coolant in a channel of complex geometry that is a basis for a generalization model of experimental data on heat transfer in various porous structures. Formulas for calculation of heat transfer of fuel rods in transversal fluid flow are acquired with the use of the abovementioned model. It is shown that the model describes a marginal case of separated flows in twisting channels where coolant constantly changes its flow direction and mixes in the communicating channels with large intensity. Dynamic speed is suggested to be identified by power for pumping. The coefficient of proportionality in general case depends on the geometry of the channel and the Reynolds number (Re). A calculation formula of the coefficient of proportionality for the narrow line rod packages is provided. The paper presents a comparison of experimental data and calculated values, which shows usability of the suggested models and calculation formulas.

  6. The Analysis of the Effect of Coolant Channel Width on Fuel Loading of the RSG-GAS Core

    International Nuclear Information System (INIS)

    Surbakti; Tukiran

    2004-01-01

    The RGS-GAS using uranium silicide fuel, plate type and 250 g U of loading is planned to increase the fuel loading to 300 g U even to 400 g U. The silicide fuel has advantages when increase the fuel loading in the same volume. Because of that case, it is necessary to analyze the effect of coolant channel width on fuel loading of the RSG-GAS core. Analyzing the effect the work which done is to generate cell and core calculation using WIMSD/4 and Batan-2DIFF codes. The WIMSD/4 code is used to generate cross section of core material and Batan-2DIFF is used to calculate the effective multiplication factor. The model that used in this calculation there are three kind of fuel loading namely, 250 g U, 300 g U and 400 g U. The coolant channel width is simulated from 1.75 mm to 2.55 mm. From that fuel loadings, it is analyzed which coolant channel width gave the best effective multiplication factor. From result of analysis showed that the best effective multiplication factor is on the coolant channel width of 2.55 mm for third of fuel loadings. (author)

  7. Development of innovative tools based on fuelling machine for ageing management of coolant channels of 220 MWe PHWRs

    International Nuclear Information System (INIS)

    Dev, Mahender; Roy, Shyamal; Bhattachrya, Sambit; Singh, Jit Pal; Patel, R.J.; Agarwal, R.G.

    2006-01-01

    PHWR coolant channels are required to be inspected periodically to satisfy the regulatory requirement, to provide information about known or suspected problem and to provide information to assist in future design. This paper describes these tools and techniques, their capabilities and experience of implementing these in reactor site

  8. Improvement of lifetime availability through design, inspection, repair and replacement of coolant channels of Indian Pressurized Heavy Water Reactors

    International Nuclear Information System (INIS)

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

    1998-01-01

    This paper covers an overview of the work carried out for the life management of the coolant channels of Indian Pressurised Heavy Water Reactors. In order to improve maintainability of the coolant channels and reduce down time needed for periodical creep adjustment, improved design of channel hardware were developed. The modular insulation panel, designed as a substitute for the jig saw panels, reduces the time needed for accessing the space around the end-fitting significantly. A compact mechanical snubber has been developed to totally eliminate the need for periodic creep adjustment. In addition, the paper also describes the technologies developed for performing some special inspection, repair and replacement tasks for the coolant channels. These include systems for garter spring repositioning by Mechanical Flexing Technique for fresh reactors and Integrated Garter Spring Repositioning System for operating reactors. A tooling system, developed for in-situ retrieval of sliver scrape samples from pressure tubes, is also described. These samples can be analysed in laboratories to yield valuable information on hydrogen concentration in pressure tube material. The current and planned activities towards development of technologies for improvement of the life time availability of the power plants are addressed. (author)

  9. Design of channel experiment equipment for measuring coolant velocity of innovative research reactor

    International Nuclear Information System (INIS)

    Muhammad Subekti; Endiah Puji Hastuti; Dedi Heriyanto

    2014-01-01

    The design of innovative high flux research reactor (RRI) requires high power so that the capability core cooling requires to be improved by designing the faster core coolant velocity near to the critical velocity limit. Hence, the critical coolant velocity as the one of the important parameter of the reactor safety shall be measured by special equipment to the velocity limit that may induce fuel element degradation. The research aims is to calculate theoretically the critical coolant velocity and to design the special experiment equipment namely EXNal for measuring the critical coolant velocity in fuel element subchannel of the RRI. EXNal design considers the critical velocity calculation result of 20.52 m/s to determine the variation of flow rate of 4.5-29.2 m 3 /h, in which the experiment could simulate the 1-4X standard coolant velocity of RSG-GAS as well as destructive test of RRI's fuel plate. (author)

  10. Fine element (F.E.) modelling of hydrogen migration and blister formation in PHWR coolant channels

    International Nuclear Information System (INIS)

    Prasad, P.S.; Dutta, B.K.; Sinha, R.K.; Kushwaha, H.S.; Mahajan, S.C.; Kakodkar, A.

    1995-01-01

    The formation of a cold spot in pressure tube due to its contact with calandria tube of PHWR coolant results in the migration of Hydrogen in pressure tube towards contact zone from its surrounding material. A 3-D finite element code SPARSH is developed to model the hydrogen redistribution and consequent hydride blister formation due to thermal and Hydrogen concentration gradients. In the present paper, the details and performance of this code are presented. (author). 6 refs., 2 figs

  11. Thermal-Hydraulic Analysis of Coolant Flow Decrease in Fuel Channels of Smolensk-3 RBMK during GDH Blockage Event

    International Nuclear Information System (INIS)

    Costa, A. L.; Cherubini, M.; D'Auria, F.; Giannotti, W.; Moskalev, A.

    2007-01-01

    One of the transients that have received considerable attention in the safety evaluation of RBMK reactors is the partial break of a group distribution header (GDH). The coolant flow rate blockage in one GDH might lead to excessive heat-up of the pressure tubes and can result in multiple fuel channels (FC) ruptures. In this work, the GDH flow blockage transient has been studied considering the Smolensk-3 RBMK NPP (nuclear power plant). In the RBMK, each GDH distributes coolant to 40-43 FC. To investigate the behavior of each FC belonging to the damaged GDH and to have a more realistic trend, one (affected) GDH has been schematised with its forty-two FC, one by one. The calculations were performed using the 0-D NK (neutron kinetic) model of the RELAP5-3.3 stand-alone code. The results show that, during the event, the mass flow rate is disturbed differently according to the power distribution established for each FC in the schematization. The start time of the oscillations in mass flow rate depends strongly on the attributed power to each FC. It was also observed that, during the event, the fuel channels at higher thermal power values tend to undergo first cladding rupture leaving the reactor to scram and safeguarding all the other FCs connected to the affected GDH.

  12. Thermal fluid dynamic behavior of coolant helium gas in a typical reactor VHTGR channel of prismatic core

    International Nuclear Information System (INIS)

    Belo, Allan Cavalcante

    2016-01-01

    The current studies about the thermal fluid dynamic behavior of the VHTGR core reactors of 4 th generation are commonly developed in 3-D analysis in CFD (computational fluid dynamics), which often requires considerable time and complex mathematical calculations for carrying out these analysis. The purpose of this project is to achieve thermal fluid dynamic analysis of flow of gas helium refrigerant in a typical channel of VHTGR prismatic core reactor evaluating magnitudes of interest such as temperature, pressure and fluid velocity and temperature distribution in the wall of the coolant channel from the development of a computer code in MATLAB considering the flow on one-dimensional channel, thereby significantly reducing the processing time of calculations. The model uses three different references to the physical properties of helium: expressions given by the KTA (German committee of nuclear safety standards), the computational tool REFPROP and a set of constant values for the entire channel. With the use of these three references it is possible to simulate the flow treating the gas both compressible and incompressible. The results showed very close values for the interest quantities and revealed that there are no significant differences in the use of different references used in the project. Another important conclusion to be observed is the independence of helium in the gas compressibility effects on thermal fluid dynamic behavior. The study also indicated that the gas undergoes no severe effects due to high temperature variations in the channel, since this goes in the channel at 914 K and exits at approximately 1263 K, which shows the excellent use of helium as a refrigerant fluid in reactor channels VHTGR. The comparison of results obtained in this work with others in the literature served to confirm the effectiveness of the one-dimensional consideration of method of gas flow in the coolant channel to replace the models made in 3-D for the pressure range and

  13. Study on parameters of self-oscillations of the coolant flow rate in an evaporating channel of a boiling-type reactor

    International Nuclear Information System (INIS)

    Proshutinskij, A.P.; Lobachev, A.G.

    1979-01-01

    The experimental data on the oscillation frequencies and amplitudes of the coolant flow rate at the limit of the thermohydraulic stability of the boiling type reactor evaporating channel are presented. The experiments have been carried out on the channel simulators of three modifications -smooth-tube, with intensifiers of a transverse crimp type and of an inner spiral ribbing type. The range of the investigated regime parameters is as follows: the pressure - 2.5-14MPa; the heat flux density is 0.015-0.8MV/m 2 , mass velocity is 252-2520 kg/(m 2 xs), the temperature at the channel entrance is from 50 deg C up to (tsub(s) -5)deg C. The experimental data analysis is carried out on the assumption that the period of parameter oscillations in the steam generating channel equals the time of the coolant transfer through the channel. The formular is obtained which provides 25% accuracy of the oscillation frequency calculation in the range of underheating parameter variation B=0.5-3.0. As a result the following conclusions have been made: the oscillation frequency of the coolant flow rate is connected with the time of its transfer through the channel and does not practically depend on the type of the heat exchange intensifiers and the degree of the flux throttling at the channel entrance; the self-oscillation amplitude of the coolant flow rate depends on the regime and structural parameters as well

  14. Numerical evaluation of various gas and coolant channel designs for high performance liquid-cooled proton exchange membrane fuel cell stacks

    International Nuclear Information System (INIS)

    Sasmito, Agus P.; Kurnia, Jundika C.; Mujumdar, Arun S.

    2012-01-01

    A careful design of gas and coolant channel is essential to ensure high performance and durability of proton exchange membrane (PEM) fuel cell stack. The channel design should allow for good thermal, water and gas management whilst keeping low pressure drop. This study evaluates numerically the performance of various gas and coolant channel designs simultaneously, e.g. parallel, serpentine, oblique-fins, coiled, parallel-serpentine and a novel hybrid parallel-serpentine-oblique-fins designs. The stack performance and local distributions of key parameters are investigated with regards to the thermal, water and gas management. The results indicate that the novel hybrid channel design yields the best performance as it constitutes to a lower pumping power and good thermal, water and gas management as compared to conventional channels. Advantages and limitation of the designs are discussed in the light of present numerical results. Finally, potential application and further improvement of the design are highlighted. -- Highlights: ► We evaluate various gas and coolant channel designs in liquid-cooled PEM fuel cell stack. ► The model considers coupled electrochemistry, channel design and cooling effect simultaneously. ► We propose a novel hybrid channel design. ► The novel hybrid channel design yields the best thermal, water and gas management which is beneficial for long term durability. ► The novel hybrid channel design exhibits the best performance.

  15. Rotation Effect on Jet Impingement Heat Transfer in Smooth Rectangular Channels with Film Coolant Extraction

    Directory of Open Access Journals (Sweden)

    James A. Parsons

    2001-01-01

    Full Text Available The effect of channel rotation on jet impingement cooling by arrays of circular jets in twin channels was studied. Impinging jet flows were in the direction of rotation in one channel and opposite to the direction of rotation in the other channel. The jets impinged normally on the smooth, heated target wall in each channel. The spent air exited the channels through extraction holes in each target wall, which eliminates cross flow on other jets. Jet rotation numbers and jet Reynolds numbers varied from 0.0 to 0.0028 and 5000 to 10,000, respectively. For the target walls with jet flow in the direction of rotation (or opposite to the direction of rotation, as rotation number increases heat transfer decreases up to 25% (or 15% as compared to corresponding results for non-rotating conditions. This is due to the changes in flow distribution and rotation induced Coriolis and centrifugal forces.

  16. Spatial dependence of the void coefficient in the interstitial coolant channel positions of a stainless steel-clad TRIGA Mark I core

    International Nuclear Information System (INIS)

    Spriggs, Gregory D.; Nelson, George W.; Doane, Harry J.

    1982-01-01

    A new top grid plate was manufactured and installed in the U of A TRIGA. The new grid plate was identical to the old grid plate with respect to the fuel element array, but included two minor modifications; 1) 3/8'' holes were drilled in six interstitial positions between fuel element rings to allow for insertion of a small diameter void rod for void coefficient measurements in the coolant channels, and 2) flux wire holes were drilled in all remaining interstitial positions. The purpose of this report is to update the previously reported void coefficient measurements with data taken in one of the coolant channel positions

  17. Turbulent heat transfer in a coolant channel of a pressurized water reactor (PWR) core

    International Nuclear Information System (INIS)

    Kumar, Sanjeev; Saha, Arun K.; Munshi, Prabhat

    2016-01-01

    Exact predictions in nuclear reactors are more crucial, because of the safety aspects. It necessitates the appropriate modeling of heat transfer phenomena in the reactors core. A two-dimensional thermal-hydraulics model is used to study the detailed analysis of the coolant region of a fuel pin. Governing equations are solved using Marker and Cell (MAC) method. Standard wall functions k-ε turbulence model is incorporated to consider the turbulent behaviour of the flow field. Validation of the code and a few results for a typical PWR running at normal operating conditions reported earlier. There were some discrepancies in the old calculations. These discrepancies have been resolved and updated results are presented in this work. 2D thermal-hydraulics model results have been compared with the 1D thermal-hydraulics model results and conclusions have been drawn. (author)

  18. LOCO - a linearised model for analysing the onset of coolant oscillations and frequency response of boiling channels

    International Nuclear Information System (INIS)

    Romberg, T.M.

    1982-12-01

    Industrial plant such as heat exchangers and nuclear and conventional boilers are prone to coolant flow oscillations which may not be detected. In this report, a hydrodynamic model is formulated in which the one-dimensional, non-linear, partial differential equations for the conservation of mass, energy and momentum are perturbed with respect to time, linearised, and Laplace-transformed into the s-domain for frequency response analysis. A computer program has been developed to integrate numerically the resulting non-linear ordinary differential equations by finite difference methods. A sample problem demonstrates how the computer code is used to analyse the frequency response and flow stability characteristics of a heated channel

  19. Simplified CFD model of coolant channels typical of a plate-type fuel element: an exhaustive verification of the simulations

    Energy Technology Data Exchange (ETDEWEB)

    Mantecón, Javier González; Mattar Neto, Miguel, E-mail: javier.mantecon@ipen.br, E-mail: mmattar@ipen.br [Instituto de Pesquisas Energéticas e Nucleares (IPEN/CNEN-SP), São Paulo, SP (Brazil)

    2017-07-01

    The use of parallel plate-type fuel assemblies is common in nuclear research reactors. One of the main problems of this fuel element configuration is the hydraulic instability of the plates caused by the high flow velocities. The current work is focused on the hydrodynamic characterization of coolant channels typical of a flat-plate fuel element, using a numerical model developed with the commercial code ANSYS CFX. Numerical results are compared to accurate analytical solutions, considering two turbulence models and three different fluid meshes. For this study, the results demonstrated that the most suitable turbulence model is the k-ε model. The discretization error is estimated using the Grid Convergence Index method. Despite its simplicity, this model generates precise flow predictions. (author)

  20. Thermohydraulic behaviour of the hot channel in a PWR type reactor under loss-of-coolant accident conditions (LOCA)

    International Nuclear Information System (INIS)

    Costa, J.R.

    1978-12-01

    An analysis is done of the core behavior for a 1861 MW(th) pressurized water reactor with two coolant loops, during the blowdown phase of a double-ended cold leg rupture, between the main feedwater pump, and the pressure vessel. The analysis is done through a detailed thermohydraulic study of the hot pin channel with RELAP4/MOD 5 code, including the Evaluatin Model options. The problem is solved separately for two values of discharge coefficient (C sub(D)= 1,0 and 0,4). The results show that the maximum clad temperature is lower than the limit value for licensing purposes. Concerning clad material oxidation, the maximum value obtained is also under the limit of acceptance. (author) [pt

  1. Simplified CFD model of coolant channels typical of a plate-type fuel element: an exhaustive verification of the simulations

    International Nuclear Information System (INIS)

    Mantecón, Javier González; Mattar Neto, Miguel

    2017-01-01

    The use of parallel plate-type fuel assemblies is common in nuclear research reactors. One of the main problems of this fuel element configuration is the hydraulic instability of the plates caused by the high flow velocities. The current work is focused on the hydrodynamic characterization of coolant channels typical of a flat-plate fuel element, using a numerical model developed with the commercial code ANSYS CFX. Numerical results are compared to accurate analytical solutions, considering two turbulence models and three different fluid meshes. For this study, the results demonstrated that the most suitable turbulence model is the k-ε model. The discretization error is estimated using the Grid Convergence Index method. Despite its simplicity, this model generates precise flow predictions. (author)

  2. 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

  3. Effects of Channel Geometry and Coolant Fluid on Thermoelectric Net Power

    DEFF Research Database (Denmark)

    Rezaniakolaei, Alireza; Rosendahl, Lasse; Sørensen, Kim

    2014-01-01

    Channel geometry has a strong influence on the heat transfer coefficient and cooling energy input in a heat sink. The net power output in a thermoelectric generator (TEG) can be defined as power generation minus the required cooling energy in TEG. This study aims to evaluate the net power generat......, and the maximum net power output occurs at smaller Reynolds number when the channel hydraulic diameter reduces....... generation in TEG for different size of hydraulic diameter of plate-fin heat sink and over a wide range of Reynolds number. The particular focus of this study is to find optimal Reynolds number in each considered channel hydraulic diameter and to explore optimal channel hydraulic diameter for maximum TEG net...

  4. A contribution to a theory of two-phase flow with phase change and addition of heat in a coolant channel of a LWR-fuel element during a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Gaballah, I.

    1978-09-01

    A contribution to a theory of two-phase flow with phase change and addition of heat in a coolant channel of a LWR-fuel element during a loss-of-coolant accident. A theory was developed for the calculation of a dispersed two phase flow with heat addition in a channel with general area change. The theory was used to study different thermodynamic and gasdynamic processes, which may occur during the emergency cooling after a LOCA of a pressurized water reactor. The basic equations were formulated and solved numerically. The heat transfer mechanism was examined. Calculations have indicated that the radiative heat flux component is small compared to the convective component. A drop size spectrum was used in the calculations. Its effect on the heat transfer was investigated. It was found that the calculation with a mean drop diameter gives good results. Significant thermal non-equilibrium has been evaluated. The effect of different operating parameters on the degree of thermal non-equilibrium was studied. The flow and heat transfer in a channel with cross-sectional area change were calculated. It was shown that the channel deformation affects the state properties and the heat transfer along the channel very strongly. (orig.) 891 GL [de

  5. Liquid metal MHD and heat transfer in a tokamak blanket slotted coolant channel

    International Nuclear Information System (INIS)

    Reed, C.B.; Hua, T.Q.; Black, D.B.; Kirillov, I.R.; Sidorenkov, S.I.; Shapiro, A.M.; Evtushenko, I.A.

    1993-01-01

    A liquid metal MHD (Magnetohydrodynamic)/heat transfer test was conducted at the ALEX (Argonne Liquid Metal Experiment) facility of ANL (Argonne National Laboratory), jointly between ANL and NIIEFA (Efremov Institute). The test section was a rectangular slotted channel geometry (meaning the channel has a high aspect ratio, in this case 10:1, and the long side is parallel to the applied magnetic field). Isothermal and heat transfer data were collected. A heat flux of ∼9 W/cm 2 was applied to the top horizontal surface (the long side) of the test section. Hartmann Numbers to 1050 (2 Tesla), interaction parameters to 9 x 10 3 , Peclet numbers of 10--200, based on the half-width of the small dimension (7mm), and velocities of 1--75 cm/sec. were achieved. The working fluid was NaK (sodium potassium eutectic). All four interior walls were bare, 300-series stainless steel, conducting walls

  6. Experiments on simulation of coolant mixing in fuel assembly head and core exit channel of WWER-440 reactor

    International Nuclear Information System (INIS)

    Kobzar, L.L; Oleksyuk, D.A.

    2006-01-01

    RRC 'Kurchatov Institute' has performed coolant mixing investigation in a head of a full-size simulator of WWER-440 fuel assembly. The experiments were focused on obtaining the data important for investigating the trends in temperature difference between the value registered by a ICIS thermocouple and the value of average temperature. The completed experiments ensure representative of configuration simulation by reproducing every construction peculiar feature of flow part of fuel assembly in the domain between the lower spacing grid and thermocouple location, and also by slightly modified fuel assembly regular elements (or analogues thereof). For the purpose of effectiveness of coolant mixing assessment within the head cross section of FA simulator, we measured coolant temperature distribution both in the place where coolant flow leaves the rod bundle simulator (in 39 data points along the cross section) and in the cross section location of regular ICIS thermocouple simulator (30 data points). The testing was conducted with pressure of (90 - 95) bar, mass coolant flow rates up to 2000 kg/(m 2 .s), temperature of coolant heating in 'hot' parts of the bundle up to 35.. and differences between coolant temperature extremes measured in rod bundle simulator outlet up to 20... Temperature fields were registered in 63 conditions that differ in coolant flow and inlet coolant temperature, electrical heating rate of FA simulator, and radial coolant distribution. In certain registered conditions we simulated coolant leakage to the space between the fuel assemblies. The received test data may be important both for investigation of dependencies between the coolant temperature in regular thermocouple location or average outlet temperature in assembly head, and for validation of CFD codes or subchannel codes (Authors)

  7. Manufacturing and characterization of porous SiC for flow channel inserts in dual-coolant blanket designs

    International Nuclear Information System (INIS)

    Bereciartu, Ainhoa; Ordas, Nerea; Garcia-Rosales, Carmen; Morono, Alejandro; Malo, Marta; Hodgson, Eric R.; Abella, Jordi; Sedano, Luis

    2011-01-01

    SiC is the primary candidate for the flow channel inserts in dual-coolant blanket concepts. Porous SiC ceramics are attractive candidates for this non-structural application, since they can satisfy the required properties through a low cost manufacturing route, compared to SiC f /SiC. This work shows first results of the manufacturing of porous SiC ceramics prepared with different amounts of Y 2 O 3 and Al 2 O 3 as sintering additives. C powders were used as pore-formers by their burnout during oxidation after sintering. Comparison of microstructure, porosity, flexural strength, thermal and electrical conductivity and corrosion under Pb-15.7Li of porous SiC without and with sintering additives is presented. The addition of 2.5 wt.% of Y 2 O 3 and Al 2 O 3 improves the mechanical properties, and reduces the thermal and electrical conductivity down to reasonable values. Preliminary corrosion tests under Pb-15.7 Li at 500 deg. C show that the absence of a dense coating on porous SiC leads to poor corrosion behavior.

  8. Manufacturing and characterization of porous SiC for flow channel inserts in dual-coolant blanket designs

    Energy Technology Data Exchange (ETDEWEB)

    Bereciartu, Ainhoa [CEIT and Tecnun (University of Navarra), Manuel de Lardizabal 15, 20018 San Sebastian (Spain); Ordas, Nerea, E-mail: nordas@ceit.es [CEIT and Tecnun (University of Navarra), Manuel de Lardizabal 15, 20018 San Sebastian (Spain); Garcia-Rosales, Carmen [CEIT and Tecnun (University of Navarra), Manuel de Lardizabal 15, 20018 San Sebastian (Spain); Morono, Alejandro; Malo, Marta; Hodgson, Eric R. [CIEMAT, Avenida Complutense 22, 28040 Madrid (Spain); Abella, Jordi [Institut Quimic de Sarria, University Ramon Llull, Via Augusta 390, 08017 Barcelona (Spain); Sedano, Luis [CIEMAT, Avenida Complutense 22, 28040 Madrid (Spain)

    2011-10-15

    SiC is the primary candidate for the flow channel inserts in dual-coolant blanket concepts. Porous SiC ceramics are attractive candidates for this non-structural application, since they can satisfy the required properties through a low cost manufacturing route, compared to SiC{sub f}/SiC. This work shows first results of the manufacturing of porous SiC ceramics prepared with different amounts of Y{sub 2}O{sub 3} and Al{sub 2}O{sub 3} as sintering additives. C powders were used as pore-formers by their burnout during oxidation after sintering. Comparison of microstructure, porosity, flexural strength, thermal and electrical conductivity and corrosion under Pb-15.7Li of porous SiC without and with sintering additives is presented. The addition of 2.5 wt.% of Y{sub 2}O{sub 3} and Al{sub 2}O{sub 3} improves the mechanical properties, and reduces the thermal and electrical conductivity down to reasonable values. Preliminary corrosion tests under Pb-15.7 Li at 500 deg. C show that the absence of a dense coating on porous SiC leads to poor corrosion behavior.

  9. ISS Internal Active Thermal Control System (IATCS) Coolant Remediation Project -2006 Update

    Science.gov (United States)

    Morrison, Russell H.; Holt, Mike

    2006-01-01

    The IATCS coolant has experienced a number of anomalies in the time since the US Lab was first activated on Flight 5A in February 2001. These have included: 1) a decrease in coolant pH, 2) increases in inorganic carbon, 3) a reduction in phosphate concentration, 4) an increase in dissolved nickel and precipitation of nickel salts, and 5) increases in microbial concentration. These anomalies represent some risk to the system, have been implicated in some hardware failures and are suspect in others. The ISS program has conducted extensive investigations of the causes and effects of these anomalies and has developed a comprehensive program to remediate the coolant chemistry of the on-orbit system as well as provide a robust and compatible coolant solution for the hardware yet to be delivered. This paper presents a status of the coolant stability over the past year as well as results from destructive analyses of hardware removed from the on-orbit system and the current approach to coolant remediation.

  10. Effect of emergency core cooling system flow reduction on channel temperature during recirculation phase of large break loss-of-coolant accident at Wolsong unit 1

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Seon Oh; Cho, Yong Jin [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Kim, Sung Joong [Dept. of Nuclear Engineering, Hanyang University, Seoul (Korea, Republic of)

    2017-08-15

    The feasibility of cooling in a pressurized heavy water reactor after a large break loss-of-coolant accident has been analyzed using Multidimensional Analysis of Reactor Safety-KINS Standard code during the recirculation phase. Through evaluation of sensitivity of the fuel channel temperature to various effective recirculation flow areas, it is determined that proper cooling of the fuel channels in the broken loop is feasible if the effective flow area remains above approximately 70% of the nominal flow area. When the flow area is reduced by more than approximately 25% of the nominal value, however, incipience of boiling is expected, after which the thermal integrity of the fuel channel can be threatened. In addition, if a dramatic reduction of the recirculation flow occurs, excursions and frequent fluctuations of temperature in the fuel channels are likely to be unavoidable, and thus damage to the fuel channels would be anticipated. To resolve this, emergency coolant supply through the newly installed external injection path can be used as one alternative means of cooling, enabling fuel channel integrity to be maintained and permanently preventing severe accident conditions. Thus, the external injection flow required to guarantee fuel channel coolability has been estimated.

  11. Effect of emergency core cooling system flow reduction on channel temperature during recirculation phase of large break loss-of-coolant accident at Wolsong unit 1

    Directory of Open Access Journals (Sweden)

    Seon Oh Yu

    2017-08-01

    Full Text Available The feasibility of cooling in a pressurized heavy water reactor after a large break loss-of-coolant accident has been analyzed using Multidimensional Analysis of Reactor Safety-KINS Standard code during the recirculation phase. Through evaluation of sensitivity of the fuel channel temperature to various effective recirculation flow areas, it is determined that proper cooling of the fuel channels in the broken loop is feasible if the effective flow area remains above approximately 70% of the nominal flow area. When the flow area is reduced by more than approximately 25% of the nominal value, however, incipience of boiling is expected, after which the thermal integrity of the fuel channel can be threatened. In addition, if a dramatic reduction of the recirculation flow occurs, excursions and frequent fluctuations of temperature in the fuel channels are likely to be unavoidable, and thus damage to the fuel channels would be anticipated. To resolve this, emergency coolant supply through the newly installed external injection path can be used as one alternative means of cooling, enabling fuel channel integrity to be maintained and permanently preventing severe accident conditions. Thus, the external injection flow required to guarantee fuel channel coolability has been estimated.

  12. Effects of spacers on blockage of coolant channels in clad melting accidents

    Energy Technology Data Exchange (ETDEWEB)

    Eggen, D. T.; Scale, T.; Hsieh, S. [Northwestern Univ., Evanston, IL (United States). The Technological Inst.

    1977-07-01

    The elements and configuration of these assemblies are representative of the current design for a GCFR. The fuel elements are stainless-steel clad, mixed-oxide spaced by a grid structure on 250 mm centers with a pitch of 9.5 mm, diameter, 7.2 mm, and cladding thickness, 0.5 m. Three series of experiments have been conducted to study the flow and disposition of molten cladding metal into a lower powered blanket region of the reactor following a loss of flow situation. The first two series used a simulant fuel-element bundle to simplify the experimental procedure and make visual observation possible. The 'fuel' was simulated by mullite rods 6.4 mm in diameter and 610 mm long. These were clad with a 50 Pb/50 Sn alloy tubing which was drawn onto the 'fuel'. The first series used cast spacers with webs of about 0.5-0.55 mm thickness placed 175 and 425 mm from the top end of the assembly. The second series used grid spacers fabricated of 0.25 mm alloy strips. This provided a more accurate representation of the hydraulic diameter. The bundle was encased in a hexagonal glass tube. The bundle was at 22/sup 0/C and the molten alloy was poured at a temperature of 260/sup 0/C (35/sup 0/C superheat). Motion pictures recorded the experiments and the bundle was sectioned for observation. The third set of experiments was done with a stainless steel bundle of 37 elements fabricated of mullite rods, 7.14 mm diameter. The stainless steel cladding had an O.D. of 8.41 mm. The element pitch was 11.1 mm. The grid spacers were prototypic. The experiment was conducted in an inert-gas tube furnace. The 'core fuel' cladding was melted in an induction furnace and the molten liquid flowed through the center seven element channels. X-ray pictures were taken after the tests and the bundle was sectioned for further study.

  13. International Standard Problems and Small Break Loss-Of-Coolant Accident (SBLOCA)

    International Nuclear Information System (INIS)

    Aksan, N.

    2008-01-01

    Best-estimate thermalhydraulic system codes are widely used to perform safety and licensing analyses of nuclear power plants and also used in the design of advance reactors. Evaluation of the capabilities and the performance of these codes can be accomplished by comparing the code predictions with measured experimental data obtained on different test facilities. In this respect, parallel to other national and international programs, OECD/Nea (OECD Nuclear Energy Agency) Committee on the Safety of Nuclear Installations (CSNI) has promoted, over the last twenty-nine years some forty-eight International Standard Problems (ISPs). These ISPs were performed in different fields as in-vessel thermalhydraulic behaviour, fuel behaviour under accident conditions, fission product release and transport, core/concrete interactions, hydrogen distribution and mixing, containment thermalhydraulic behaviour. 80% of these ISPs were related to the working domain of Principal Working Group no. 2 on Coolant System Behaviour (PWG2). The ISPs have been one of the major PWG2 activities for many years. The individual ISP comparison reports include the analysis and conclusions of the specific ISP exercises. A global review and synthesis on the contribution that ISPs have made to address nuclear reactor safety issues was initiated by CSNI-PWG2 and an overview on the subject of small break LOCA ISP's is given in this paper based on a report prepared by a CSNI-PWG2 writing group. In addition, the relevance of small break LOCA in a PWR with relation to nuclear reactor safety and the reorientation of the reactor safety program after TMI-2 accident, specifically small break LOCA, are shortly summarized. Five small break LOCA related ISP's are considered, since these were used for the assessment of the advanced best-estimate codes. The considered ISP's deal with the phenomenon typical of small break LOCAs in Western design PWRs. The experiments in four integral test facilities, LOBI, SPES, BETHSY

  14. International Standard Problems and Small Break Loss-of-Coolant Accident (SBLOCA

    Directory of Open Access Journals (Sweden)

    N. Aksan

    2008-01-01

    Full Text Available Best-estimate thermal-hydraulic system codes are widely used to perform safety and licensing analyses of nuclear power plants and also used in the design of advance reactors. Evaluation of the capabilities and the performance of these codes can be accomplished by comparing the code predictions with measured experimental data obtained on different test facilities. OECD/NEA Committee on the Safety of Nuclear Installations (CSNI has promoted, over the last twenty-nine years, some forty-eight international standard problems (ISPs. These ISPs were performed in different fields as in-vessel thermal-hydraulic behaviour, fuel behaviour under accident conditions, fission product release and transport, core/concrete interactions, hydrogen distribution and mixing, containment thermal-hydraulic behaviour. 80% of these ISPs were related to the working domain of principal working group no.2 on coolant system behaviour (PWG2 and were one of the major PWG2 activities for many years. A global review and synthesis on the contribution that ISPs have made to address nuclear reactor safety issues was initiated by CSNI-PWG2 and an overview on the subject of small break LOCA ISPs is given in this paper based on a report prepared by a writing group. In addition, the relevance of small break LOCA in a PWR with relation to nuclear reactor safety and the reorientation of the reactor safety program after TMI-2 accident are shortly summarized. The experiments in four integral test facilities, LOBI, SPES, BETHSY, ROSA IV/LSTF and the recorded data during a steam generator tube rupture transient in the DOEL-2 PWR (Belgium were the basis of the five small break LOCA related ISP exercises, which deal with the phenomenon typical of small break LOCAs in Western design PWRs. Some lessons learned from these small break LOCA ISPs are identified in relation to code deficiencies and capabilities, progress in the code capabilities, possibility of scaling, and various additional aspects

  15. Mechanical and Radiological Characterization of Different parts of an Irradiation Coolant Channel Tube from Atucha I Nuclear Plant; Caracterizacion Mecanica y Radiologica de Partes de Canales Refrigerantes Irradiados Extraidos del Reactor de la Central Nuclear Atucha I

    Energy Technology Data Exchange (ETDEWEB)

    Piquin, Ruben [Instituto Balseiro, Universidad Nacional de Cuyo, Centro Atomico Bariloche, Universidad Nacional de Mar del Plata (Argentina)

    2001-07-01

    The widespread replacement of reactor internals has generated a substantial volume of active material. It is essential to work with these components at least in a partial way before the next planned stop, which will take place during the second semester of the year 2002. Due to the fact that the reactor internals pool and the storage pool for irradiated nuclear fuel have limited capacities, it has been proposed to compact an experimental shift of 50 irradiated coolant channels, that are currently placed in storage pools. Basically the processed waste will be put in baskets at the bottom pools.The alternative choice proposes to divide an irradiation coolant channel tube into different parts: stainless steel section, zircaloy-4 section and stainless steel section with hardened zones with cobalt alloys named Estelite-6. The person in charge has already planned the constructive and operative solutions but the mechanical characterization of the different parts of the channel tube is necessary in order to dimension the compaction tool needed for the semi-industrial installation.In the present special report, two well-differentiated actions will be described. The necessary compacted strength of the irradiation coolant channel tube will be estimated for the stainless steel section and the zircaloy-4 section starting from experiment with unirradiated material and considering effects of radiation damage and hydrides on the ductility.These results will be used to design the necessary compacted tools for the semi-industrial installation. The necessary equipment for the radiological characterization of the different material sections already specified will be described and the most important emitting particles of radiation that could be detected will be mentioned. Also the decontamination process to use including the radiological characterization of every stage of the process will be described in order to establish the decontamination factor. Finally the most important

  16. Experimental investigation of boiling-water nuclear-reactor parallel-channel effects during a postulated loss-of-coolant accident

    International Nuclear Information System (INIS)

    Conlon, W.M.; Lahey, R.T. Jr.

    1982-12-01

    This report describes an experimental study of the influence of parallel channel effects (PCE) on the distribution of emergency core spray cooling water in a Boiling Water Nuclear Reactor (BWR) following a postulated design basis loss of coolant accident (LCA). The experiments were conducted in a scaled test section in which the reactor coolant was simulated by Freon-114 at conditions similar to those postulated to occur in the reactor vessel shortly after a LOCA. A BWR/4 was simulated by a (PCE) test section which contained three parallel heated channels to simulate fuel assemblies; a core bypass channel, and a jet pump channel. The test section also inlcuded scaled regions to simulate the lower and upper plena, downcomer, and steam separation regions of a BWR. A series of nine transient experiments were conducted, in which the lower plenum vaporization rate and heater rod power were varied while the core spray flow rate was held constant to simulate that of a BWR/4. During these experiments the flow distribution and heat transfer phenomena were observed and measured

  17. Research of the fluid flow in a radially orientated coolant channel of a turbine blade; Untersuchung der Stroemung in einem radial gerichteten Kuehlkanal eines Turbinenlaufrades

    Energy Technology Data Exchange (ETDEWEB)

    Hein, O.

    1999-07-01

    Due to rotation (Coriolis forces) in a coolant channel a secondary flow is superimposed to the basic flow. This leads to a change in the local heat transfer over the surface of the coolant channel as well as a change in the overall value of the heat transfer. Also the pressure loss over the channel length will change by rotation. By means of computational fluid dynamics (Finite Element Method) it was achieved to figure out the interaction between changing fluid flow and heat transfer. To validate the results obtained by a numerical flow simulation, a new measurement technique was developed. A laser-two-focus velocimeter has been combined with a rotation prism which allows continued measurements in a rotating scaled up channel. (orig.) [German] Bedingt durch die Rotationsbewegung eines Kuehlkanals wird die Grundstroemung von einem Sekundaerwirbel ueberlagert (Corioliskraefte). Durch diese Einfluesse aendert sich sowohl der lokale Waermeuebergang ueber der Kanaloberflaeche als auch die globalen Waermeuebertragungsraten ueber dem gesamten Kanal. Ebenfalls aendert sich durch die Rotation der Druckverlust ueber der Kanallaenge. Durch eine numerische Stroemungssimulation (Finite-Element-Methode) war es moeglich, einen detaillierten Zusammenhang zwischen dem veraenderten Stroemungsverhalten und dem Waermeuebertragungsverhalten darzustellen. Um die numerisch gewonnenen Ergebnisse experimentell abzusichern, wurde eine neuartige Messtechnik entwickelt. Ein Laser-2-Fokus-Velozimeter wurde mit einem Bilddrehprisma kombiniert, und dies erlaubte eine kontinuierliche Messung in einem rotierenden vergroesserten Modellkanal. (orig.)

  18. GLOBAL LOGISTICS AND INTERNATIONAL CHANNEL DEVELOPMENT

    Directory of Open Access Journals (Sweden)

    Fani Mateska

    2015-01-01

    Full Text Available Business Logistics and Supply Chain Management (SCM are relatively new terms that emerged in recent years concerning to be modern terms compared to the more traditional fields of production, marketing or finance. By opening new stores globally, from which to source products, by increasing the number of retailers, wholesalers, agents and distributers in the global supply chain, as well as easing global transport, these events dramatically changed the way business looked at managing its physical operations. Therefore, this paper offers deep understanding of the main characteristics, new trends and evaluation of global logistics which referred to as global supply chain management. Global logistics plays critical role in the growth and development of world trade, and in the integration of business operations on a worldwide scale. This paper also emphasizes the differences between traditional and global supply chains, by explaining several alternative types of distribution channels that are part of the international distribution channels. Also, real world examples of global organizations and their global supply chain management processes are accurately described in the end of this paper.

  19. Specificities of reactor coolant pumps units with lead and lead-bismuth coolant

    International Nuclear Information System (INIS)

    Beznosov, A.V.; Anotonenkov, M.A.; Bokov, P.A.; Baranova, V.S.; Kustov, M.S.

    2009-01-01

    The analysis results of impact of lead and lead-bismuth coolants specific properties on the coolants flow features in flow channels of the main and auxiliary circulating pumps are presented. Impossibility of cavitation initiation in flow channels of vane pumps pumping lead and lead-bismuth coolants was demonstrated. The experimental research results of discontinuity of heavy liquid metal coolant column were presented and conditions of gas cavitation initiation in coolant flow were discussed. Invalidity of traditional calculation methods of water and sodium coolants circulation pumps calculations for lead and lead-bismuth coolants circulation pumps was substantiated [ru

  20. Continuous internal channels formed in aluminum fusion welds

    Science.gov (United States)

    Gault, J.; Sabo, W.

    1967-01-01

    Process produces continuous internal channel systems on a repeatable basis in 2014-T6 aluminum. Standard machining forms the initial channel, which is filled with tungsten carbide powder. TIG machine fusion welding completes formation of the channel. Chem-mill techniques enlarge it to the desired size.

  1. International Space Station Active Thermal Control Sub-System On-Orbit Pump Performance and Reliability Using Liquid Ammonia as a Coolant

    Science.gov (United States)

    Morton, Richard D.; Jurick, Matthew; Roman, Ruben; Adamson, Gary; Bui, Chinh T.; Laliberte, Yvon J.

    2011-01-01

    The International Space Station (ISS) contains two Active Thermal Control Sub-systems (ATCS) that function by using a liquid ammonia cooling system collecting waste heat and rejecting it using radiators. These subsystems consist of a number of heat exchangers, cold plates, radiators, the Pump and Flow Control Subassembly (PFCS), and the Pump Module (PM), all of which are Orbital Replaceable Units (ORU's). The PFCS provides the motive force to circulate the ammonia coolant in the Photovoltaic Thermal Control Subsystem (PVTCS) and has been in operation since December, 2000. The Pump Module (PM) circulates liquid ammonia coolant within the External Active Thermal Control Subsystem (EATCS) cooling the ISS internal coolant (water) loops collecting waste heat and rejecting it through the ISS radiators. These PM loops have been in operation since December, 2006. This paper will discuss the original reliability analysis approach of the PFCS and Pump Module, comparing them against the current operational performance data for the ISS External Thermal Control Loops.

  2. Basic experimental study with visual observation on elimination of the re-criticality issue using the MELT-II facility. Simulated fuel-escape behavior through a coolant channel

    International Nuclear Information System (INIS)

    Matsuba, Ken-ichi; Imahori, Shinji; Isozaki, Mikio

    2004-11-01

    In a core disruptive accident of fast reactors, fuel escape from the reactor core is a key phenomenon for prevention of re-criticality with significant mechanical-energy release subsequent to formation of a large-scale fuel pool with high mobility. Therefore, it is effective to study possibility of early fuel escape through probable escape paths such as a control-rod-guide-tube space well before high-mobility-pool formation. The purpose of the present basic experimental study is to clarify the mechanism of fuel-escape under a condition expected in the reactor situation, in which some amount of coolant may be entrapped into the molten-fuel pool. The following results have been obtained through basic experiments in which molten Wood's metal (components: 60wt%Bi-20wt%Sn-20wt%In, density at the room temperature: 8700 kg/m 3 , melting point: 78.8degC) is ejected into an coolant channel filled with water. (1) In the course of melt ejection, a small quantity of coolant is forced to be entrapped into the melt pool as a result of thermal interactions leading to high-pressure rise within the coolant channel. (2) Melt ejection is accelerated by pressure build-up which results from vapor pressure of entrapped coolant within the melt pool. (3) Average melt-ejection rate tends to increase in lower coolant-subcooling conditions, in which pressure build-up within the melt pool is enhanced. These results indicate a probability of a phenomenon in which melt ejection is accelerated by entrapment of coolant within a melt pool. Through application of the mechanism of confirmed phenomenon into the reactor condition, it is suggested that fuel escape is enhanced by entrapment of coolant within a fuel pool. (author)

  3. Dark matter annihilation with s-channel internal Higgsstrahlung

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Jason; Liao, Jiajun, E-mail: liaoj@hawaii.edu; Marfatia, Danny

    2016-08-10

    We study the scenario of fermionic dark matter that annihilates to standard model fermions through an s-channel axial vector mediator. We point out that the well-known chirality suppression of the annihilation cross section can be alleviated by s-channel internal Higgsstrahlung. The shapes of the cosmic ray spectra are identical to that of t-channel internal Higgsstrahlung in the limit of a heavy mediating particle. Unlike the general case of t-channel bremsstrahlung, s-channel Higgsstrahlung can be the dominant annihilation process even for Dirac dark matter. Since the s-channel mediator can be a standard model singlet, collider searches for the mediator are easily circumvented.

  4. Dark matter annihilation with s-channel internal Higgsstrahlung

    International Nuclear Information System (INIS)

    Kumar, Jason; Liao, Jiajun; Marfatia, Danny

    2016-01-01

    We study the scenario of fermionic dark matter that annihilates to standard model fermions through an s-channel axial vector mediator. We point out that the well-known chirality suppression of the annihilation cross section can be alleviated by s-channel internal Higgsstrahlung. The shapes of the cosmic ray spectra are identical to that of t-channel internal Higgsstrahlung in the limit of a heavy mediating particle. Unlike the general case of t-channel bremsstrahlung, s-channel Higgsstrahlung can be the dominant annihilation process even for Dirac dark matter. Since the s-channel mediator can be a standard model singlet, collider searches for the mediator are easily circumvented.

  5. Chemical aspects of hydrogen ingress in zirconium and zircaloy pressure tubes: ageing management of Indian PHWR coolant channels - determination of hydrogen and deuterium

    International Nuclear Information System (INIS)

    Sayi, Y.S.; Shankaran, P.S.; Yadav, C.S.; Ramanjaneyulu, P.S.; Venugopal, V.; Ramakumar, K.L.; Chhapru, G.C.; Prasad, R.; Jain, H.C.; Sood, D.D.

    2009-02-01

    Pressurized heavy water reactors (PHWRs) use zirconium and zirconium based alloys as clad and coolant tubes since its beginning. The first ever zircaloy-2 pressure tube failure occurred in 1983 at Ontario Hydro's Pickering Unit 2 in Canada which necessitated a thorough examination of causes of such failure. The failure was attributed to massive hydriding at the failed spot of pressure tube. Continuous usage of zirconium alloys could result in their hydrogen and deuterium pick-up leading to hydrogen/ deuterium embrittlement. The life of the zircaloy coolant channels is dictated by hydrogen/deuterium content and hence ageing management of the pressure tubes is essential for ensuring their trouble-free usage. It is desirable to have a sound knowledge on the chemical aspects of zirconium and zirconium based alloys metallurgy, the mechanistic principles of hydrogen ingress into the pressure tubes during in reactor service, and identifying suitable analytical methodologies for precise and accurate determination of hydrogen in wafer thin sliver samples carved out from insides of pressure tubes without causing any structural damage so that it can continue to remain in service. This is desirable so that the ageing management does not result in cost-escalation. This report is divided in to three main parts. The first part deals with the chemical aspects of zirconium and zirconium based alloy metallurgy, the mechanism of hydrogen pick-up and hydride formation in zirconium matrix. The second part describes various methodologies and their limitations, available for hydrogen/deuterium determination. The third part deals in detail, about the extensive investigations carried out at Radioanalytical Chemistry Division (RACD) in Radiochemistry and Isotope Group for establishing an indigenously developed hot vacuum extraction system in combination with quadrupole mass spectrometry for precise determination of hydrogen and deuterium in wafer thin sliver sample of zircaloy. The

  6. Knock-Limited Power Outputs from a CFR Engine Using Internal Coolants. 3; Four Alkyl Amines, Three Alkanolamines, Six Amides, and Eight Heterocyclic Compounds

    Science.gov (United States)

    Imming, Harry S.; Bellman, Donald R.

    1947-01-01

    An investigation of the antiknock effectiveness of various additive-water solutions when used as internal coolants has been conducted at the NACA Cleveland laboratory. Nine compounds have been previously run in a CFR engine and the results are presented. In an effort to find a good anti-knock-coolant additive with more desirable physical properties than those of the nine compounds previously investigated, water solutions of four alkyl amines, three alkanolamines, six amides, and eight heterocyclic compounds were investigated and the results are presented.

  7. Experimental analysis of upward vertical two-phase flow in four-cusp channels simulating the conditions of a typical nuclear reactor channel, degraded by a loss of coolant accident

    International Nuclear Information System (INIS)

    Assad, A.C.A.

    1984-01-01

    The present work deals with an experimental analysis of upward vertical two-phase flow in channels with circular and four-cusp cross-sections. The latter simulates the conditions of a typical nuclear reactor channel, degraded by a loss of coolant accident. Simultaneous flow of air and water has been employed to simulate adiabatic steam-water flow. The installation of air-water separators helped eliminate instabilities during pressure-drop measurements. The gamma ray attenuation was utilized for the void fraction determination. For the four-cusp geommetry, new criteria for two-phase flow regime transitions have been determined, as well as new correlatins for pressure drop and void fraction, as function of the Lockhart-Martinelli factor and vapour mass-fraction, respectively. (Author) [pt

  8. Measurement of vibrations in the primary coolant circuit and in the vertical experimental channel of the RA reactor

    International Nuclear Information System (INIS)

    Ristic, B.; Rakic, R.; Milosevic, M.; Jerkovic, M.

    1966-01-01

    Full text: Beginning of the work dates from 1962 with the initial objective: study of the wear-out of the bearings of the centrifugal pumps in the heavy water system. It has been expected that the increase of wear-out would initiate increase of vibration amplitudes and noise. During further study the initial task was broadened to other fields, mainly appearance of material fatigue in components of the heavy water coolant system. During operation mechanical energy is generated due to non existing equilibrium of the pump rotor, wear-out of the bearing, turbulence in the pump, cavitation process and pulsation of the operating environment. This energy is transformed into noise and vibration energy which is spread through surrounding walls and pipes causing noise finally. Obtained results were only qualitatively tested at present. For quantitative testing it would be necessary to obtain data about the material, in addition to the diagrams obtained by measurements. It would be possible to calculate the fatigue of the material at measuring points as well as estimation of the time when material fatigue would become critical [sr

  9. The numerical simulation of the WWER-440/V-213 reactor pressure vessel internals response to maximum hypothetical large break loss of coolant accident

    International Nuclear Information System (INIS)

    Hermansky, P.; Krajcovic, M.

    2012-01-01

    The reactor internals are designed to ensure cooling of the fuel, to ensure the movement of emergency control assemblies under all operating conditions including accidents and facilitate removal of the fuel and of the internals following an accident This paper presents results of the numerical simulation of the WWER-440/V213 reactor vessel internals dynamic response to maximum hypothetical Large-Break Loss of Coolant Accident. The purpose of this analysis is to determine the reactor vessel internals response due to rapid depressurization and to prove no such deformations occur in the reactor vessel internals which would prevent timely and proper activation of the emergency control assemblies. (Authors)

  10. RELAP simulation and experimental verification of transient boiling conditions in narrow coolant channels, at low temperature and pressure

    International Nuclear Information System (INIS)

    Kunze, J.F.; Loyalka, S.K.; Hultsch, R.A.; Oladiran, O.; McKibben, J.C.

    1990-01-01

    This paper reports on benchmark experiments needed to verify the accuracy of thermal hydraulic codes (such as RELAP5/MOD2) with respect to their capability to simulate transient boiling conditions both with and without a closed recirculation path in narrow channels, under essentially atmospheric pressure conditions characteristic of plate-type research reactors. An experimental apparatus with this objective has been constructed, and data for surface heat flux of 1.2 x 10 5 w/m 2 are reported

  11. Heat transfer characteristics of rectangular coolant channels with various aspect ratios in the plasma-facing components under fully developed MHD laminar flow

    International Nuclear Information System (INIS)

    Takase, K.; Hasan, M.Z.

    1995-01-01

    Convective heat transfer in MHD laminar flow through rectangular channels in the plasma-facing components of a fusion reactor has been analyzed numerically to investigate the effects of channel aspect ratio, defined as the ratio of the lengths of the plasma-facing side to the other side. The adverse effect of the nonuniformity of surface heat flus on Nusselt number (Nu) at the plasma-facing side can be alleviated by increasing the aspect ratio of a rectangular duct. At the center and corner of the plasma-facing side of a square duct, the Nu of non-MHD flow are 6.8 and 2.2, respectively, for uniform surface heat flux. In the presence of a strong magnetic field, Nu at the center and corner increases to 22 and 3.6, respectively. However, when the heat flux is highly nonuniform, as in the plasma-facing components, Nu decreases from 22 to 3.1 at the center and from 3.6 to 3.1 at the corner. When the aspect ratio is increased to 4, Nu at the center and corner increase to 5 and 4.7. Along the circumference of a rectangular channel, there are locations where the wall temperature is equal to or less than the bulk coolant temperature, thus making the Nu with conventional definition infinity or negative. The ratio between Nu of MHD flow and Nu of non-MHD flow for various aspect ratios is constant in the region of Hartmann number of more than 200 at least. On the other hand, its ratio increases monotonously with increasing the aspect ratio

  12. Single-phase coolant flow CFD simulations inside the CANDU channel for the 37 and the 43 elements bundles

    International Nuclear Information System (INIS)

    Pauna, E.; Olteanu, G.; Catana, A.

    2013-01-01

    In this paper, a Computation Fluid Dynamics (CFD) simulation was performed in order to find the flow conditions in the CANDU Channel for the standard (37 elements) and the new designed bundle (43 elements) using the CFD Code S aturne software. Due to the fact that the code is a single-phase one it was considered an inlet temperature of 250 O C, a flow rate of 24.17 kg/s, an outlet pressure of 10.3 MPa and a linear power of 800 kW/m. The flow conditions were achieved by using a CFD typical chain of steps which was performed starting from preprocessing (geometry, mesh and boundary conditions), through solver and post-processing. Open Source platform (Salome-Meca geometry and mesh modules, the Code S aturne solver, Paraview and Visit for post-processing) were used as computational tool kit and an unsteady state was considered. Some simplifications were considered: the tube creep was not taken into account and all the bundles were considered aligned. The three dimensional thermal-hydraulic distributions of the temperature, pressure and velocity parameters offered information for the geometry comparison and the results were in agreement with some experimental data. CFD analysis results provided valuable data regarding the thermal-hydraulic operating conditions inside the CANDU reactor channel. (authors)

  13. Variegated operation of MAPS reactors after enmasse' coolant channel replacement: a tale-tell signature of high standard fuel bundle production quality

    International Nuclear Information System (INIS)

    Jena, J.K.; Sahu, J.K.; Arularasan, V.; Sivagurnathan, D.; Rathakrishnan, S.; Ramamurthy, K.

    2009-01-01

    After the Enmasse' Coolant Channel Replacement (EMCCR) of both the reactors of Madras Atomic Power Station (MAPS), they have put up a good performance, as far as core integrity is considered. This is a tale-tell signature of the high quality of the fuel bundles manufactured by Nuclear Fuel Complex (NFC), Hyderabad. Both the reactor cores have been loaded with various types of fuel bundles viz. Natural Uranium (NU), Depleted Uranium (DU), and Deeply Depleted Uranium (DDU) and were operated at different power level with different flux configuration at different stages of operation. Even around 1026 low burn up bundle (<2500 MWD/TeU) were transferred from MAPS-1 to MAPS-2, first time in the history of PHWRS. During all such variegated operations, the Primary Heat Transport (PHT) system 131 I activity, which is synonymous with the core integrity, was maintaining low for most of the reactor operation period. However, recently a low burn up fuel bundle failure has been observed in MAPS-1. Even though the overall failure rate is very low, the cause of such failure needs to be ascertained for taking appropriate action to maintain the high standards of quality in the manufacturing process of the fuel bundles. (author)

  14. Evaluation of CNA I coolant channel behaviour during an accidental transient using ICARE2 V2 mod2.3 code

    International Nuclear Information System (INIS)

    Marino, Edgardo J.L.

    1999-01-01

    Using the input data language of ICARE2 V2 Mod.3 code, the fuel element and coolant channel assembly of CNA I type was described. This input data was utilized to analyze the system behavior and determine the degradation produced during a hypothetical accidental transient at CNA I. The boundary conditions were determined through a previous calculation with RELAP5/MOD 3.2 code. The results had shown characteristic degradation phenomena's. The temperature of bundle components increases fast after 6.11 h in the first case and 5.28 h in the second case, due to the energy release by cladding oxidation. It was correlated with instantaneous hydrogen production and energy contribution. The cumulated hydrogen production was estimated as 0.15 Kg in the first case and ∼ 5 times greater in the second case. Fission product release from the gap due to cladding rupture took place from 6.25 h in the first case and 5.65 h in the second. Relocation started after 6.81 h in the first case and 5.68 in the second, because the cladding dislocation condition is reached. UO 2 dissolution by molten Zircaloy was observed at different levels in the calculation domain. (author)

  15. Work related to increasing the exploitation and experimental possibilities of the RA reactor, 05. Independent CO2 loop for cooling the samples irradiated in the RA vertical experimental channels (I-IV), Part II, IZ-240-0379-1963, Vol. II Head of the low temperature RA reactor coolant loop

    International Nuclear Information System (INIS)

    Pavicevic, M.

    1963-07-01

    The objective of the project was to design the head of the CO 2 coolant loop for cooling the materials during irradiation in the RA reactor. Six heads of coolant loops will be placed in the RA reactor, two in the region of heavy water in the experimental channels VEK-6 and four in the graphite reflector in the channels VEK-G. maximum generated heat in the heads of the coolant loop is 10500 kcal/h and minimum generated heat is 1500 kcal/h. The loops are cooled by CO 2 gas, coolant flow is 420 kg/h, and the pressure is 4.5 atu. There is a need to design and construct the secondary coolant loop for the low temperature coolant loop. This volume includes technical specifications of the secondary CO 2 loop with instructions for construction and testing; needed calculations; specification of materials; cost estimation for materials, equipment and construction; and graphical documentation [sr

  16. Work related to increasing the exploitation and experimental possibilities of the RA reactor, 05. Independent CO2 loop for cooling the samples irradiated in the RA vertical experimental channels (IIV), Part I, IZ-240-o379-1963, Vol. I, Head of the low temperature RA reactor coolant loop

    International Nuclear Information System (INIS)

    Pavicevic, M.

    1963-07-01

    The objective of the project was to design the head of the CO 2 coolant loop for cooling the materials during irradiation in the RA reactor. Six heads of coolant loops will be placed in the RA reactor, two in the region of heavy water in the experimental channels VEK-6 and four in the graphite reflector in the channels VEK-G. Materials for irradiation are metallurgy and chemical samples. In addition to the project objectives, this volume includes technical specifications of the coolant loop head, thermal calculations, calculations of mechanical stress, antireactivity and activation of the construction materials, cost estimation, scheme of the coolant loop head, diagrams of CO 2 gas temperature, thermal neutron flux distribution, design specifications of two proposed solutions for head of low temperature coolant loop [sr

  17. Computer program MCAP-TOSS calculates steady-state fluid dynamics of coolant in parallel channels and temperature distribution in surrounding heat-generating solid

    Science.gov (United States)

    Lee, A. Y.

    1967-01-01

    Computer program calculates the steady state fluid distribution, temperature rise, and pressure drop of a coolant, the material temperature distribution of a heat generating solid, and the heat flux distributions at the fluid-solid interfaces. It performs the necessary iterations automatically within the computer, in one machine run.

  18. Theoretical Calculations of the Effect on Lattice Parameters of Emptying the Coolant Channels in a D{sub 2}O- Moderated and Cooled Natural Uranium Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Weissglas, P [The Swedish State Power Board, Stockholm (Sweden)

    1960-11-15

    The purpose of the present study was to evaluate theoretically the effect of coolant boiling and subsequent void formation in a pressurized D{sub 2}O moderated and cooled reactor. The fuel rods were arranged in a cluster geometry and clad in Zr-2. The coolant was separated from the moderator by a Zr-2 shroud. In this geometry the following problems have been given special attention: l) calculation of the effective resonance integral, 2) thermal disadvantage factors, 3) fast fission effects, 4) leakage effects, 5) changes in epithermal absorption. No account has up to now been taken of the variation of these effects with position in the reactor and burnup. Some comparisons of the theoretical methods and measurements have been attempted. It is concluded that at the present time it is not possible to calculate the void coefficient with any accuracy but it may be possible to give an upper limit from theoretical consideration.

  19. Analytical evaluation of local fault in sodium cooled small fast reactor (4S). Preliminary evaluation of partial blockage in coolant channel

    International Nuclear Information System (INIS)

    Nishimura, Satoshi; Ueda, Nobuyuki

    2007-01-01

    Local faults are fuel failures that result from heat removal imbalance within a single subassembly especially in FBRs. Although the occurrence frequency of local faults is quite low, the licensing body required local faults evaluations in previous FBR plants to confirm the potential for the occurrence of severe fuel subassembly failure and its propagation. A conceptual design of 4S (Super-Safe, Small and Simple) is a sodium cooled fast reactor, which aims at an application to dispersed energy source and long core lifetime. It has a dense arrangement of fuel pins to achieve a long lifetime. Therefore, from the viewpoint of thermal hydraulics, the 4S reactor is considered to have more potential for coolant boiling and fuel pin failure caused by formation of local blockage, comparing these potential in the conventional FBRs. The objective of the present study is to evaluate the effect of local blockage on the coolant flow pattern and temperature rise in the 4S-type fuel subassembly under the normal operation condition. A series of three-dimensional thermal-hydraulic analysis in a single subassembly with local blockage was conducted by the commercialized CFD code 'PHOENICS'. Analytical results show that the peak coolant temperature behind the blockage rises with increasing the blockage area, however, the coolant boiling does not occur under the present analytical conditions. On the other hand, it is found that the liquid phase formation caused by eutectic reactions will occur between the metallic fuel and the cladding under the local blockage condition. However, the penetration rate of liquid phase at fuel-cladding interface is quit low. Therefore, it is expected that rapid fuel pin failure and its propagation to surrounding pins due to liquid phase formation will not occur. (author)

  20. Thermal fluid dynamic behavior of coolant helium gas in a typical reactor VHTGR channel of prismatic core; Comportamento termofluidodinamico do gas refrigerante helio em um canal topico de reator VHTGR de nucleo prismatico

    Energy Technology Data Exchange (ETDEWEB)

    Belo, Allan Cavalcante

    2016-08-01

    The current studies about the thermal fluid dynamic behavior of the VHTGR core reactors of 4{sup th} generation are commonly developed in 3-D analysis in CFD (computational fluid dynamics), which often requires considerable time and complex mathematical calculations for carrying out these analysis. The purpose of this project is to achieve thermal fluid dynamic analysis of flow of gas helium refrigerant in a typical channel of VHTGR prismatic core reactor evaluating magnitudes of interest such as temperature, pressure and fluid velocity and temperature distribution in the wall of the coolant channel from the development of a computer code in MATLAB considering the flow on one-dimensional channel, thereby significantly reducing the processing time of calculations. The model uses three different references to the physical properties of helium: expressions given by the KTA (German committee of nuclear safety standards), the computational tool REFPROP and a set of constant values for the entire channel. With the use of these three references it is possible to simulate the flow treating the gas both compressible and incompressible. The results showed very close values for the interest quantities and revealed that there are no significant differences in the use of different references used in the project. Another important conclusion to be observed is the independence of helium in the gas compressibility effects on thermal fluid dynamic behavior. The study also indicated that the gas undergoes no severe effects due to high temperature variations in the channel, since this goes in the channel at 914 K and exits at approximately 1263 K, which shows the excellent use of helium as a refrigerant fluid in reactor channels VHTGR. The comparison of results obtained in this work with others in the literature served to confirm the effectiveness of the one-dimensional consideration of method of gas flow in the coolant channel to replace the models made in 3-D for the pressure range

  1. Calculation of activity concentration and dose rates from online radioactivity measurement in primary coolant channel of TAPS-III and IV

    International Nuclear Information System (INIS)

    Chaudhury, Sanhita; Agarwal, Chhavi; Goswami, A.; Mhatre, Amol; Chaturvedi, T.P.; Tawde, N.; Gathibandhe, Manohar; Dash, S.C.

    2011-05-01

    Radioactivity measurement using CdZnTe detector and dose measurement using teletector were done at several locations of primary heat transport (PHT) system of the Tarapur Atomic Power Station-III and IV reactor during shut down as well as operating condition of the reactors. The detector efficiency for the required counting geometry was simulated using MCNP code. Using this simulated efficiency and the experimental count rate (cps), the activity concentrations (Bq/mL) of different radionuclides in coolant water were calculated. The dose rates for the counted locations were also simulated using Monte Carlo code and it matched well with the experimentally obtained dose rate. (author)

  2. Coolant leakage detection device

    International Nuclear Information System (INIS)

    Ito, Takao.

    1983-01-01

    Purpose: To surely detect the coolant leakage at a time when the leakage amount is still low in the intra-reactor inlet pipeway of FBR type reactor. Constitution: Outside of the intra-reactor inlet piping for introducing coolants at low temperature into a reactor core, an outer closure pipe is furnished. The upper end of the outer closure pipe opens above the liquid level of the coolants in the reactor, and a thermocouple is inserted to the opening of the upper end. In such a structure, if the coolants in the in-reactor piping should leak to the outer closure pipe, coolants over-flows from the opening thereof, at which the thermocouple detects the temperature of the coolants at a low temperature, thereby enabling to detect the leakage of the coolants at a time when it is still low. (Kamimura, M.)

  3. HANARO secondary coolant management

    International Nuclear Information System (INIS)

    Kim, Seon Duk.

    1998-02-01

    In this report, the basic theory for management of water quality, environmental factors influencing to the coolant, chemicals and its usage for quality control of coolant are mentioned, and water balance including the loss rate by evaporation (34.3 m 3 /hr), discharge rate (12.665 m 3 /hr), concentration ratio and feed rate (54.1 m 3 /hr) are calculated at 20 MW operation. Also, the analysis data of HANSU Limited for HANARO secondary coolant (feed water and circulating coolant) - turbidity, pH, conductivity, M-alkalinity, Ca-hardness, chloride ion, total iron ion, phosphoric ion and conversion rate are reviewed. It is confirmed that the feed water has good quality and the circulating coolant has been maintained within the control specification in general, but some items exceeded the control specification occasionally. Therefore it is judged that more regular discharge of coolant is needed. (author). 6 refs., 17 tabs., 18 figs

  4. Study of Internal Channel Surface Roughnesses Manufactured by Selective Laser Melting in Aluminum and Titanium Alloys

    Science.gov (United States)

    Pakkanen, Jukka; Calignano, Flaviana; Trevisan, Francesco; Lorusso, Massimo; Ambrosio, Elisa Paola; Manfredi, Diego; Fino, Paolo

    2016-08-01

    Interest in additive manufacturing (AM) has gained considerable impetus over the past decade. One of the driving factors for AM success is the ability to create unique designs with intrinsic characteristics as, e.g., internal channels used for hydraulic components, cooling channels, and heat exchangers. However, a couple of the main problems in internal channels manufactured by AM technologies are the high surface roughness obtained and the distortion of the channel shape. There is still much to understand in these design aspects. In this study, a cylindrical geometry for internal channels to be built with different angles with respect to the building plane in AlSi10Mg and Ti6Al4V alloys by selective laser melting was considered. The internal surfaces of the channels produced in both materials were analyzed by means of a surface roughness tester and by optical and electron microscopy to evaluate the effects of the material and design choices.

  5. Modeling the spatial distribution of the parameters of the coolant in the reactor volume

    International Nuclear Information System (INIS)

    Nikonov, S.P.

    2011-01-01

    In this paper the approach to the question about the spatial distribution of the parameters of the coolant in-reactor volume. To describe the in-core space is used specially developed preprocessor. When the work of the preprocessor in the first place, is recreated on the basis of available information (mostly-the original drawings) with high accuracy three-dimensional description of the structures of the reactor volume and, secondly, are prepared on this basis blocks input to the nodal system code improved estimate ATHLET, allows to take into account the hydrodynamic interaction between the spatial control volumes. As an example the special case of solutions of international standard problem on the reconstruction of the transition process in the third unit of the Kalinin nuclear power plant, due to the shutdown of one of the four Main Coolant Pumps in operation at the rated capacity (first download). Model-core area consists of approximately 58 000 control volumes and spatial relationships. It shows the influence of certain structural units of the core to the distribution of the mass floe rate of its height. It is detected a strong cross-flow coolant in the area over the baffle. Moreover, we study the distribution of the coolant temperature at the assembly head of WWER-1000 reactor. It is shown that in the region of the top of the assembly head, where we have installation of thermocouples, the flow coolant for internal assemblies core is formed by only from guide channel Reactor control and protected system Control rod flow, or a mixture of the guide channel flow and flow from the area in front of top grid head assembly (the peripheral assemblies). It is shown that the magnitude of the flow guide channels affects not only the position of control rods, but also the presence of a particular type of measuring channels (Self powered neutron detector sensors or Temperature control sensors) in the cassette. (Author)

  6. Multirods burst tests under loss-of-coolant conditions

    International Nuclear Information System (INIS)

    Kawasaki, S.; Uetsuka, H.; Furuta, T.

    1983-01-01

    In order to know the upper limit of coolant flow area restriction in a fuel assembly under loss-of-coolant accidents in LWRs, burst tests of fuel bundles were performed. Each bundle consisted of 49 rods(7x7 rods), and bursts were conducted in flowing steam. In some cases, 4 rods were replaced by control rods with guide tubes in a bundle. After the burst, the ballooning behavior of each rod and the degree of coolant flow area restriction in the bundle were measured. Ballooning behavior of rods and degree of coolant flow channel restriction in bundles with control rods were not different from those without control rods. The upper limit of coolant flow channel restriction under loss-of-coolant conditions was estimated to be about 80%. (author)

  7. CANDU with supercritical water coolant: conceptual design features

    International Nuclear Information System (INIS)

    Spinks, N.

    1997-01-01

    An advanced CANDU reactor, with supercritical water as coolant, has many attractive design features. The pressure exceeds 22 MPa but coolant temperatures in excess of 370 degrees C can be reached without encountering the two-phase region with its associated fuel-dry-out and flow-instability problems. Increased coolant temperature leads to increased plant thermodynamic efficiency reducing unit energy cost through reduced specific capital cost and reduced fueling cost. Increased coolant temperature leads to reduced void reactivity via reduced coolant in-core density. Light water becomes a coolant option. To preserve neutron economy, an advanced fuel channel is needed and is described below. A supercritical-water-cooled CANDU can evolve as fuel capabilities evolve to withstand increasing coolant temperatures. (author)

  8. International and Pan-European News Channels: An overview

    OpenAIRE

    Thierry Vissol

    2005-01-01

    The European television industry has evolved in a dramatic way since the beginning of the 90’s. The number of channels has been multiplied by 12, the share of the public television decreased to around 30 %, and the number of thematic channels is in thousands. In this general landscape, generalist TV channels (public or private), that still hold the bulk of the audience are nevertheless losing ground. In the same time the trust of their viewers in the quality of the information transmitted in ...

  9. The power of perceived service quality in international marketing channels

    NARCIS (Netherlands)

    Ruyter, de J.C.; Wetzels, M.G.M.; Lemmink, J.G.A.M.

    1996-01-01

    States that providing quality customer service is increasingly regarded as an important basis for establishing and maintaining solid and long-term relationships between manufacturers and distributors in marketing channels. Discusses the results of a research project which was undertaken to study the

  10. The sodium coolant

    International Nuclear Information System (INIS)

    Rodriguez, G.

    2004-01-01

    The sodium is the best appropriate coolant for the fast neutrons reactors technology. Thus the fast neutrons reactors development is intimately bound to the sodium technology. This document presents the sodium as a coolant point of view: atomic structure and characteristics, sodium impacts on the fast neutron reactors technology, chemical properties of the sodium and the consequences, quality control in a nuclear reactor, sodium treatment. (A.L.B.)

  11. Extended Life Coolant Testing

    Science.gov (United States)

    2016-06-06

    number. PLEASE DO NOT RETURN YOUR FORM TO THE ABOVE ADDRESS. 1. REPORT DATE (DD-MM-YYYY) 06-06-2016 2. REPORT TYPE Interim Report 3. DATES COVERED ... Corrosion Testing of Traditional and Extended Life Coolants 5b. GRANT NUMBER 5c. PROGRAM ELEMENT NUMBER 6. AUTHOR(S) Hansen, Gregory A. T...providing vehicle specific coolants. Several laboratory corrosion tests were performed according to ASTM D1384 and D2570, but with a 2.5x extended time

  12. Research on coolant radiochemistry

    International Nuclear Information System (INIS)

    Yeon, Jei Won; Kim, W. H.; Park, Y. J.; Im, J. K.; Jung, Y. J.; Jee, K. Y.; Choi, K. C.

    2004-04-01

    The final objective of this study is to develop the technology on the reduction of radioactive material formed in reactor coolant circuit. The contents of this study are composed of the simulation of primary cooling system, chemistry measurement technology in the high-temperature high-pressure environments, and coolant chemistry control technology. The main results are as follows; High-temperature and high-pressure loop system was designed and fabricated, which is to inducing CRUD growth condition on the surface of cladding. The high-temperature pH measurement system was established with YSZ sensing electrode and Ag/AgCl reference electrode. The performance of pH electrode was confirmed in the temperature range 200∼280 .deg. C. Coolant chemistry control technologies such as the neutron irradiation technique of boric acid solution, the evaluation on high-temperature electrochemical behavior of coolant, and the measurement of physicochemical properties of micro-particles were developed. The results of this study can be useful for the understanding of chemical phenomena occurred in reactor coolant and for the study on the reduction of radioactive material in primary coolant, which will be carried out in the next research stage

  13. User's guide to EPIC, a computer program to calculate the motion of fuel and coolant subsequent to pin failure in an LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Pizzica, P.A.; Garner, P.L.; Abramson, P.B.

    1979-10-01

    The computer code EPIC models fuel and coolant motion which results from internal fuel pin pressure (from fission gas or fuel vapor) and possibly from the generation of sodium vapor pressure in the coolant channel subsequent to pin failure in a liquid-metal fast breeder reactor. The EPIC model is restricted to conditions where fuel pin geometry is generally preserved and is not intended to treat the total disruption of the pin structure. The modeling includes the ejection of molten fuel from the pin into a coolant channel with any amount of voiding through a clad breach which may be of any length or which may extend with time. One-dimensional Eulerian hydrodynamics is used to treat the motion of fuel and fission gas inside a molten fuel cavity in the fuel pin as well as the mixture of two-phase sodium and fission gas in the coolant channel. Motion of fuel in the coolant channel is tracked with a type of particle-in-cell technique. EPIC is a Fortran-IV program requiring 400K bytes of storage on the IBM 370/195 computer. 21 refs., 2 figs.

  14. 1st International Conference on Hydraulic Design in Water Resources Engineering : Channels and Channel Control Structures

    CERN Document Server

    1984-01-01

    The development of water resources has proceeded at an amazing speed around the world in the last few decades. The hydraulic engineer has played his part: in constructing much larger artificial channels than ever before, larger and more sophisticated control structures, and systems of irrigation, drainage and water supply channels in which the flow by its nature is complex and unsteady requiring computer-based techniques at both the design and operation stage. It seemed appropriate to look briefly at some of the developments in hydraulic design resulting from this situation. Hence the idea of the Conference was formed. The Proceedings of the Conference show that hydraulic engineers have been able to acquire a very substantial base of design capability from the experience of the period referred to. The most outstanding development to have occurred is in the combination of physical and mathematical modelling, which in hydraulic engineering has followed a parallel path to that in other branches of engineering sc...

  15. Coolant leakage detecting device

    International Nuclear Information System (INIS)

    Yamauchi, Kiyoshi; Kawai, Katsunori; Ishihara, Yoshinao.

    1995-01-01

    The device of the present invention judges an amount of leakage of primary coolants of a PWR power plant at high speed. Namely, a mass of coolants contained in a pressurizer, a volume controlling tank and loop regions is obtained based on a preset relational formula and signals of each of process amount, summed up to determine the total mass of coolants for every period of time. The amount of leakage for every period of time is calculated by a formula of Karman's filter based on the total mass of the primary coolants for every predetermined period of time, and displays it on CRT. The Karman's filter is formed on every formula for several kinds of states formed based on the preset amount of the leakage, to calculate forecasting values for every mass of coolants. An adaptable probability for every preset leakage amount is determined based on the difference between the forecast value and the observed value and the scattering thereof. The adaptable probability is compared with a predetermined threshold value, which is displayed on the CRT. This device enables earlier detection of leakage and identification of minute leakage amount as compared with the prior device. (I.S.)

  16. Comparison of THALES and VIPRE-01 Subchannel Codes for Loss of Flow and Single Reactor Coolant Pump Rotor Seizure Accidents using Lumped Channel APR1400 Geometry

    Energy Technology Data Exchange (ETDEWEB)

    Oezdemir, Erdal; Moon, Kang Hoon; Oh, Seung Jong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of); Kim, Yongdeog [KHNP-CRI, Daejeon (Korea, Republic of)

    2014-10-15

    Subchannel analysis plays important role to evaluate safety critical parameters like minimum departure from nucleate boiling ratio (MDNBR), peak clad temperature and fuel centerline temperature. In this study, two different subchannel codes, VIPRE-01 (Versatile Internals and Component Program for Reactors: EPRI) and THALES (Thermal Hydraulic AnaLyzer for Enhanced Simulation of core) are examined. In this study, two different transient cases for which MDNBR result play important role are selected to conduct analysis with THALES and VIPRE-01 subchannel codes. In order to get comparable results same core geometry, fuel parameters, correlations and models are selected for each code. MDNBR results from simulations by both code are agree with each other with negligible difference. Whereas, simulations conducted by enabling conduction model in VIPRE-01 shows significant difference from the results of THALES.

  17. Reactor coolant cleanup device

    International Nuclear Information System (INIS)

    Igarashi, Noboru.

    1986-01-01

    Purpose: To enable to introduce reactor water at high temperature and high pressure as it is, as well as effectively adsorb to eliminate cobalt in reactor water. Constitution: The coolant cleanup device comprises a vessel main body inserted to coolant pipeway circuits in a water cooled reactor power plant and filters contained within the vessel main body. The filters are prepared by coating and baking powder of metal oxides such as manganese ferrite having a function capable of adsorbing cobalt in the coolants onto the surface of supports made of metals or ceramics resistant to strong acids and alkalies in the form of three-dimensional network structure, for example, zircaloy-2, SUS 303 and the zirconia (baking) to form a basic filter elements. The basic filter elements are charged in plurality to the vessel main body. (Kawaiami, Y.)

  18. Challenges Related to the Use of Liquid Metal and Molten Salt Coolants in Advanced Reactors. Report of the collaborative project COOL of the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

    International Nuclear Information System (INIS)

    2013-05-01

    The International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) was launched in 2000, based on a resolution by the IAEA General Conference (GC(44)/RES/21). INPRO aims at helping to ensure that nuclear energy is available in the twenty-first century in a sustainable manner, and seeks to bring together all interested Member States, both technology holders and technology users, to jointly consider actions to achieve desired innovations. INPRO is taking care of the specific needs of developing countries. One of the aims of INPRO is to develop options for enhanced sustainability through promotion of technical and institutional innovations in nuclear energy technology through collaborative projects among IAEA Member States. Collaboration among INPRO members is fostered on selected innovative nuclear technologies to bridge technology gaps. Collaborative projects have been selected so that they complement other national and international R and D activities. The INPRO Collaborative Project COOL on Investigation of Technological Challenges Related to the Removal of Heat by Liquid Metal and Molten Salt Coolants from Reactor Cores Operating at High Temperatures investigated the technological challenges of cooling reactor cores that operate at high temperatures in advanced fast reactors, high temperature reactors and accelerator driven systems by using liquid metals and molten salts as coolants. The project was initiated in 2008 and was led by India; experts from Brazil, China, Germany, India, Italy and the Republic of Korea participated and provided chapters of this report. The INPRO Collaborative Project COOL addressed the following fields of research regarding liquid metal and molten salt coolants: (i) survey of thermophysical properties; (ii) experimental investigations and computational fluid dynamics studies on thermohydraulics, specifically pressure drop and heat transfer under different operating conditions; (iii) monitoring and control of coolant

  19. Reactor having coolant recycling pump

    International Nuclear Information System (INIS)

    Goto, Tadashi; Karatsuka, Shigeki; Yamamoto, Hajime.

    1991-01-01

    In a coolant recycling pump for an LMFBR type reactor, vertical grooves are formed to a static portion which surrounds a pump shaft as far as the lower end thereof. Sodium mists present in an annular gap of the pump shaft form a rotational flow, lose its centrifugal force at the grooved portion and are collected positively to the grooved portion. Further, since the rotational flow in the grooved channel is in a state of a cavity flow, the pressure is released in the grooved portion and a secondary eddy current is formed thereby providing a depressurized state. Accordingly, by a synergestic effect of the centrifugal force and the cavity flow, sodium mists can be recovered completely. (T.M.)

  20. Reactor Coolant Temperature Measurement using Ultrasonic Technology

    Energy Technology Data Exchange (ETDEWEB)

    Jung, JaeCheon [KEPCO International Nuclear graduate School, Ulsan (Korea, Republic of); Seo, YongSun; Bechue, Nicholas [Krohne Messtechnik GmbH, Duisburg (Germany)

    2016-10-15

    In NPP, the primary piping temperature is detected by four redundant RTDs (Resistance Temperature Detectors) installed 90 degrees apart on the RCS (Reactor Coolant System) piping circumferentially. Such outputs however, if applied to I and C systems would not give balanced results. The discrepancy can be explained by either thermal stratification or improper arrangement of thermo-wells and RTDs. This phenomenon has become more pronounced in the hot-leg piping than in the cold-leg. Normally, the temperature difference among channels is in the range of 1°F in Korean nuclear power Plants. Consequently, a more accurate pipe average temperate measurement technique is required. Ultrasonic methods can be used to measure average temperatures with relatively higher accuracy than RTDs because the sound wave propagation in the RCS pipe is proportional to the average temperature around pipe area. The inaccuracy of RCS temperature measurement worsens the safety margin for both DNBR and LPD. The possibility of this discrepancy has been reported with thermal stratification effect. Proposed RCS temperature measurement system based on ultrasonic technology offers a countermeasure to cope with thermal stratification effect on hot-leg piping that has been an unresolved issue in NPPs. By introducing ultrasonic technology, the average internal piping temperature can be measured with high accuracy. The inaccuracy can be decreased less than ±1℉ by this method.

  1. Coolant system decontamination

    International Nuclear Information System (INIS)

    Anstine, L.D.; James, D.B.; Melaika, E.A.; Peterson, J.P.

    1981-01-01

    An improved method for decontaminating the coolant system of water cooled nuclear power reactors and for regenerating the decontamination solution is described. A small amount of one or more weak-acid organic complexing agents is added to the reactor coolant, and the pH is adjusted to form a decontamination solution which is circulated throughout the coolant system to dissolve metal oxides from the interior surfaces and complex the resulting metal ions and radionuclide ions. The coolant containing the complexed metal ions and radionuclide ions is passed through a strong-base anion exchange resin bed which has been presaturated with a solution containing the complexing agents in the same ratio and having the same pH as the decontamination solution. As the decontamination solution passes through the resin bed, metal-complexed anions are exchanged for the metal-ion-free anions on the bed, while metal-ion-free anions in the solution pass through the bed, thus removing the metal ions and regenerating the decontamination solution. (author)

  2. Research on Coolant Radiochemistry

    International Nuclear Information System (INIS)

    Ha, Yeong Keong; Kim, W. H.; Yeon, J. W.; Jung, Y. J.; Choi, K. C.; Choi, K. S.; Park, Y. J.; Cho, Y. H.

    2007-06-01

    The final objective of this study is to develop a method for reducing radioactive materials formed in the reactor coolant circuit. This second stage research was categorized into the following three subgroups: the development of the estimation technique of microscopic chemical variation at high temperatures and pressures, the fundamental study on the thermodynamics at high temperatures and pressures, and the study on the deposition of metal oxides and the determination of the main factors responsible for the growth of CRUD. First, in the development of the estimation technique of microscopic chemical change at high temperatures and pressures, the technique for measuring coolant chemistry such as pH, conductivity and Eh was developed to be appropriate for the high temperature and pressure condition. The coolant chemistry measuring system including the self-devised high temperature pH sensor can be applied to the field of nuclear reactor and contribute on a large scale in the automation of the coolant chemistry control and the establishment of the real-time on-line measuring technique. Secondly, the dissociation constant of water and the solubility of metal oxides were measured in the fundamental study on the thermodynamics at high temperatures and pressures. Finally, in the study on the deposition of metal oxides and the determination of the main factors responsible for the growth of CRUD, the careful investigation of the deposition phenomena of micro particles on the cladding surface showed that subcooled boiling and the dissolved hydrogen are the main factors responsible for the growth of CRUD. In addition, the basis was provided for the construction of a new particle behavior model in the reactor coolant circuit

  3. ISP 33. OECD/NEA/CSNI International Standard Problem n. 33. Pactel natural circulation stepwise coolant inventory reduction experiment. Comparison report. Volume 1 + 2

    International Nuclear Information System (INIS)

    Purhonen, H.; Kouhia, J.; Holmstrom, H.

    1994-12-01

    This is the comparison report of the CSNI ISP n.33, which is based on a natural circulation experiment with various coolant inventories conducted in Pactel facility (Finland), a 1/305 volumetrically scaled, full-height simulator of a Russian type VVER-440 pressurized water reactor. It presents all submitted blind calculational results from different countries, using various codes (Athlet, Cathare2, etc.) and compares them with the experimental data. The Pactel facility and the ISP 33 experiment are described, and the summaries of the participants, the computer codes and the nodalizations used for the blind calculations are presented

  4. Nuclear reactor of pressurized liquid coolant type

    International Nuclear Information System (INIS)

    Costes, D.

    1976-01-01

    The reactor comprises a vertical concrete pressure vessel, a bell-housing having an open lower end and disposed coaxially with the interior of the pressure vessel so as to delimit therewith a space filled with gas under pressure for the thermal insulation of the internal vessel wall, a pressurizing device for putting the coolant under pressure within the bell-housing and comprising a volume of control gas in contact with a large free surface of coolant in order that an appreciable variation in volume of liquid displaced within the coolant circuit inside the bell-housing should correspond to a small variation in pressure of the control gas. 9 claims, 3 drawing figures

  5. LWR primary coolant pipe rupture test rig

    International Nuclear Information System (INIS)

    Yoshitoshi, Shyoji

    1978-01-01

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

  6. New light on ion channel imaging by total internal reflection fluorescence (TIRF) microscopy.

    Science.gov (United States)

    Yamamura, Hisao; Suzuki, Yoshiaki; Imaizumi, Yuji

    2015-05-01

    Ion channels play pivotal roles in a wide variety of cellular functions; therefore, their physiological characteristics, pharmacological responses, and molecular structures have been extensively investigated. However, the mobility of an ion channel itself in the cell membrane has not been examined in as much detail. A total internal reflection fluorescence (TIRF) microscope allows fluorophores to be imaged in a restricted region within an evanescent field of less than 200 nm from the interface of the coverslip and plasma membrane in living cells. Thus the TIRF microscope is useful for selectively visualizing the plasmalemmal surface and subplasmalemmal zone. In this review, we focused on a single-molecule analysis of the dynamic movement of ion channels in the plasma membrane using TIRF microscopy. We also described two single-molecule imaging techniques under TIRF microscopy: fluorescence resonance energy transfer (FRET) for the identification of molecules that interact with ion channels, and subunit counting for the determination of subunit stoichiometry in a functional channel. TIRF imaging can also be used to analyze spatiotemporal Ca(2+) events in the subplasmalemma. Single-molecule analyses of ion channels and localized Ca(2+) signals based on TIRF imaging provide beneficial pharmacological and physiological information concerning the functions of ion channels. Copyright © 2015 The Authors. Production and hosting by Elsevier B.V. All rights reserved.

  7. An Analysis of Freight Forwarder Operations in an International Distribution Channel.

    Science.gov (United States)

    1987-01-01

    44 3. International Marketing Mix ....................... 45 4. Security Assistance Distribution Channel .......... 69 5...an item is ultimately derived from the interaction of variables in the marketing mix . Of those variables, the distribution functions seem to allow the...Component of the Marketing Mix ,"Proceedings, NCPDM Fall Meeting, National council of Physical Distribution Management, San Francisco, CA., 1982. 7

  8. Compartmentalized safety coolant injection system

    International Nuclear Information System (INIS)

    Johnson, F.T.

    1983-01-01

    A safety coolant injection system for nuclear reactors wherein a core reflood tank is provided to afford more reliable reflooding of the reactor core in the event of a break in one of the reactor coolant supply loops. Each reactor coolant supply loop is arranged in a separate compartment in the containment structure to contain and control the flow of spilled coolant so as to permit its use during emergency core cooling procedures. A spillway allows spilled coolant in the compartment to pass into the emergency water storage tank from where it can be pumped back to the reactor vessel. (author)

  9. Channel follower leakage restrictor

    International Nuclear Information System (INIS)

    Williamson, H.E.; Smith, B.A.

    1977-01-01

    An improved means is provided to control coolant leakage between the flow channel and the lower tie plate of a nuclear fuel assembly. The means includes an opening in the lower tie plate and a movable element adjacent thereto. The coolant pressure within the tie plate biases the movable means toward the inner surface of the surrounding flow channel to compensate for any movement of the flow channel away from the lower tie plate to thereby control the leakage of coolant flow from the fuel assemblies to the spaces among the fuel assemblies of the core. 9 figures

  10. Coolant Void Reactivity Analysis of CANDU Lattice

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jin Su; Lee, Hyun Suk; Tak, Tae Woo; Lee, Deok Jung [UNIST, Ulsan (Korea, Republic of)

    2016-05-15

    Models of CANDU-6 and ACR-700 fuel lattices were constructed for a single bundle and 2 by 2 checkerboard to understand the physics related to CVR. Also, a familiar four factor formula was used to predict the specific contributions to reactivity change in order to achieve an understanding of the physics issues related to the CVR. At the same time, because the situation of coolant voiding should bring about a change of neutron behavior, the spectral changes and neutron current were also analyzed. The models of the CANDU- 6 and ACR-700 fuel lattices were constructed using the Monte Carlo code MCNP6 using the ENDF/B-VII.0 continuous energy cross section library based on the specification from AECL. The CANDU fuel lattice was searched through sensitivity studies of each design parameter such as fuel enrichment, fuel pitch, and types of burnable absorber for obtaining better behavior in terms of CVR. Unlike the single channel coolant voiding, the ACR-700 bundle has a positive reactivity change upon 2x2 checkerboard coolant voiding. Because of the new path for neutron moderation, the neutrons from the voided channel move to the no-void channel where they lose energy and come back to the voided channel as thermal neutrons. This phenomenon causes the positive CVR when checkerboard voiding occurs. The sensitivity study revealed the effects of the moderator to fuel volume ratio, fuel enrichment, and burnable absorber on the CVR. A fuel bundle with low moderator to fuel volume ratio and high fuel enrichment can help achieve negative CVR.

  11. Platinum atom location on the internal walls of nanocavities investigated by ion channeling analysis

    International Nuclear Information System (INIS)

    Kinomura, A.; Williams, J.S.; Tsubouchi, N.; Horino, Y.

    2002-01-01

    Atomic locations of Pt trapped at hydrogen-induced cavities in Si have been investigated by ion channeling analysis. A Pt dose of 1x10 14 cm -2 , corresponding to a monolayer coverage of the internal walls of cavities, was implanted into cavity-containing samples. The gettering of Pt to the cavities was induced by annealing at 850 deg. C for 1 h. Clear channeling effects were observed in aligned and random backscattering spectra for the , and axes. Angular yield profiles for three crystalline axes exhibited dips with a narrowing of Pt signal half width compared with the Si matrix. Results suggested that the Pt atoms trapped at the cavities are closely aligned with the Si atomic strings bounding axial channels in Si

  12. Sensitivity calculation of the coolant temperature regarding the thermohydraulic parameters

    International Nuclear Information System (INIS)

    Andrade Lima, F.R. de; Silva, F.C. da; Thome Filho, Z.D.; Alvim, A.C.M.; Oliveira Barroso, A.C. de.

    1985-01-01

    It's studied the application of the Generalized Perturbation Theory (GPT) in the sensitivity calculation of thermalhydraulic problems, aiming at verifying the viability of the extension of the method. For this, the axial distribution, transient, of the coolant temperature in a PWR channel are considered. Perturbation expressions are developed using the GPT formalism, and a computer code (Tempera) is written, to calculate the channel temperature distribution and the associated importance function, as well as the effect of the thermalhydraulic parameters variations in the coolant temperature (sensitivity calculation). The results are compared with those from the direct calculation. (E.G.) [pt

  13. Triboengineering problems of lead coolant in innovative fast reactors

    International Nuclear Information System (INIS)

    Beznosov, A.V.; Novozhilova, O.O.; Shumilkov, A.I.; Lvov, A.V.; Bokova, T.A.; Makhov, K.A.

    2013-01-01

    Graphical abstract: Models of experimental sites for research of processes tribology in heavy liquid metal coolant. -- Highlights: • The contact a pair of heavy liquid metal coolant for reactors on fast neutrons. • The hydrostatic bearings main circulation pumps. • Oxide coating and degree of wear of friction surfaces in heavy liquid metal coolant. -- Abstract: So far, there are plenty of works dedicated to studying the phenomenon of friction. However, there are none dedicated to functioning of contact pairs in heavy liquid-metal coolants for fast neutron, reactor installations (Kogaev and Drozdov, 1991; Modern Tribology, 2008; Drozdov et al., 1986). At the Nizhny Novgorod State Technical University, such research is conducted in respect to friction, bearings of main circulating pumps, interaction of sheaths of neutron absorber rods with their covers, of the reactor control and safety system, refueling systems, and interaction of coolant flows with, channel borders. As a result of experimental studies, the characteristic of friction pairs in the heavy, liquid metal coolant shows the presence dependences of oxide film on structural materials of the wear. The inapplicability of existing calculation methods for assessing the performance of the bearing nodes, in the heavy liquid metal coolant is shown

  14. PKA-induced internalization of slack KNa channels produces dorsal root ganglion neuron hyperexcitability.

    Science.gov (United States)

    Nuwer, Megan O; Picchione, Kelly E; Bhattacharjee, Arin

    2010-10-20

    Inflammatory mediators through the activation of the protein kinase A (PKA) pathway sensitize primary afferent nociceptors to mechanical, thermal, and osmotic stimuli. However, it is unclear which ion conductances are responsible for PKA-induced nociceptor hyperexcitability. We have previously shown the abundant expression of Slack sodium-activated potassium (K(Na)) channels in nociceptive dorsal root ganglion (DRG) neurons. Here we show using cultured DRG neurons, that of the total potassium current, I(K), the K(Na) current is predominantly inhibited by PKA. We demonstrate that PKA modulation of K(Na) channels does not happen at the level of channel gating but arises from the internal trafficking of Slack channels from DRG membranes. Furthermore, we found that knocking down the Slack subunit by RNA interference causes a loss of firing accommodation analogous to that observed during PKA activation. Our data suggest that the change in nociceptive firing occurring during inflammation is the result of PKA-induced Slack channel trafficking.

  15. Device for preventing leakage of coolant in nuclear fuel assembly

    International Nuclear Information System (INIS)

    Kobayashi, Yukio; Sekiguchi, Mamoru; Yoshida, Hideo.

    1975-01-01

    Object: To prevent leakage of coolant from between lower tie plate and channel box without causing deformation of the channel box and also without the possibility of disturbing the installation and removal of the box by the provision of a thin plate provided with leakage holes for the lower tie plate. Structure: Static water pressure within the lower tie plate is adapted to act upon the bear side of a flat plate for leakage prevention through leakage holes formed in the tie plate, thus urging the flat plate against the channel box inner surface. Meanwhile, static water pressure having been led through the leakage holes in the flat plate is adapted to press the flat plate in the vertical direction, thus urging the flat plate against the channel box inner surface and thereby preventing leakage of the coolant through a gap between the channel box and lower tie plate. (Yoshino, Y.)

  16. SGS Modeling of the Internal Energy Equation in LES of Supersonic Channel Flow

    Science.gov (United States)

    Raghunath, Sriram; Brereton, Giles

    2011-11-01

    DNS of fully-developed turbulent supersonic channel flows (Reτ = 190) at up to Mach 3 indicate that the turbulent heat fluxes depend only weakly on Mach number, while the viscous dissipation and pressure dilatation do so strongly. Moreover, pressure dilatation makes a significant contribution to the internal energy budget at Mach 3 and higher. The balance between these terms is critical to determining the temperature (and so molecular viscosity) from the internal energy equation and so, in LES of these flows, it is essential to use accurate SGS models for the viscous dissipation and the pressure dilatation. In this talk, we present LES results for supersonic channel flow, using SGS models for these terms that are based on the resolved-scale dilatation, an inverse timescale, and SGS momentum fluxes, which intrinsically represent this Mach number effect.

  17. Improving Coolant Effectiveness through Drill Design Optimization in Gundrilling

    Science.gov (United States)

    Woon, K. S.; Tnay, G. L.; Rahman, M.

    2018-05-01

    Effective coolant application is essential to prevent thermo-mechanical failures of gun drills. This paper presents a novel study that enhances coolant effectiveness in evacuating chips from the cutting zone using a computational fluid dynamic (CFD) method. Drag coefficients and transport behaviour over a wide range of Reynold numbers were first established through a series of vertical drop tests. With these, a CFD model was then developed and calibrated with a set of horizontal drilling tests. Using this CFD model, critical drill geometries that lead to poor chip evacuation including the nose grind contour, coolant hole configuration and shoulder dub-off angle in commercial gun drills are identified. From this study, a new design that consists a 20° inner edge, 15° outer edge, 0° shoulder dub-off and kidney-shaped coolant channel is proposed and experimentally proven to be more superior than all other commercial designs.

  18. Exchange rate pass-through and the role of international distribution channels

    OpenAIRE

    Desiraju, Ramarao; Shrikhande, Milind

    1996-01-01

    Manufacturers selling in foreign markets often do not completely pass on the effects of fluctuations in exchange rates to the prices of their products. Our paper addresses this puzzle and studies the effects of the international distribution channel on exchange rate pass-through. We develop an exchange rate pass-through model that takes into account the role of an intermediary between a domestic manufacturer and its consumers in a foreign market. We find that the magnitude of the pass-through...

  19. Subtype-specific, bi-component inhibition of SK channels by low internal pH

    DEFF Research Database (Denmark)

    Peitersen, Torben; Jespersen, Thomas; Jorgensen, Nanna K

    2006-01-01

    The effects of low intracellular pH (pH(i) 6.4) on cloned small-conductance Ca2+-activated K+ channel currents of all three subtypes (SK1, SK2, and SK3) were investigated in HEK293 cells using the patch-clamp technique. In 400 nM internal Ca2+ [Ca2+]i, all subtypes were inhibited by pH(i) 6...

  20. Dual coolant blanket concept

    International Nuclear Information System (INIS)

    Malang, S.; Schleisiek, K.

    1994-11-01

    A self-cooled liquid metal breeder blanket with helium-cooled first wall ('Dual Coolant Blanket Concept') for a fusion DEMO reactor is described. This is one of the four blanket concepts under development in the frame of the European fusion technology program with the aim to select in 1995 the two most promising ones for further development. Described are the design of the blankets including the ancillary loop system and the results of the theoretical and experimental work in the fields of neutronics, magnetohydrodynamics, thermohydraulics, mechanical stresses, compatibility and purification of lead-lithium, tritium control, safety, reliability, and electrically insulating coatings. The remaining open questions and the required R and D programme are identified. (orig.) [de

  1. Hydrodynamics and heat transfer in reactor components cooled by liquid metal coolants in single/two phase. 11. meeting of the International Association for Hydraulic Research (IAHR) Working Group. Working material

    International Nuclear Information System (INIS)

    2005-01-01

    This Working Material includes the papers presented at the International Meeting 'Hydrodynamics and heat transfer in reactor components cooled by liquid metal coolants in single/two-phase', which was held 5-9 July 2004 at the State Scientific Center of Russian Federation - Institute for Physics and Power Engineering named after A.I. Leypunsky, in Obninsk near Moscow. The objectives of the meeting were to discuss new results obtained in the field of liquid metal coolant and to recommend the lines of further general physics and applied investigations, with the purpose of validating existing and codes under development for liquid metal cooled advanced and new generation nuclear reactors. Most of the contributions present results of experimental and numerical investigations into velocity, temperature and heat transfer in fuel subassemblies of fast reactors cooled by sodium or lead. In the frame of the meeting a benchmark problem devoted to heat transfer in the model subassembly of the fast reactor BREST-OD-300 was proposed. Experts from 5 countries (Japan, Netherlands, Spain, Republic of Korea, and Russia) took part in this benchmark exercise. The results of the benchmark calculations are summarized in the Working Material. The results of hydrodynamic studies of pressure head chambers and collector systems of liquid metal cooled reactors are presented in a number of papers. Also attention was given to the generalization of experimental data on hydraulic losses in the pipelines in case of mutual influence of local pressure drops, and to the modeling of natural convection in the fuel subassemblies and circuits with liquid metal cooling. Special emphasis at the meeting was placed on thermal hydraulics issues related to the development and design of target systems, such as heat removal in the target unit of the cascade subcritical reactor cooled by liquid salt; the target complex MK-1 for accelerator driven systems cooled by eutectic lead-bismuth alloy; and the test

  2. Hydrodynamics and heat transfer in reactor components cooled by liquid metal coolants in single/two phase. 11. meeting of the International Association for Hydraulic Research (IAHR) Working Group. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-07-01

    This Working Material includes the papers presented at the International Meeting 'Hydrodynamics and heat transfer in reactor components cooled by liquid metal coolants in single/two-phase', which was held 5-9 July 2004 at the State Scientific Center of Russian Federation - Institute for Physics and Power Engineering named after A.I. Leypunsky, in Obninsk near Moscow. The objectives of the meeting were to discuss new results obtained in the field of liquid metal coolant and to recommend the lines of further general physics and applied investigations, with the purpose of validating existing and codes under development for liquid metal cooled advanced and new generation nuclear reactors. Most of the contributions present results of experimental and numerical investigations into velocity, temperature and heat transfer in fuel subassemblies of fast reactors cooled by sodium or lead. In the frame of the meeting a benchmark problem devoted to heat transfer in the model subassembly of the fast reactor BREST-OD-300 was proposed. Experts from 5 countries (Japan, Netherlands, Spain, Republic of Korea, and Russia) took part in this benchmark exercise. The results of the benchmark calculations are summarized in the Working Material. The results of hydrodynamic studies of pressure head chambers and collector systems of liquid metal cooled reactors are presented in a number of papers. Also attention was given to the generalization of experimental data on hydraulic losses in the pipelines in case of mutual influence of local pressure drops, and to the modeling of natural convection in the fuel subassemblies and circuits with liquid metal cooling. Special emphasis at the meeting was placed on thermal hydraulics issues related to the development and design of target systems, such as heat removal in the target unit of the cascade subcritical reactor cooled by liquid salt; the target complex MK-1 for accelerator driven systems cooled by eutectic lead-bismuth alloy; and the test

  3. Enhancement of centrifugal compressor stable operation using an internal recirculation channel

    Directory of Open Access Journals (Sweden)

    Růžička M.

    2016-12-01

    Full Text Available Centrifugal compressors with a high pressure ratio are widely used in small aircraft turbine engines and turbocharges. At high rotational speeds they have a narrow stable operating region and the commonly used impellers with back swept blades are not able to ensure requested stability. In order to achieve a wider stable operating region, some other anti-surge measures can be used such as an Internal Recirculation Channel (IRC located in the inlet area of the compressor impeller. This paper analyses the influence of various IRC inlet slot geometries on air flow parameters. The flow velocity profiles downstream of the IRC outlet slot are evaluated and the influence of various channel configurations on compressor performance map is shown. Results from experiments as well as CFD simulations enable us to analyse the airflow in IRC and find out its most suitable geometry.

  4. Influence of internal channel geometry of gas turbine blade on flow structure and heat transfer

    Science.gov (United States)

    Szwaba, Ryszard; Kaczynski, Piotr; Telega, Janusz; Doerffer, Piotr

    2017-12-01

    This paper presents the study of the influence of channel geometry on the flow structure and heat transfer, and also their correlations on all the walls of a radial cooling passage model of a gas turbine blade. The investigations focus on the heat transfer and aerodynamic measurements in the channel, which is an accurate representation of the configuration used in aeroengines. Correlations for the heat transfer coefficient and the pressure drop used in the design of internal cooling passages are often developed from simplified models. It is important to note that real engine passages do not have perfect rectangular cross sections, but include a corner fillets, ribs with fillet radii and a special orientation. Therefore, this work provides detailed fluid flow and heat transfer data for a model of radial cooling geometry which has very realistic features.

  5. CAREM-25: considerations about primary coolant chemistry

    International Nuclear Information System (INIS)

    Chocron, Mauricio; Iglesias, Alberto M.; Raffo Calderon, Maria C.; Villegas, Marina

    2000-01-01

    World operating experience, in conjunction with basic studies has been modifying chemistry specifications for the primary coolant of water cooled nuclear reactors along with the reactor type and structural materials involved in the design. For the reactor CAREM-25, the following sources of information have been used: 1) Experience gained by the Chemistry Department of the National Atomic Energy Commission (CNEA, Argentina); 2) Participation of the Chemistry Department (CNEA) in international cooperation projects; 3) Guidelines given by EPRI, Siemens-KWU, AECL, etc. Given the main objectives: materials integrity, low radiation levels and personnel safety, which are in turn a balance between the lowest corrosion and activity transport achievable and considering that the CAREM-25 is a pressurized vessel integrated reactor, a group of guidelines for the chemistry and additives for the primary coolant have been given in the present work. (author)

  6. BWR fuel assembly bottom nozzle with one-way coolant flow valve

    International Nuclear Information System (INIS)

    Taleyarkhan, R.P.

    1987-01-01

    In a nuclear reactor having a flow of coolant/moderator fluid therein, at least one fuel assembly installed in the fluid flow, the fuel assembly is described comprising in combination: a bundle of elongated fuel rods disposed in side-by-side relationship so as to form an array of spaced fuel rods; an outer tubular flow channel surrounding the fuel rods so as to direct the flow of coolant/moderator fluid along the fuel rods; bottom and top nozzles mounted at opposite ends of the flow channel and having an inlet and outlet respectively for allowing entry and exit of the flow of coolant/moderator fluid into and from the flow channel and along the fuel rods therein; and a coolant flow direction control device operatively disposed in the bottom nozzle so as to open the inlet thereof to the flow of coolant/moderator fluid in an inflow direction into the flow channel through the bottom nozzle inlet but close the inlet to the flow of coolant/moderator fluid from the flow channel through the bottom nozzle inlet upon reversal of coolant/moderator fluid flow from the inflow direction

  7. Secondary coolant purification system

    International Nuclear Information System (INIS)

    Stiteler, F.Z.; Donohue, J.P.

    1978-01-01

    The present invention combines the attributes of volatile chemical addition, continuous blowdown, and full flow condensate demineralization. During normal plant operation (defined as no primary to secondary leakage) condensate from the condenser is pumped through a full flow condensate demineralizer system by the condensate pumps. Volatile chemical additions are made. Dissolved and suspended solids are removed in the condensate polishers by ion exchange and/or filtration. At the same time a continuous blowdown of approximately 1 percent of the main steaming rate of the steam generators is maintained. Radiation detectors monitor the secondary coolant. If these monitors indicate no primary to secondary leakage, the blowdown is cooled and returned directly to the condensate pump discharge. If one of the radiation monitors should indicate a primary to secondary leak, when the temperature of the effluent exiting from the blowdown heat exchanger is compatible with the resin specifications of the ion exchangers, the bypass valve causes the blowdown flow to pass through the blowdown ion exchangers

  8. Single-phase convection heat transfer characteristics of pebble-bed channels with internal heat generation

    International Nuclear Information System (INIS)

    Meng Xianke; Sun Zhongning; Xu Guangzhan

    2012-01-01

    Graphical abstract: The core of the water-cooled pebble bed reactor is the porous channels which stacked with spherical fuel elements. The gaps between the adjacent fuel elements are complex because they are stochastic and often shift. We adopt electromagnetic induction heating method to overall heat the pebble bed. By comparing and analyzing the experimental data, we get the rule of power distribution and the rule of heat transfer coefficient with particle diameter, heat flux density, inlet temperature and working fluid's Re number. Highlights: ► We adopt electromagnetic induction heating method to overall heat the pebble bed to be the internal heat source. ► The ball diameter is smaller, the effect of the heat transfer is better. ► With Re number increasing, heat transfer coefficient is also increasing and eventually tends to stabilize. ► The changing of heat power makes little effect on the heat transfer coefficient of pebble bed channels. - Abstract: The reactor core of a water-cooled pebble bed reactor includes porous channels that are formed by spherical fuel elements. This structure has notably improved heat transfer. Due to the variability and randomness of the interstices in pebble bed channels, heat transfer is complex, and there are few studies regarding this topic. To study the heat transfer characters of pebble bed channels with internal heat sources, oxidized stainless steel spheres with diameters of 3 and 8 mm and carbon steel spheres with 8 mm diameters are used in a stacked pebble bed. Distilled water is used as a refrigerant for the experiments, and the electromagnetic induction heating method is used to heat the pebble bed. By comparing and analyzing the experimental results, we obtain the governing rules for the power distribution and the heat transfer coefficient with respect to particle diameter, heat flux density, inlet temperature and working fluid Re number. From fitting of the experimental data, we obtain the dimensionless average

  9. Pressure control of a proton beam-irradiated water target through an internal flow channel-induced thermosyphon.

    Science.gov (United States)

    Hong, Bong Hwan; Jung, In Su

    2017-07-01

    A water target was designed to enhance cooling efficiency using a thermosyphon, which is a system that uses natural convection to induce heat exchange. Two water targets were fabricated: a square target without any flow channel and a target with a flow channel design to induce a thermosyphon mechanism. These two targets had the same internal volume of 8 ml. First, visualization experiments were performed to observe the internal flow by natural convection. Subsequently, an experiment was conducted to compare the cooling performance of both water targets by measuring the temperature and pressure. A 30-MeV proton beam with a beam current of 20 μA was used to irradiate both targets. Consequently, the target with an internal flow channel had a lower mean temperature and a 50% pressure drop compared to the target without a flow channel during proton beam irradiation. Copyright © 2017 Elsevier Ltd. All rights reserved.

  10. Decontamination of main coolant pumps

    International Nuclear Information System (INIS)

    Roofthooft, R.

    1988-01-01

    Last year a number of main coolant pumps in Belgian nuclear power plants were decontaminated. A new method has been developed to reduce the time taken for decontamination and the volume of waste to be treated. The method comprises two phases: Oxidation with permanganate in nitric acid and dissolution in oxalic acid. The decontamination of main coolant pumps can now be achieved in less than one day. The decontamination factors attained range between 15 and 150. (orig.) [de

  11. Multi-state reliability for coolant pump based on dependent competitive failure model

    International Nuclear Information System (INIS)

    Shang Yanlong; Cai Qi; Zhao Xinwen; Chen Ling

    2013-01-01

    By taking into account the effect of degradation due to internal vibration and external shocks. and based on service environment and degradation mechanism of nuclear power plant coolant pump, a multi-state reliability model of coolant pump was proposed for the system that involves competitive failure process between shocks and degradation. Using this model, degradation state probability and system reliability were obtained under the consideration of internal vibration and external shocks for the degraded coolant pump. It provided an effective method to reliability analysis for coolant pump in nuclear power plant based on operating environment. The results can provide a decision making basis for design changing and maintenance optimization. (authors)

  12. Effects of Rotation at Different Channel Orientations on the Flow Field inside a Trailing Edge Internal Cooling Channel

    Directory of Open Access Journals (Sweden)

    Matteo Pascotto

    2013-01-01

    Full Text Available The flow field inside a cooling channel for the trailing edge of gas turbine blades has been numerically investigated with the aim to highlight the effects of channel rotation and orientation. A commercial 3D RANS solver including a SST turbulence model has been used to compute the isothermal steady air flow inside both static and rotating passages. Simulations were performed at a Reynolds number equal to 20000, a rotation number (Ro of 0, 0.23, and 0.46, and channel orientations of γ=0∘, 22.5°, and 45°, extending previous results towards new engine-like working conditions. The numerical results have been carefully validated against experimental data obtained by the same authors for conditions γ=0∘ and Ro = 0, 0.23. Rotation effects are shown to alter significantly the flow field inside both inlet and trailing edge regions. These effects are attenuated by an increase of the channel orientation from γ=0∘ to 45°.

  13. Rapid internalization of the oncogenic K+ channel K(V10.1.

    Directory of Open Access Journals (Sweden)

    Tobias Kohl

    Full Text Available K(V10.1 is a mammalian brain voltage-gated potassium channel whose ectopic expression outside of the brain has been proven relevant for tumor biology. Promotion of cancer cell proliferation by K(V10.1 depends largely on ion flow, but some oncogenic properties remain in the absence of ion permeation. Additionally, K(V10.1 surface populations are small compared to large intracellular pools. Control of protein turnover within cells is key to both cellular plasticity and homeostasis, and therefore we set out to analyze how endocytic trafficking participates in controlling K(V10.1 intracellular distribution and life cycle. To follow plasma membrane K(V10.1 selectively, we generated a modified channel of displaying an extracellular affinity tag for surface labeling by α-bungarotoxin. This modification only minimally affected K(V10.1 electrophysiological properties. Using a combination of microscopy and biochemistry techniques, we show that K(V10.1 is constitutively internalized involving at least two distinct pathways of endocytosis and mainly sorted to lysosomes. This occurs at a relatively fast rate. Simultaneously, recycling seems to contribute to maintain basal K(V10.1 surface levels. Brief K(V10.1 surface half-life and rapid lysosomal targeting is a relevant factor to be taken into account for potential drug delivery and targeting strategies directed against K(V10.1 on tumor cells.

  14. Analysis and prediction of dimensions and cost of laser micro-machining internal channel fabrication process

    Directory of Open Access Journals (Sweden)

    Brabazon D.

    2010-06-01

    Full Text Available This paper presents the utilisation of Response Surface Methodology (RSM as the prediction tool for the laser micro-machining process. Laser internal microchannels machined using pulsed Nd:YVO4 laser in polycarbonate were investigated. The experiments were carried out according to 33 factorial Design of Experiment (DoE. In this work the three input process set as control parameters were laser power, P; pulse repetition frequency, PRF; and sample translation speed, U. Measured responses were the channel width and the micro-machining operating cost per metre of produced microchannels. The responses were sufficiently predicted within the set micro-machining parameters limits. Two factorial interaction (2FI and quadratic polynomial regression equations for both responses were constructed. It is proposed that the developed prediction equations can be used to find locally optimal micro-machining process parameters under experimental and operational conditions.

  15. Reactor coolant pump transportation incident

    International Nuclear Information System (INIS)

    Noce, D.

    1992-01-01

    This paper reports on an incident, which occurred on August 27, 1991, in which a Reactor Coolant Pump motor en route from Surry Power Station to Westinghouse repair facilities struck the overpass at the junction of Interstate 64 and Jefferson Avenue in Newport News, Virginia. The transport container that housed the reactor coolant pump motor failed to clear the overpass. The force of the impact dislodged the container and motor from the truck bed, and it landed on the acceleration land and road shoulder. Upon impact, the container broke open and exposed the reactor coolant pump motor. Incidental radioactively contaminated water that remained in the motor coolers drained onto the road, contaminating the aggregate as well as the underlying gravel

  16. Reactor coolant flow measurements at Point Lepreau

    International Nuclear Information System (INIS)

    Brenciaglia, G.; Gurevich, Y.; Liu, G.

    1996-01-01

    The CROSSFLOW ultrasonic flow measurement system manufactured by AMAG is fully proven as reliable and accurate when applied to large piping in defined geometries for such applications as feedwater flows measurement. Its application to direct reactor coolant flow (RCF) measurements - both individual channel flows and bulk flows such as pump suction flow - has been well established through recent work by AMAG at Point Lepreau, with application to other reactor types (eg. PWR) imminent. At Point Lepreau, Measurements have been demonstrated at full power; improvements to consistently meet ±1% accuracy are in progress. The development and recent customization of CROSSFLOW to RCF measurement at Point Lepreau are described in this paper; typical measurement results are included. (author)

  17. FILM-30: A Heat Transfer Properties Code for Water Coolant

    International Nuclear Information System (INIS)

    MARSHALL, THERON D.

    2001-01-01

    A FORTRAN computer code has been written to calculate the heat transfer properties at the wetted perimeter of a coolant channel when provided the bulk water conditions. This computer code is titled FILM-30 and the code calculates its heat transfer properties by using the following correlations: (1) Sieder-Tate: forced convection, (2) Bergles-Rohsenow: onset to nucleate boiling, (3) Bergles-Rohsenow: partially developed nucleate boiling, (4) Araki: fully developed nucleate boiling, (5) Tong-75: critical heat flux (CHF), and (6) Marshall-98: transition boiling. FILM-30 produces output files that provide the heat flux and heat transfer coefficient at the wetted perimeter as a function of temperature. To validate FILM-30, the calculated heat transfer properties were used in finite element analyses to predict internal temperatures for a water-cooled copper mockup under one-sided heating from a rastered electron beam. These predicted temperatures were compared with the measured temperatures from the author's 1994 and 1998 heat transfer experiments. There was excellent agreement between the predicted and experimentally measured temperatures, which confirmed the accuracy of FILM-30 within the experimental range of the tests. FILM-30 can accurately predict the CHF and transition boiling regimes, which is an important advantage over current heat transfer codes. Consequently, FILM-30 is ideal for predicting heat transfer properties for applications that feature high heat fluxes produced by one-sided heating

  18. Sodium as a reactor coolant

    International Nuclear Information System (INIS)

    Cesar, S.B.G.

    1989-01-01

    This work is related to the use of sodium as a reactor coolant, to the advantages and problems related to its use, its mechanical, thermophysics, eletronical, magnetic and nuclear properties. It is mainly a bibliographic review, with the aim of gathering the necessary information to persons initiating in the study of sodium and also as reference source. (author) [pt

  19. Vertical reactor coolant pump instabilities

    International Nuclear Information System (INIS)

    Jones, R.M.

    1985-01-01

    The investigation conducted at the Tennessee Valley Authority's Sequoyah Nuclear Power Plant to determine and correct increasing vibrations in the vertical reactor coolant pumps is described. Diagnostic procedures to determine the vibration causes and evaluate the corractive measures taken are also described

  20. Calculation of coolant temperature sensitivity related to thermohydraulic parameters

    International Nuclear Information System (INIS)

    Silva, F.C. da; Andrade Lima, F.R. de

    1985-01-01

    It is verified the viability to apply the generalized Perturbation Theory (GPT) in the calculation of sensitivity for thermal-hydraulic problems. It was developed the TEMPERA code in FORTRAN-IV to transient calculations in the axial temperature distribution in a channel of PWR reactor and the associated importance function, as well as effects of variations of thermalhydraulic parameters in the coolant temperature. The results are compared with one which were obtained by direct calculation. (M.C.K.) [pt

  1. Experimental study on single-phase convection heat transfer characteristics of pebble bed channels with internal heat generation

    International Nuclear Information System (INIS)

    Meng Xianke; Sun Zhongning; Zhou Ping; Xu Guangzhan

    2012-01-01

    The water-cooled pebble bed reactor core is the porous channels stacked with spherical fuel elements, having evident effect on enhancing heat transfer. Owing to the variability and randomness characteristics of it's interstice, pebble bed channels have a very complex heat transfer situation and have little correlative research. In order to research the heat transfer characters of pebble bed channels with internal heat source, electromagnetic induction heating method was adopted for overall heating the pebble bed which was composed of 8 mm diameter steel balls, and the internal heat transfer characteristics were researched. By comparing and analyzing the experimental data, the rule of power distribution and heat transfer coefficient with heat flux density, inlet temperature and working fluid's Re were got. According to the experimental data fitting, the dimensionless average heat transfer coefficient correlation criteria was got. The fitting results are good agreement with the experimental results within 12% difference. (authors)

  2. Advances in Forecasting and Prevention of Resonances Between Coolant Acoustical Oscillations and Fuel Rod Vibrations

    Energy Technology Data Exchange (ETDEWEB)

    Proskuryakov, Konstantin Nicolaevich [NPP, NPEI, 14, Krasnokazarmennaya str. Moscow, 111250 (Russian Federation)

    2009-06-15

    To prevent the appearance of the conditions for resonance interaction between the fluid flow and the reactor internals (RI), fuel rod (FR ) and fuel assemblies (FA) it is necessary to de-tune Eigen frequency of coolant pressure oscillations (EFCPO) and natural frequency of mechanical element's oscillations and also of the system which is formed by the comprising of these elements. Other words it is necessary to de-tune acoustic resonance frequency and natural frequencies of RI, FR and FA. While solving these problems it is necessary to have a theoretical and settlement substantiation of an oscillation frequency band of the coolant outside of which there is no resonant interaction with structure vibrations. The presented work is devoted to finding the solution of this problem. There are results of an estimation of width of such band as well as the examples of a preliminary quantitative estimation of Q - factors of coolant acoustic oscillatory circuit formed by the equipment of the NPP. Abnormal growth of intensity of pressure pulsations in a mode with definite value of reactor capacity have been found out by measurements on VVER - 1000 reactor. This phenomenon has been found out casually and its original reason had not been identified. Paper shows that disappearance of this effect could be reached by realizing outlet of EFCPO from so-called, pass bands of frequencies (PBF). PBF is located symmetrical on both parties from frequency of own oscillations of FA. Methods, algorithms of calculations and quantitative estimations are developed for EFCPO, Q and PBF in various modes of operation NPP with VVER-1000. Results of calculations allow specifying area of resonant interaction EFCPO with vibrations of FR, FA and a basket of reactor core. For practical realization of the received results it is offered to make corresponding additions to the design documentation and maintenance instructions of the equipment of the NPP with VVER-1000. The improvement of these documents

  3. Nuclear reactor coolant and cover gas system

    International Nuclear Information System (INIS)

    George, J.A.; Redding, A.H.; Tower, S.N.

    1976-01-01

    A core cooling system is disclosed for a nuclear reactor of the type utilizing a liquid coolant with a cover gas above free surfaces of the coolant. The disclosed system provides for a large inventory of reactor coolant and a balanced low pressure cover gas arrangement. A flow restricting device disposed within a reactor vessel achieves a pressure of the cover gas in the reactor vessel lower than the pressure of the reactor coolant in the vessel. The low gas pressure is maintained over all free surfaces of the coolant in the cooling system including a coolant reservoir tank. Reactor coolant stored in the reservoir tank allows for the large reactor coolant inventory provided by the invention

  4. Transforming 'Female' Programmes: Don’t Tell the Bride from International TV to Italian Digital Channels for Women

    NARCIS (Netherlands)

    Penati, Cecilia

    2016-01-01

    textabstractContemporary Italian digital channels explicitly targeting women (such as Real Time, Lei, Fox Life, La5, and La7d) represent a privileged observatory for some general trends in the international circulation of content and incorporation of foreign formats into national television (TV). In

  5. New insights on the voltage dependence of the KCa3.1 channel block by internal TBA.

    Science.gov (United States)

    Banderali, Umberto; Klein, Hélène; Garneau, Line; Simoes, Manuel; Parent, Lucie; Sauvé, Rémy

    2004-10-01

    We present in this work a structural model of the open IKCa (KCa3.1) channel derived by homology modeling from the MthK channel structure, and used this model to compute the transmembrane potential profile along the channel pore. This analysis showed that the selectivity filter and the region extending from the channel inner cavity to the internal medium should respectively account for 81% and 16% of the transmembrane potential difference. We found however that the voltage dependence of the IKCa block by the quaternary ammonium ion TBA applied internally is compatible with an apparent electrical distance delta of 0.49 +/- 0.02 (n = 6) for negative potentials. To reconcile this observation with the electrostatic potential profile predicted for the channel pore, we modeled the IKCa block by TBA assuming that the voltage dependence of the block is governed by both the difference in potential between the channel cavity and the internal medium, and the potential profile along the selectivity filter region through an effect on the filter ion occupancy states. The resulting model predicts that delta should be voltage dependent, being larger at negative than positive potentials. The model also indicates that raising the internal K+ concentration should decrease the value of delta measured at negative potentials independently of the external K+ concentration, whereas raising the external K+ concentration should minimally affect delta for concentrations >50 mM. All these predictions are born out by our current experimental results. Finally, we found that the substitutions V275C and V275A increased the voltage sensitivity of the TBA block, suggesting that TBA could move further into the pore, thus leading to stronger interactions between TBA and the ions in the selectivity filter. Globally, these results support a model whereby the voltage dependence of the TBA block in IKCa is mainly governed by the voltage dependence of the ion occupancy states of the selectivity filter.

  6. Coolant monitoring systems for PWR reactors

    International Nuclear Information System (INIS)

    Luzhnov, A.M.; Morozov, V.V.; Tsypin, S.G.

    1987-01-01

    The ways of improving information capacity of existing monitoring systems and the necessity of designing new ones for coolant monitoring are reviewed. A wide research program on development of coolant monitoring systems in PWR reactors is analyzed. The possible applications of in-core and out-of-core detectors for coolant monitoring are demonstrated

  7. Computational Fluid Dynamics Study of Channel Geometric Effect for Fischer-Tropsch Microchannel Reactor

    International Nuclear Information System (INIS)

    Na, Jonggeol; Jung, Ikhwan; Kshetrimayum, Krishnadash S.; Park, Seongho; Park, Chansaem; Han, Chonghun

    2014-01-01

    Driven by both environmental and economic reasons, the development of small to medium scale GTL(gas-to-liquid) process for offshore applications and for utilizing other stranded or associated gas has recently been studied increasingly. Microchannel GTL reactors have been preferred over the conventional GTL reactors for such applications, due to its compactness, and additional advantages of small heat and mass transfer distance desired for high heat transfer performance and reactor conversion. In this work, multi-microchannel reactor was simulated by using commercial CFD code, ANSYS FLUENT, to study the geometric effect of the microchannels on the heat transfer phenomena. A heat generation curve was first calculated by modeling a Fischer-Tropsch reaction in a single-microchannel reactor model using Matlab-ASPEN integration platform. The calculated heat generation curve was implemented to the CFD model. Four design variables based on the microchannel geometry namely coolant channel width, coolant channel height, coolant channel to process channel distance, and coolant channel to coolant channel distance, were selected for calculating three dependent variables namely, heat flux, maximum temperature of coolant channel, and maximum temperature of process channel. The simulation results were visualized to understand the effects of the design variables on the dependent variables. Heat flux and maximum temperature of cooling channel and process channel were found to be increasing when coolant channel width and height were decreased. Coolant channel to process channel distance was found to have no effect on the heat transfer phenomena. Finally, total heat flux was found to be increasing and maximum coolant channel temperature to be decreasing when coolant channel to coolant channel distance was decreased. Using the qualitative trend revealed from the present study, an appropriate process channel and coolant channel geometry along with the distance between the adjacent

  8. Avoiding Internal Capsule Stimulation With a New Eight-Channel Steering Deep Brain Stimulation Lead.

    Science.gov (United States)

    van Dijk, Kees J; Verhagen, Rens; Bour, Lo J; Heida, Ciska; Veltink, Peter H

    2017-10-15

    Novel deep brain stimulation (DBS) lead designs are currently entering the market, which are hypothesized to provide a way to steer the stimulation field away from neural populations responsible for side effects and towards populations responsible for beneficial effects. The objective of this study is to assess the performances of a new eight channel steering-DBS lead and compare this with a conventional cylindrical contact (CC) lead. The two leads were evaluated in a finite element electric field model combined with multicompartment neuron and axon models, representing the internal capsule (IC) fibers and subthalamic nucleus (STN) cells. We defined the optimal stimulation setting as the configuration that activated the highest percentage of STN cells, without activating any IC fibers. With this criterion, we compared monopolar stimulation using a single contact of the steering-DBS lead and CC lead, on three locations and four orientations of the lead. In addition, we performed a current steering test case by dividing the current over two contacts with the steering-DBS lead in its worst-case orientation. In most cases, the steering-DBS lead is able to stimulate a significantly higher percentage of STN cells compared to the CC lead using single contact stimulation or using a two contact current steering protocol when there is approximately a 1 mm displacement of the CC lead. The results also show that correct placement and orientation of the lead in the target remains an important aspect in achieving the optimal stimulation outcome. Currently, clinical trials are set up in Europe with a similar design as the steering-DBS lead. Our results illustrate the importance of the orientation of the new steering-DBS lead in avoiding side effects induced by stimulation of IC fibers. Therefore, in clinical trials sufficient attention should be paid to implanting the steering DBS-lead in the most effective orientation. © 2017 International Neuromodulation Society.

  9. Trace organics in AGR coolants

    International Nuclear Information System (INIS)

    Smith, R.; Green, L.O.; Johnson, P.A.V.

    1980-01-01

    Several analytical techniques have been employed in previous studies of the stable organic compounds arising from the radiolysis of methane/carbon monoxide/carbon dioxide coolants. The majority of this early information was collected from the Windscale AGR prototype. Analyses were also carried out on the liquors obtained from the WAGR humidryers. Three classes of compound were found in the liquors; aliphatic acids in the aqueous phase and methyl ketones and aromatic hydrocarbons in the oily phase. Acetic acid was found to be the predominant carboxylic acid. This paper outlines the major findings from a recent analytical survey of coolants taken over a wide range of dose rate, pressure, temperature and composition, from materials testing reactor facilities, WAGR and CAGR. (author)

  10. Multi-rod burst test under a loss-of coolant accident condition, (4)

    International Nuclear Information System (INIS)

    Otomo, Takashi; Hashimoto, Masao; Kawasaki, Satoru; Furuta, Teruo; Uetsuka, Hiroshi

    1983-06-01

    Multi-rod burst test of No.7808 bundle was performed in steam to estimate quantitative coolant flow channel restriction caused by the ballooning of zircaloy claddings in a fuel assembly during a LOCA transient in LWRs. The test was conducted under the condition that the initial internal pressure in each rod was 35kg/cm 2 (RT) and the heating rate was 9 0 C/s in steam with flow rate of 0.4g/cm 2 .min. The following results were obtained; (1) Maximum and burst pressures in rods were in the range 45 to 48kg/cm 2 and 41 to 45kg/cm 2 , respectively. The burst temperature of cladding were estimated to be 850 to 880 0 C. (2) Axial portions of tubes with greater than 34% strain were observed in the range 0 to 40mm in most rod. The mean length was 19mm in the bundle. (3) The degree of maximum increase in cross-sectional area is 54.2% in the bundle(7 x 7) and 66.9% in the internal rods(5 x 5). (4) Maximum channel area restriction was 40.5% in the bundle(7 x 7) and 51.4% in the internal rods(5 x 5). (author)

  11. Coolant clean-up and recycle systems

    International Nuclear Information System (INIS)

    Ito, Takao.

    1979-01-01

    Purpose: To increase the service life of mechanical seals in a shaft sealing device, eliminate leakages and improve the safety by providing a recycle pump for feeding coolants to a coolant clean-up device upon reactor shut-down and adapting the pump treat only low temperature and low pressure coolants. Constitution: The system is adapted to partially take out coolants from the pipeways of a recycling pump upon normal operation and feed them to a clean-up device. Upon reactor shut-down, the recycle pump is stopped and coolants are extracted by the recycle pump for shut-down into the clean-up device. Since the coolants are not fed to the clean-up device by the recycle pump during normal operation as conducted so far, high temperature and high pressure coolants are not directly fed to the recycle pump, thereby enabling to avoid mechanical problems in the pump. (Kamimura, M.)

  12. Transforming 'Female' Programmes: Don’t Tell the Bride from International TV to Italian Digital Channels for Women

    Directory of Open Access Journals (Sweden)

    Cecilia Penati

    2016-08-01

    Full Text Available Contemporary Italian digital channels explicitly targeting women (such as Real Time, Lei, Fox Life, La5, and La7d represent a privileged observatory for some general trends in the international circulation of content and incorporation of foreign formats into national television (TV. In fact, their schedules rely heavily on the genre of factual entertainment, which was first devised and used in international contexts, such as the UK and the US, and only in the second phase was imported into other national TV environments.After introducing this scenario and the main issues connected to the circulation of international content on these channels, the article will focus on the BBC docu-reality Don’t Tell the Bride. The show was extensively circulated (UK and US versions in the circuit of Italian women channels before a national version was produced by the pay-TV brand Lei. The article will reflect on how the national version Non ditelo alla sposa added further shades to the original meaning of the format. While the international version only allowed the Italian audience to engage with the programme through a romantic and escapist approach, the local adaptation inserted practical advice in the storytelling on how to arrange a low budget yet traditional wedding day, coupled with the sarcastic and ironic points of view of the characters involved.

  13. Slow coolant phaseout could worsen warming

    Science.gov (United States)

    Reese, April

    2018-03-01

    In the summer of 2016, temperatures in Phalodi, an old caravan town on a dry plain in northwestern India, reached a blistering 51°C—a record high during a heat wave that claimed more than 1600 lives across the country. Wider access to air conditioning (AC) could have prevented many deaths—but only 8% of India's 249 million households have AC. As the nation's economy booms, that figure could rise to 50% by 2050. And that presents a dilemma: As India expands access to a life-saving technology, it must comply with international mandates—the most recent imposed just last fall—to eliminate coolants that harm stratospheric ozone or warm the atmosphere.

  14. Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Thermo- and fluid-dynamic effects

    Energy Technology Data Exchange (ETDEWEB)

    Seeliger, André, E-mail: a.seeliger@hszg.de [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany); Alt, Sören; Kästner, Wolfgang; Renger, Stefan [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany); Kryk, Holger; Harm, Ulrich [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany)

    2016-08-15

    Highlights: • Borated coolant supports corrosion at zinc-coated installations in PWR after LOCA. • Dissolved zinc is injected into core by ECCS during sump recirculation phase. • Corrosion products can reach and settle at further downstream components. • Corrosion products can cause head losses at spacers and influence decay heat removal. • Preventive procedures were tested at semi-technical scale facilities. - Abstract: Within the framework of the German reactor safety research, generic experimental investigations were carried out aiming at thermal-hydraulic consequences of physicochemical mechanisms, caused by dissolution of zinc in boric acid during corrosion processes at hot-dip galvanized surfaces of containment internals at lower coolant temperatures and the subsequent precipitation of solid zinc borates in PWR core regions of higher temperature. This constellation can occur during sump recirculation operation of ECCS after LOCA. Hot-dip galvanized compounds, which are installed inside a PWR containment, may act as zinc sources. Getting in contact with boric acid coolant, zinc at their surfaces is released into coolant in form of ions due to corrosion processes. As a long-term behavior resp. over a time period of several days, metal layers of zinc and zinc alloys can dissolve extensively. First fundamental studies at laboratory scale were done at the Helmholtz-Zentrum Dresden-Rossendorf (HZDR). Their experimental results were picked up for the definition of boundary conditions for experiments at semi-technical scale at the Hochschule Zittau/Görlitz (HSZG). Electrical heating rods with zircaloy cladding tubes have been used as fuel rod simulators. As near-plant core components, a 3 × 3 configuration of heating rods (HRC) and a shortened, partially heatable PWR fuel assembly dummy were applied into cooling circuits. The HRC module includes segments of spacers for a suitable representation of a heating channel geometry. Formations of different solid

  15. Method for producing components with internal architectures, such as micro-channel reactors, via diffusion bonding sheets

    Science.gov (United States)

    Alman, David E [Corvallis, OR; Wilson, Rick D [Corvallis, OR; Davis, Daniel L [Albany, OR

    2011-03-08

    This invention relates to a method for producing components with internal architectures, and more particularly, this invention relates to a method for producing structures with microchannels via the use of diffusion bonding of stacked laminates. Specifically, the method involves weakly bonding a stack of laminates forming internal voids and channels with a first generally low uniaxial pressure and first temperature such that bonding at least between the asperites of opposing laminates occurs and pores are isolated in interfacial contact areas, followed by a second generally higher isostatic pressure and second temperature for final bonding. The method thereby allows fabrication of micro-channel devices such as heat exchangers, recuperators, heat-pumps, chemical separators, chemical reactors, fuel processing units, and combustors without limitation on the fin aspect ratio.

  16. Experimental investigations of flow distribution in coolant system of Helium-Cooled-Pebble-Bed Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Ilić, M.; Schlindwein, G., E-mail: georg.schlindwein@kit.edu; Meyder, R.; Kuhn, T.; Albrecht, O.; Zinn, K.

    2016-02-15

    Highlights: • Experimental investigations of flow distribution in HCPB TBM are presented. • Flow rates in channels close to the first wall are lower than nominal ones. • Flow distribution in central chambers of manifold 2 is close to the nominal one. • Flow distribution in the whole manifold 3 agrees well with the nominal one. - Abstract: This paper deals with investigations of flow distribution in the coolant system of the Helium-Cooled-Pebble-Bed Test Blanket Module (HCPB TBM) for ITER. The investigations have been performed by manufacturing and testing of an experimental facility named GRICAMAN. The facility involves the upper poloidal half of HCPB TBM bounded at outlets of the first wall channels, at outlet of by-pass pipe and at outlets of cooling channels in breeding units. In this way, the focus is placed on the flow distribution in two mid manifolds of the 4-manifold system: (i) manifold 2 to which outlets of the first wall channels and inlet of by-pass pipe are attached and (ii) manifold 3 which supplies channels in breeding units with helium coolant. These two manifolds are connected with cooling channels in vertical/horizontal grids and caps. The experimental facility has been built keeping the internal structure of manifold 2 and manifold 3 exactly as designed in HCPB TBM. The cooling channels in stiffening grids, caps and breeding units are substituted by so-called equivalent channels which provide the same hydraulic resistance and inlet/outlet conditions, but have significantly simpler geometry than the real channels. Using the conditions of flow similarity, the air pressurized at 0.3 MPa and at ambient temperature has been used as working fluid instead of HCPB TBM helium coolant at 8 MPa and an average temperature of 370 °C. The flow distribution has been determined by flow rate measurements at each of 28 equivalent channels, while the pressure distribution has been obtained measuring differential pressure at more than 250 positions. The

  17. An experimental investigation of heat transfer from a reactor fuel channel to surrounding water

    International Nuclear Information System (INIS)

    Gillespie, G.E.

    An important feature of the CANDU-PHW reactor is that each fuel channel is surrounded by cool heavy-water moderator that can act as a sink for heat generated in the fuel if other means of heat removal were to fail. During postulated loss-of-coolant accidents there are two scenarios in which the primary cooling system may not prevent fuel-channel overheating. These situations arise when: (1) for a particular break size and location, called the critical break, the coolant flow through a portion of the reactor core stagnates before the emergency coolant injection system restores circulation, or, (2) the emergency coolant injection system fails to operate. In either case, the heat generated in the fuel is transferred mainly by radiation to the pressure tube and calandria tube, and then by boiling heat transfer to the moderator. This paper describes a simple one-dimensional model developed to analyse the thermal behaviour of a fuel channel when the internal pressure is high. Also described is a series of experiments in which the pressure-tube segment is pressurized and heated at a constant rate until it contacts a surrounding calandria-tube segment. Predictions of the one-dimensional model are compared with the experimental results

  18. Dimensional accuracy of internal cooling channel made by selective laser melting (SLM And direct metal laser sintering (DMLS processes in fabrication of internally cooled cutting tools

    Directory of Open Access Journals (Sweden)

    Ghani S. A. C.

    2017-01-01

    Full Text Available Selective laser melting(SLM and direct metal laser sintering(DMLS are preferred additive manufacturing processes in producing complex physical products directly from CAD computer data, nowadays. The advancement of additive manufacturing promotes the design of internally cooled cutting tool for effectively used in removing generated heat in metal machining. Despite the utilisation of SLM and DMLS in a fabrication of internally cooled cutting tool, the level of accuracy of the parts produced remains uncertain. This paper aims at comparing the dimensional accuracy of SLM and DMLS in machining internally cooled cutting tool with a special focus on geometrical dimensions such as hole diameter. The surface roughness produced by the two processes are measured with contact perthometer. To achieve the objectives, geometrical dimensions of identical tool holders for internally cooled cutting tools fabricated by SLM and DMLS have been determined by using digital vernier calliper and various magnification of a portable microscope. In the current study, comparing internally cooled cutting tools made of SLM and DMLS showed that generally the higher degree of accuracy could be obtained with DMLS process. However, the observed differences in surface roughness between SLM and DMLS in this study were not significant. The most obvious finding to emerge from this study is that the additive manufacturing processes selected for fabricating the tool holders for internally cooled cutting tool in this research are capable of producing the desired internal channel shape of internally cooled cutting tool.

  19. Coolant inlet device for nuclear reactors

    International Nuclear Information System (INIS)

    Ando, Hiroshi; Abe, Yasuhiro; Iwabuchi, Toshihiko; Yamamoto, Kenji.

    1969-01-01

    Herein disclosed is a coolant inlet device for liquid-metal cooled reactors which employs a coolant distributor serving also as a supporting means for the reactor core. The distributor is mounted within the reactor vessel so as to slide horizontally on supporting lugs, and is further slidably connected via a junction pipe to a coolant inlet conduit protruding through the floor of the vessel. The distributor is adapted to uniformly disperse the highly pressured coolant over the reactor core so as to reduce the stresses sustained by the reactor vessel as well as the supporting lugs. Moreover, the slidable nature of the distributor allows thermal shock and excessive coolant pressures to be prevented or alleviated, factors which posed major difficulties in conventional coolant inlet devices. (Owens, K. J.)

  20. Assessment of Loss-of-Coolant Effect on Pressurized Heavy Water Reactors

    International Nuclear Information System (INIS)

    Kim, Won Young; Park, Joo Hwan; Kim, Bong Ghi

    2009-01-01

    A CANDU reactor is a heavy-water-moderated, natural uranium fuelled reactor with a pressure tube. The reactor contains a horizontal cylindrical vessel (calandria) and each pressure tube is isolated from the heavy-water moderator in a calandria. This allows the moderator system to be operated of a high-pressure and of a high-temperature coolant in pressure tube. This causes the pressurized liquid coolant in the channel to void and therefore give rise to a reactivity transient in the event of a break or fault in the coolant circuit. In particular, all CANDU reactors are well known to have a positive void reactivity coefficient and thus this phenomenon may lead to a positive feedback, which can cause a large power pulse. We assess the loss-of-coolant effect by coolant void reactivity versus fuel burnup, four factor parameters for fresh fuel and equilibrium fuel, reactivity change due to the change of coolant density and reactivity change in the case of half- and full-core coolant

  1. Organic coolant for ARIES-III

    International Nuclear Information System (INIS)

    Sze, D.K.; Sviatoslavsky, I.; Sawan, M.; Gierszewski, P.; Hollies, R.; Sharafat, S.; Herring, S.

    1991-04-01

    ARIES-III is a D-He 3 reactor design study. It is found that the organic coolant is well suited for the D-He 3 reactor. This paper discusses the unique features of the D-He 3 reactor, and the reason that the organic coolant is compatible with those features. The problems associated with the organic coolant are also discussed. 8 refs., 2 figs., 6 tabs

  2. Physical properties of organic coolants

    International Nuclear Information System (INIS)

    Debbage, A.G.; Garton, D.A.; Kinneir, J.H.

    1963-03-01

    Density, viscosity, specific heat, vapour pressure and calorific value were measured within the temperature range 100 - 400 deg C for mixtures of Santowax R with pyrolytic high boiler and Santowax R with O.M.R.E. radiolytic high boiler; in addition measurements were made on Santowax OM, X-7 standard, X-7 loop coolant and O.M.R.E. coolant supplied by Atomic Energy of Canada Ltd. The accuracy of the measurements made were density (± 1/4%), viscosity (± 2%), specific heat (± 2%), vapour pressure (± 2%) and calorific value (± 1/2%). Thermal conductivity was calculated from an improved form of the Smiths equation with an accuracy within ± 6%. Equations fitted to the vapour pressure results were used to provide data outside the experimental range for burnout correlation purposes. The general effect of high boiler content on the specific heat and calorific values was small. The differences in physical property values for corresponding values of either pyrolytic or radiolytic high boiler were small for density (0.3%) and specific heat (2%), but quite large for viscosity (70%) with the pyrolytic high boiler mixture giving the higher value. The chemical analysis of all materials was based on gas chromatography and the relationship between this and an earlier distillation method established. (author)

  3. Detailed channel thermal-hydraulic calculation of nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    Zhukov, A.V.; Sorokin, A.P.; Ushakov, P.A.; Yur'ev, Yu.S.

    1981-01-01

    The system of equations of mass balance, quantity of motion and energy used in calculation of nuclear reactor fuel assemblies is obtained. The equation system is obtained on the base of integral equations of hydrodynamics interaction in assemblies of smooth fuel elements and fuel elements with wire packing. The calculation results of coolant heating distributions by the fast reactor assembly channels are presented. The analysis of the results obtained shows that interchannel exchange essentially uniforms the coolant heating distribution in the peripheral range of the assembly but it does not remove non-uniformity caused by power distribution non-uniformity in the cross section. Geometry of the peripheral assembly range plays an essential role in the heating distribution. Change of the calculation gap between the peripheral fuel elements and assembly shells can result either in superheating or in subcooling in the peripheral channels relatively to joint internal channels of the assembly. Heat supply to the coolant passing through interassembly gaps decreases temperature in the assembly periphery and results in the increase of temperature non-uniformity by the perimeter of peripheral fuel elements. It is concluded that the applied method of the channel-by-channel calculation is ef-- fective in thermal-physical calculation of nuclear reactor fuel assemblies and it permits to solve a wide range of problems [ru

  4. Cleaning of aluminum after machining with coolants

    International Nuclear Information System (INIS)

    Roop, B.

    1992-01-01

    An x-ray photoemission spectroscopic study was undertaken to compare the cleaning of the Advanced Photon Source (APS) aluminum extrusion storage ring vacuum chambers after machining with and without water soluble coolants. While there was significant contamination left by the coolants, the cleaning process was capable of removing the residue. The variation of the surface and near surface composition of samples machined either dry or with coolants was negligible after cleaning. The use of such coolants in the machining process is therefore recommended

  5. Coolant clean-up system in the primary coolant circuit for nuclear reactor

    International Nuclear Information System (INIS)

    Saito, Michio.

    1981-01-01

    Purpose: To maintain the quality of coolants at a prescribed level by distillating coolants in the primary coolant circuit for a BWR type reactor to remove impurities therefrom, taking out the condensates from the top of the distillation column and extracting impurities in a concentrated state from the bottom. Constitution: Coolant water for cooling the core is recycled by a recycling pump by way of a recycling pipeway in a reactor. The coolants extracted from an extraction pipeway connected to the recycling pipeway are fed into a distillation column, where distillation is taken place. Impurities in the coolants, that is, in-core corrosion products, fission products generated in the reactor core, etc. are separated by the distillation, concentrated and solidified in the bottom of the distillation column. While on the other hand, condensates removed with the impurities, that is, coolants cleaned-up are recycled to the coolant water for cooling the reactor core. (Moriyama, K.)

  6. Design and fabrication of magnetic coolant filter

    Science.gov (United States)

    Prashanth, B. N.

    2017-07-01

    Now a day's use of coolants in industry has become dominant because of high production demands. Coolants not only help in speeding up the production but also provide many advantages in the metal working operation. As the consumption of coolants is very high a system is badly in need, so as to recirculate the used coolant. Also the amount of hazardous waste generated by industrial plants has become an increasingly costly problem for the manufactures and an additional stress on the environment. Since the purchase and disposal of the spent cutting fluids is becoming increasingly expensive, fluid recycling is a viable option for minimizing the cost. Separation of metallic chips from the coolants by using magnetic coolant separation has proven a good management and maintenance of the cutting fluid. By removing the metallic chips, the coolant life is greatly extended, increases the machining quality and reduces downtime. Above being the case, a magnetic coolant filter is developed which utilizes high energy permanent magnets to develop a dense magnetic field along a narrow flow path into which the contaminated coolant is directed. The ferromagnetic particles captured and aligned by the dense magnetic field, from the efficient filter medium. This enables the unit to remove ferromagnetic particles from the coolant. Magnetic coolant filters use the principle of magnetic separation to purify the used coolant. The developed magnetic coolant separation has the capability of purifying 40 litres per minute of coolant with the size of the contaminants ranging from 1 µm to 30 µm. The filter will be helpful in saving the production cost as the cost associated with the proposed design is well justified by the cost savings in production. The magnetic field produced by permanent magnets will be throughout the area underneath the reservoir. This produces magnetic field 30mm above the coolant reservoir. Very fine particles are arrested without slip. The magnetic material used will not

  7. A review and integration of empirical evidence on internal economy in marketing channel relationships

    DEFF Research Database (Denmark)

    Jensen, Nils Bøgelund; Skytte, Hans

    Papiret gennemgår empirisk forskning omkring intern økonomi i distributionskanaler. Ved intern økonomi forstås graden af vertikal integration og beslutningsprocesser i distributionskanaler. På basis af gennemgangen diskuteres konklusioner og resultaterne integreres....

  8. Mathematical model of the reactor coolant pump

    International Nuclear Information System (INIS)

    Kozuh, M.

    1989-01-01

    The mathematical model of reactor coolant pump is described in this paper. It is based on correlations for centrifugal reactor coolant pumps. This code is one of the elements needed for the simulation of the whole NPP primary system. In subroutine developed according to this model we tried in every possible detail to incorporate plant specific data for Krsko NPP. (author)

  9. Labelling Of Coolant Flow Anomaly Using Fractal Structure

    International Nuclear Information System (INIS)

    Djainal, Djen Djen

    1996-01-01

    This research deals with the instrumentation of the detection and characterization of vertical two-phase flow coolant. This type of work is particularly intended to find alternative method for the detection and identification of noise in vertical two-phase flow in a nuclear reactor environment. Various new methods have been introduced in the past few years, an attempt to developed an objective indicator off low patterns. One of new method is Fractal analysis which can complement conventional methods in the description of highly irregular fluctuations. In the present work, Fractal analysis was applied to analyze simulated boiling coolant signal. This simulated signals were built by sum random elements in small subchannels of the coolant channel. Two modes are defined and both are characterized by their void fractions. In the case of uni modal -PDF signals, the difference between these modes is relatively small. On other hand, bimodal -PDF signals have relative large range. In this research, Fractal dimension can indicate the characters of that signals simulation

  10. LIMBO computer code for analyzing coolant-voiding dynamics in LMFBR safety tests

    International Nuclear Information System (INIS)

    Bordner, G.L.

    1979-10-01

    The LIMBO (liquid metal boiling) code for the analysis of two-phase flow phenomena in an LMFBR reactor coolant channel is presented. The code uses a nonequilibrium, annular, two-phase flow model, which allows for slip between the phases. Furthermore, the model is intended to be valid for both quasi-steady boiling and rapid coolant voiding of the channel. The code was developed primarily for the prediction of, and the posttest analysis of, coolant-voiding behavior in the SLSF P-series in-pile safety test experiments. The program was conceived to be simple, efficient, and easy to use. It is particularly suited for parametric studies requiring many computer runs and for the evaluation of the effects of model or correlation changes that require modification of the computer program. The LIMBO code, of course, lacks the sophistication and model detail of the reactor safety codes, such as SAS, and is therefore intended to compliment these safety codes

  11. Applied model of through-wall crack of coolant vessels of WWER-type reactors

    International Nuclear Information System (INIS)

    Petrosyan, V.; Hovakimyan, T.; Vardanyan, M.; Khachatryan, A.; Minasyan, K.

    2010-01-01

    We propose an applied-model of Through-Wall Crack (TWC) for WWER-type units primary vessels. The model allows to simulate the main morphological parameters of real TWC, i.e. length, area of inlet and outlet openings, channel depth and small and large size unevenness of the crack surface. The model can be used for developing and improving the coolant-leak detectors for the primary circuit vessels of WWER-units. Also, it can be used for research of the coolant two-phase leakage phenomenon through narrow cracks/channels and thermo-physical processes in heat-insulation layer of the Main Coolant Piping (MCP) during the leak

  12. Fabrication and Characteristics of Sintered Cutting Stainless Steel Fiber Felt with Internal Channels and an Al2O3 Coating

    Directory of Open Access Journals (Sweden)

    Shufeng Huang

    2018-03-01

    Full Text Available A novel sintered cutting stainless steel fiber felt with internal channels (SCSSFFC composed of a stainless-steel fiber skeleton, three-dimensional interconnected porous structure and multiple circular microchannels is developed. SCSSFFC has a jagged and rough surface morphology and possesses a high specific surface area, which is approximately 2.4 times larger than that of the sintered bundle-drawing stainless steel fiber felt with internal channels (SBDSSFFC and is expected to enhance adhesive strength. The sol-gel and wet impregnation methods are adopted to prepare SCSSFFC with an Al2O3 coating (SCSSFFC/Al2O3. The adhesive strength of SCSSFFC/Al2O3 is investigated using ultrasonic vibration and thermal shock tests. The experimental results indicate that the weight loss rate of the Al2O3 coating has a 4.2% and 8.42% reduction compared with those of SBDSSFFCs based on ultrasonic vibration and thermal shock tests. In addition, the permeability of SCSSFFC/Al2O3 is investigated based on forced liquid flow tests. The experimental results show that the permeability and inertial coefficients of SCSSFFC/Al2O3 are mainly affected by the coating rate, porosity and open ratio; however, the internal microchannel diameter has little influence. It is also found that SCSSFFC/Al2O3 yields superior permeability, as well as inertial coefficients compared with those of other porous materials reported in the literature.

  13. Organic coolant in Winnipeg riverbed sediments

    International Nuclear Information System (INIS)

    Guthrie, J.E.; Acres, O.E.

    1979-03-01

    Between January and May 1977 a prolonged leak of organic coolant occurred from the Whiteshell Nuclear Research Establishment's nuclear reactor, and a minimum of 1450 kg of coolant entered the Winnipeg River and was deposited on the riverbed. The level of radioactivity associated with this coolant was low, contributing less than 0.2 μGy (0.02 mrad) a year to the natural background gamma radiation field from the riverbed. The concentration of coolant in the water samples never exceeded 0.02 mg/L, the lower limit of detection. The mortality of crayfish, held in cages where the riverbed was covered with the largest deposits of coolant, was not significantly different from that in the control cages upstream of the outfall. No evidence of fish kill was found. (author)

  14. Primary coolant circuits in FBR type reactors

    International Nuclear Information System (INIS)

    Kutani, Masushiro.

    1985-01-01

    Purpose: To eliminate the requirement of a pump for the forcive circulation of primary coolants and avoid the manufacturing difficulty of equipments. Constitution: In primary coolant circuits of an LMFBR type reactor having a recycling path forming a closed loop between a reactor core and a heat exchanger, coolants recycled through the recycling path are made of a magnetic fluid comprising liquid sodium incorporated with fine magnetic powder, and an electromagnet is disposed to the downstream of the heat exchanger. In the above-mentioned structure, since the magnetic fluid as the primary coolants losses its magnetic property when heated in the reactor core but recovers the property at a lower temperature after the completion of the heat exchange, the magnetic fluid can forcively be flown through the recycling path under the effect of the electromagnet disposed to the down stream of the heat exchanger to thereby forcively recycle the primary coolants. (Kawakami, Y.)

  15. Neutronic Analysis on Coolant Options in a Hybrid Reactor System for High Level Waste Transmutation

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Seong Hee; Kim, Myung Hyun [Kyung Hee University, Seoul (Korea, Republic of)

    2014-10-15

    A fusion-fission hybrid reactor (FFHR) which is a combination of plasma fusion tokamak as a fast neutron source and a fission reactor as of fusion blanket is another potential candidate. In FFHR, fusion plasma machine can supply high neutron-rich and energetic 14.1MeV (D, T) neutrons compared to other options. Therefore it has better capability in HLW incineration. While, it has lower requirements compared to pure fusion. Much smaller-sized tokamak can be achievable in a near term because it needs relatively low plasma condition. FFHR has also higher safety potential than fast reactors just as ADSR because it is subcritical reactor system. FFHR proposed up to this time has many design concepts depending on the design purpose. FFHR may also satisfy many design requirement such as energy multiplication, tritium production, radiation shielding for magnets, fissile breeding for self-sustain ability also waste transmutation. Many types of fuel compositions and coolant options have been studied. Effect of choices for fuel and coolant was studied for the transmutation purpose FFHR by our team. In this study LiPb coolant was better than pure Li coolant both for neutron multiplication and tritium breeding. However, performance of waste transmutation was reduced with increased neutron absorption at coolant caused by tritium breeding. Also, LiPb as metal coolant has a problem of massive MHD pressure drop in coolant channels. Therefore, in a previous study, waste transmutation performance was evaluated with light water coolant option which may be a realistic choice. In this study, a neutronic analysis was done for the various coolant options with a detailed computation. One of solutions suggested is to use the pressure tubes inside of first wall and second wall In this work, performance of radioactive waste transmutation was compared with various coolant options. On the whole, keff increases with all coolants except for FLiBe, therefore required fusion power is decreased. In

  16. A system for the discharge of gas bubbles from the coolant flow of a nuclear reactor cooled by forced circulation

    International Nuclear Information System (INIS)

    Markfort, D.; Kaiser, A.; Dohmen, A.

    1975-01-01

    In a reactor cooled by forced circulation the gas bubbles carried along with the coolant flow are separated before entering the reactor core or forced away into the external zones. For this purpose the coolant is radially guided into a plenum below the core and deflected to a tangential direction by means of flow guide elements. The flow runs spirally downwards. On the bubbles, during their dwell time in this channel, the buoyant force and a force towards the axis of symmetry of the tank are exerted. The major part of the coolant is directed into a radial direction by means of a guiding apparatus in the lower section of the channel and guided through a chimney in the plenum to the center of the reactor core. This inner chimney is enclosed by an outer chimney for the core edge zones through which coolant with a small share of bubbles is taken away. (RW) [de

  17. Flow boiling test of GDP replacement coolants

    International Nuclear Information System (INIS)

    Park, S.H.

    1995-01-01

    The tests were part of the CFC replacement program to identify and test alternate coolants to replace CFC-114 being used in the uranium enrichment plants at Paducah and Portsmouth. The coolants tested, C 4 F 10 and C 4 F 8 , were selected based on their compatibility with the uranium hexafluoride process gas and how well the boiling temperature and vapor pressure matched that of CFC-114. However, the heat of vaporization of both coolants is lower than that of CFC-114 requiring larger coolant mass flow than CFC-114 to remove the same amount of heat. The vapor pressure of these coolants is higher than CFC-114 within the cascade operational range, and each coolant can be used as a replacement coolant with some limitation at 3,300 hp operation. The results of the CFC-114/C 4 F 10 mixture tests show boiling heat transfer coefficient degraded to a minimum value with about 25% C 4 F 10 weight mixture in CFC-114 and the degree of degradation is about 20% from that of CFC-114 boiling heat transfer coefficient. This report consists of the final reports from Cudo Technologies, Ltd

  18. Coolant cleanup method in a nuclear reactor

    International Nuclear Information System (INIS)

    Kubota, Masayoshi; Nishimura, Shigeoki; Takahashi, Sankichi; Izumi, Kenkichi; Motojima, Kenji.

    1983-01-01

    Purpose : To effectively adsorb to remove low molecular weight organic substances from iron exchange resins for use in the removal of various radioactive nucleides contained in reactor coolants. Method : Reactor coolants are recycled by a main recyling pump in a nuclear reactor and a portion of the coolants is cooled and, thereafter, purified in a coolant desalter. While on the other hand, high pressure steams generated from the reactor are passed through a turbine, cooled in a condensator, eliminated with claddings or the likes by the passage through a filtration desalter using powderous ion exchange resins and then further passed through a desalter (filled with granular ion exchange resins). For instance, an adsorption and removing device for organic substances (resulted through the decomposition of ion exchange resins) precoated with activated carbon powder or filled with granular activated carbon is disposed at the downstream for each of the desalters. In this way, the organic substances in the coolants are eliminated to prevent the reduction in the desalting performance of the ion exchange resins caused by the formation of complexes between organic substances and cobalt in the coolants, etc. In this way, the coolant cleanup performance is increased and the amount of wasted ion exchange resins can be decreased. (Horiuchi, T.)

  19. Exhaust temperature analysis of four stroke diesel engine by using MWCNT/Water nanofluids as coolant

    Science.gov (United States)

    Muruganandam, M.; Mukesh Kumar, P. C.

    2017-10-01

    There has been a continuous improvement in designing of cooling system and in quality of internal combustion engine coolants. The liquid engine coolant used in early days faced many difficulties such as low boiling, freezing points and inherently poor thermal conductivity. Moreover, the conventional coolants have reached their limitations of heat dissipating capacity. New heat transfer fluids have been developed and named as nanofluids to try to replace traditional coolants. Moreover, many works are going on the application of nanofluids to avail the benefits of them. In this experimental investigation, 0.1, 0.3 and 0.5% volume concentrations of multi walled carbon nanotube (MWCNT)/water nanofluids have been prepared by two step method with surfactant and is used as a coolant in four stroke single cylinder diesel engine to assess the exhaust temperature of the engine. The nanofluid prepared is characterized with scanning electron microscope (SEM) to confirm uniform dispersion and stability of nanotube with zeta potential analyzer. Experimental tests are performed by various mass flow rate such as 270 300 330 LPH (litre per hour) of coolant nanofluids and by changing the load in the range of 0 to 2000 W and by keeping the engine speed constant. It is found that the exhaust temperature decreases by 10-20% when compared to water as coolant at the same condition.

  20. Fast measurements of the in-core coolant velocity in a BWR by neutron noise analysis

    International Nuclear Information System (INIS)

    Hagen, T.H.J.J. van der; Hoogenboom, J.E.

    1988-01-01

    A method to determine in-core coolant velocities from neutron noise within short time intervals has been developed. The accuracy of the method was determined by using a simulation set-up and by using signals of a twin self-powered neutron detector installed in the core of the Dodewaard BWR in the Netherlands. In-core coolant velocities can be estimated within 2.5 s with a standard deviation (due to statistics) less than 2.1%. The method is suitable for velocity monitoring as is shown by the application to a stepwise velocity change of the coolant in a model of a coolant channel of a BWR. The presented technique was applied to determine the variations of the coolant velocity in the Dodewaard core during normal operation and during pressure steps. Only minor variations of the coolant velocity were detected during normal reactor conditions. An increase of those variations with pressure lowering - indicating a lower thermal hydraulic stability - could be detected. A clear velocity response to pressure steps could be determined which was also reflected in the cross-spectrum of the velocity with the vessel pressure and with the in-core neutron flux. (author)

  1. Nonlinear dynamic analysis of nuclear reactor primary coolant systems

    International Nuclear Information System (INIS)

    Saffell, B.F. Jr.; Macek, R.W.; Thompson, T.R.; Lippert, R.F.

    1979-01-01

    The ADINA computer code is utilized to perform mechanical response analysis of pressurized reactor primary coolant systems subjected to postulated loss-of-coolant accident (LOCA) loadings. Specifically, three plant analyses are performed utilizing the geometric and material nonlinear analysis capabilities of ADINA. Each reactor system finite element model represents the reactor vessel and internals, piping, major components, and component supports in a single coupled model. Material and geometric nonlinear capabilities of the beam and truss elements are employed in the formulation of each finite element model. Loadings applied to each plant for LOCA dynamic analysis include steady-state pressure, dead weight, strain energy release, transient piping hydraulic forces, and reactor vessel cavity pressurization. Representative results are presented with some suggestions for consideration in future ADINA code development

  2. TACT1- TRANSIENT THERMAL ANALYSIS OF A COOLED TURBINE BLADE OR VANE EQUIPPED WITH A COOLANT INSERT

    Science.gov (United States)

    Gaugler, R. E.

    1994-01-01

    As turbine-engine core operating conditions become more severe, designers must develop more effective means of cooling blades and vanes. In order to design reliable, cooled turbine blades, advanced transient thermal calculation techniques are required. The TACT1 computer program was developed to perform transient and steady-state heat-transfer and coolant-flow analyses for cooled blades, given the outside hot-gas boundary condition, the coolant inlet conditions, the geometry of the blade shell, and the cooling configuration. TACT1 can analyze turbine blades, or vanes, equipped with a central coolant-plenum insert from which coolant-air impinges on the inner surface of the blade shell. Coolant-side heat-transfer coefficients are calculated with the heat transfer mode at each station being user specified as either impingement with crossflow, forced convection channel flow, or forced convection over pin fins. A limited capability to handle film cooling is also available in the program. The TACT1 program solves for the blade temperature distribution using a transient energy equation for each node. The nodal energy balances are linearized, one-dimensional, heat-conduction equations which are applied at the wall-outer-surface node, at the junction of the cladding and the metal node, and at the wall-inner-surface node. At the mid-metal node a linear, three-dimensional, heat-conduction equation is used. Similarly, the coolant pressure distribution is determined by solving the set of transfer momentum equations for the one-dimensional flow between adjacent fluid nodes. In the coolant channel, energy and momentum equations for one-dimensional compressible flow, including friction and heat transfer, are used for the elemental channel length between two coolant nodes. The TACT1 program first obtains a steady-state solution using iterative calculations to obtain convergence of stable temperatures, pressures, coolant-flow split, and overall coolant mass balance. Transient

  3. Experimental distribution of coolant in the IPR-R1 Triga nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z., E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores; Palma, Daniel A.P., E-mail: dapalma@cnen.gov.b [Comissao Nacional de Energia Nuclear (CNEN/RJ), Rio de Janeiro, RJ (Brazil); Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Reis, Patricia A.L., E-mail: claubia@nuclear.ufmg.b, E-mail: dora@nuclear.ufmg.b [Universidade Federal de Minas Gerais (DEN/UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    2011-07-01

    The IPR-R1 is a typical TRIGA Mark I light-water and open pool type reactor. The core has an annular configuration of six rings and is cooled by natural circulation. The core coolant channels extend from the bottom grid plate to the top grid plate. The cooling water flows through the holes in the bottom grid plate, passes through the lower unheated region of the element, flows upwards through the active region, passes through the upper unheated region, and finally leaves the channel through the differential area between a triangular spacer block on the top of the fuel element and a round hole in the grid. Direct measurement of the flow rate in a coolant channel is difficult because of the bulky size and low accuracy of flow meters. The flow rate through the channel may be determined indirectly from the heat balance across the channel using measurements of the water inlet and outlet temperatures. This paper presents the experiments performed in the IPR-R1 reactor to monitoring some thermo-hydraulic parameters in the core coolant channels, such as: the radial and axial temperature profile, temperature, velocity, mass flow rate, mass flux and Reynolds's number. Some results were compared with theoretical predictions, as it was expected the variables follow the power distribution (or neutron flux) in the core. (author)

  4. Experimental distribution of coolant in the IPR-R1 Triga nuclear reactor core

    International Nuclear Information System (INIS)

    Mesquita, Amir Z.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Reis, Patricia A.L.

    2011-01-01

    The IPR-R1 is a typical TRIGA Mark I light-water and open pool type reactor. The core has an annular configuration of six rings and is cooled by natural circulation. The core coolant channels extend from the bottom grid plate to the top grid plate. The cooling water flows through the holes in the bottom grid plate, passes through the lower unheated region of the element, flows upwards through the active region, passes through the upper unheated region, and finally leaves the channel through the differential area between a triangular spacer block on the top of the fuel element and a round hole in the grid. Direct measurement of the flow rate in a coolant channel is difficult because of the bulky size and low accuracy of flow meters. The flow rate through the channel may be determined indirectly from the heat balance across the channel using measurements of the water inlet and outlet temperatures. This paper presents the experiments performed in the IPR-R1 reactor to monitoring some thermo-hydraulic parameters in the core coolant channels, such as: the radial and axial temperature profile, temperature, velocity, mass flow rate, mass flux and Reynolds's number. Some results were compared with theoretical predictions, as it was expected the variables follow the power distribution (or neutron flux) in the core. (author)

  5. Continuous surveillance of reactor coolant circuit integrity

    International Nuclear Information System (INIS)

    1986-01-01

    Continuous surveillance is important to assuring the integrity of a reactor coolant circuit. It can give pre-warning of structural degradation and indicate where off-line inspection should be focussed. These proceedings describe the state of development of several techniques which may be used. These involve measuring structural vibration, core neutron noise, acoustic emission from cracks, coolant leakage, or operating parameters such as coolant temperature and pressure. Twenty three papers have been abstracted and indexed separately for inclusion in the data base

  6. Determination of temperature distributions in fast reactor core coolants

    International Nuclear Information System (INIS)

    Tillman, M.

    1975-04-01

    An analytical method of determination of a temperature distribution in the coolant medium in a fuel assembly of a liquid-metal-fast-breeder-reactor (LMFBR) is presented. The temperature field obtained is applied for a constant velocity (slug flow) fluid flowing, parallel to the fuel pins of a square and hexagonal array assembly. The coolant subchannels contain irregular boundaries. The geometry of the channel due to the rod adjacent to the wall (edge rod) differs from the geometry of the other channels. The governing energy equation is solved analytically, assuming series solutions for the Poisson and diffusion equations, and the total solution is superposed by the two. The boundary conditions are specified by symmetry considerations, assembly wall insulation and a continuity of the temperature field and heat fluxes. The initial condition is arbitrary. The method satisfies the boundary conditions on the irregular boundaries and the initial condition by a least squares technique. Computed results are presented for various geometrical forms, with ratio of rod pitch-to-diameter typical for LMFBR cores. These results are applicable for various fast-reactors, and thus the influence of the transient solution (which solves the diffusion equation) on the total depends on the core parameters. (author)

  7. Condition monitoring of main coolant pumps, Dhruva

    International Nuclear Information System (INIS)

    Prasad, V.; Satheesh, C.; Acharya, V.N.; Tikku, A.C.; Mishra, S.K.

    2002-01-01

    Full text: Dhruva is a 100 MW research reactor with natural uranium fuel, heavy water as moderator and primary coolant. Three Centrifugal pumps circulate the primary coolant across the core and the heat exchangers. Each pump is coupled to a flywheel (FW) assembly in order to meet operational safety requirements. All the 3 main coolant pump (MCP) sets are required to operate during operation of the reactor. The pump-sets are in operation since the year 1984 and have logged more than 1,00,000 hrs. Frequent breakdowns of its FW bearings were experienced during initial years of operation. Condition monitoring of these pumps, largely on vibration based parameters, was initiated on regular basis. Break-downs of main coolant pumps reduced considerably due to the fair accurate predictions of incipient break-downs and timely maintenance efforts. An effort is made in this paper to share the experience

  8. Coolant processing device for nuclear reactor

    International Nuclear Information System (INIS)

    Kizawa, Hideo; Funakoshi, Toshio; Izumoji, Yoshiaki

    1981-01-01

    Purpose: To reduce an entire facility cost by concentrating and isolating tritium accumulated in coolants, removing the tritium out of the system, and returning hydrogen gas generated at a reactor accident to a recombiner in a closed loop by the switching of a valve. Constitution: Coolant from a reactor cooling system processed by a chemical volume control system facility (CVCS) and coolant drain from various devices processed by a liquid waste disposing system facility (LWDS) are fed to a tritium isolating facility, in which they are isolated into concentrated tritium water and dilute tritium water. The concentrated tritium water is removed out of the system and stored. The dilute tritium water is reused as supply water for coolant. If an accident occurs to cause hydrogen to be generated, a closed loop is formed between the containment vessel and the recombiner, the hydrogen is recombined with oxygen in the air of the closed loop to be thus returned to water. (Kamimura, M.)

  9. Fatigue management considering LWR coolant environments

    International Nuclear Information System (INIS)

    Park, Heung Bae; Jin, Tae eun

    2000-01-01

    Design fatigue curve for structural material in the ASME Boiler and Pressure Vessel Code do not explicitly address the effects of reactor coolant environments on fatigue life. Environmentally assisted cracking (EAC) of low-alloy steels in light water reactor (LWR) coolant environments has been a concern ever since the early 1970's. And, recent fatigue test data indicate a significant decrease in fatigue lives of carbon steels, low-alloy steels and austenitic stainless steels in LWR coolant environments. For these reasons, fatigue of major components has been identified as a technical issue remaining to be resolved for life management and license renewal of nuclear power plants. In the present paper, results of recent investigations by many organizations are reviewed to provide technical justification to support the development of utility approach regarding the management of fatigue considering LWR coolant environments for the purpose of life management and license renewal of nuclear power plants. (author)

  10. Selection of nuclear reactor coolant materials

    International Nuclear Information System (INIS)

    Shi Lisheng; Wang Bairong

    2012-01-01

    Nuclear material is nuclear material or materials used in nuclear industry, the general term, it is the material basis for the construction of nuclear power, but also a leader in nuclear energy development, the two interdependent and mutually reinforcing. At the same time, nuclear materials research, development and application of the depth and breadth of science and technology reflects a nation and the level of the nuclear power industry. Coolant also known as heat-carrier agent, is an important part of the heart nuclear reactor, its role is to secure as much as possible to the economic output in the form fission energy to heat the reactor to be used: the same time cooling the core, is controlled by the various structural components allowable temperature. This paper described the definition of nuclear reactor coolant and characteristics, and then addressed the requirements of the coolant material, and finally were introduced several useful properties of the coolant and chemical control. (authors)

  11. Standardized sampling system for reactor coolants

    International Nuclear Information System (INIS)

    Divine, J.R.; Munson, L.F.; Nelson, J.L.; McDowell, R.L.; Jankowski, M.W.

    1982-09-01

    A three-pronged approach was developed to reach the objectives of acceptable coolant sampling, assessment of occupational exposure from corrosion products, and model development for the transport and buildup of corrosion products. Emphasis is on sampler design

  12. Flow structure and heat exchange analysis in internal cooling channel of gas turbine blade

    Science.gov (United States)

    Szwaba, Ryszard; Kaczynski, Piotr; Doerffer, Piotr; Telega, Janusz

    2016-08-01

    This paper presents the study of the flow structure and heat transfer, and also their correlations on the four walls of a radial cooling passage model of a gas turbine blade. The investigations focus on heat transfer and aerodynamic measurements in the channel, which is an accurate representation of the configuration used in aeroengines. Correlations for the heat transfer coefficient and the pressure drop used in the design of radial cooling passages are often developed from simplified models. It is important to note that real engine passages do not have perfect rectangular cross sections, but include corner fillet, ribs with fillet radii and special orientation. Therefore, this work provides detailed fluid flow and heat transfer data for a model of radial cooling geometry which possesses very realistic features.

  13. Reactor coolant pump seals: improving their performance

    International Nuclear Information System (INIS)

    Pothier, N.E.; Metcalfe, R.

    1986-06-01

    Large CANDU plants are benefitting from transient-resistant four-year reliable reactor coolant pump seal lifetimes, a direct result of AECL's 20-year comprehensive seal improvement program involving R and D staff, manufacturers, and plant designers and operators. An overview of this program is presented, which covers seal modification design, testing, post-service examination, specialized maintenance and quality control. The relevancy of this technology to Light Water Reactor Coolant Pump Seals is also discussed

  14. Coolant clean up system in nuclear reactor

    International Nuclear Information System (INIS)

    Tajima, Fumio; Iwami, Hiroshi.

    1981-01-01

    Purpose: To decrease the amount of main steams and improve the plant heat efficiency by the use of condensated water as coolants for not-regenerative heat exchangers in a coolant clean up system of a nuclear reactor. Constitution: In a coolant clean up system of a nuclear reactor, a portion of condensates is transferred to the shell of a non-regenerative heat exchanger by way of a condensate pump for non-regenerative heat exchanger through a branched pipeway provided to the outlet of a condensate desalter for using the condensates as the coolants for the shell of the heat exchanger and the condensates are then returned to the inlet of a feedwater heater after the heat exchange. The branched flow rate of the condensates is controlled by the flow rate control valve mounted in the pipeway. Condensates passed through the heat exchanger and the condensates not passed through the heat exchanger are mixed and heated in a heater and then fed to the nuclear reactor. In a case where no feedwater is necessary to the nuclear reactor such as upon shutdown of the reactor, the condensates are returned by way of feedwater bypass pipeway to the condensator. By the use of the condensates as the coolants for the heat exchanger, the main steam loss can be decreased and the thermal load for the auxiliary coolant facility can be reduced. (Kawakami, Y.)

  15. Coolant Design System for Liquid Propellant Aerospike Engines

    Science.gov (United States)

    McConnell, Miranda; Branam, Richard

    2015-11-01

    Liquid propellant rocket engines burn at incredibly high temperatures making it difficult to design an effective coolant system. These particular engines prove to be extremely useful by powering the rocket with a variable thrust that is ideal for space travel. When combined with aerospike engine nozzles, which provide maximum thrust efficiency, this class of rockets offers a promising future for rocketry. In order to troubleshoot the problems that high combustion chamber temperatures pose, this research took a computational approach to heat analysis. Chambers milled into the combustion chamber walls, lined by a copper cover, were tested for their efficiency in cooling the hot copper wall. Various aspect ratios and coolants were explored for the maximum wall temperature by developing our own MATLAB code. The code uses a nodal temperature analysis with conduction and convection equations and assumes no internal heat generation. This heat transfer research will show oxygen is a better coolant than water, and higher aspect ratios are less efficient at cooling. This project funded by NSF REU Grant 1358991.

  16. Autoantibody-induced internalization of CNS AQP4 water channel and EAAT2 glutamate transporter requires astrocytic Fc receptor.

    Science.gov (United States)

    Hinson, Shannon R; Clift, Ian C; Luo, Ningling; Kryzer, Thomas J; Lennon, Vanda A

    2017-05-23

    Aquaporin-4 (AQP4) water channel-specific IgG distinguishes neuromyelitis optica (NMO) from multiple sclerosis and causes characteristic immunopathology in which central nervous system (CNS) demyelination is secondary. Early events initiating the pathophysiological outcomes of IgG binding to astrocytic AQP4 are poorly understood. CNS lesions reflect events documented in vitro following IgG interaction with AQP4: AQP4 internalization, attenuated glutamate uptake, intramyelinic edema, interleukin-6 release, complement activation, inflammatory cell recruitment, and demyelination. Here, we demonstrate that AQP4 internalization requires AQP4-bound IgG to engage an astrocytic Fcγ receptor (FcγR). IgG-lacking Fc redistributes AQP4 within the plasma membrane and induces interleukin-6 release. However, AQP4 endocytosis requires an activating FcγR's gamma subunit and involves astrocytic membrane loss of an inhibitory FcγR, CD32B. Interaction of the IgG-AQP4 complex with FcγRs triggers coendocytosis of the excitatory amino acid transporter 2 (EAAT2). Requirement of FcγR engagement for internalization of two astrocytic membrane proteins critical to CNS homeostasis identifies a complement-independent, upstream target for potential early therapeutic intervention in NMO.

  17. Influence of internal tides on Antarctic Bottom Water propagation through abyssal channels

    Science.gov (United States)

    Morozov, Eugene

    2010-05-01

    Antarctic Bottom Water (AABW) propagates in the Atlantic Ocean from the Weddell Sea to the north through narrow passages in submarine ridges. Submarine ridges are regions of strong internal tide generation in the ocean that causes mixing and eventually AABW loses its distinguishing properties such as low temperature and salinity. The Vema Fracture Zone (11 N) and Romanche Fracture Zone (equator) in the Mid-Atlantic Ridge (MAR) are pathways for AABW to the Northeast Atlantic. The deep basin of the Northeast Atlantic (Canary Basin and Gambia Abyssal Plain) are filled with the bottom water propagating through the Vema FZ rather than through the equatorial fracture zones because strong internal tides and mixing over the slopes of the MAR near the equator cause warming of AABW and decrease of its density. Further propagation of AABW through the Kane Gap is low. Recent field measurements in the fracture zones confirm this concept based on modeling results. Results of recent cruises are presented.

  18. Sound velocity in the coolant of boiling nuclear reactors

    International Nuclear Information System (INIS)

    Proskuryakov, K.N.; Parshin, D.A.; Novikov, K.S.; Galivec, E.Yu.

    2009-01-01

    To prevent resonant interaction between acoustic resonance and natural frequencies of FE, FA and RI oscillations, it is necessary to determine the value of EACPO. Based on results of calculations of EACPO and natural frequencies of FR, FA and RI oscillations values, it would be possible to reveal the dynamical loadings on metal that are dangerous for the initiation of cracking process in the early stage of negative condition appearance. To calculate EACPO it is necessary to know the Speed Velocity in Coolant. Now we do not have any data about real values of such important parameter as pressure pulsations propagation velocity in two phase environments, especially in conditions with variations of steam content along the length of FR, with taking into account the type of local resistances, flow geometry etc. While areas of resonant interaction of the single-phase liquid coolant with equipment and internals vibrations are estimated well enough, similar estimations in the conditions of presence of a gas and steam phase in the liquid coolant are inconvenient till now. Paper presents results of calculation of velocity of pressure pulsations distribution in two-phase flow formed in core of RBMK-1000 reactors. Feature of the developed techniques is that not only thermodynamic factors and effect of a speed difference between water and steam in a two phase flow but also geometrical features of core, local resistance, non heterogeneity in the two phase environment and power level of a reactor are considered. Obtained results evidence noticeable decreasing of velocity propagation of pressure pulsations in the presence of steam actions in the liquids. Such estimations for real RC of boiling nuclear reactors with steam-liquid coolant are obtained for the first time. (author)

  19. Application of damage function analysis to reactor coolant circuits

    International Nuclear Information System (INIS)

    MacDonald, D.D.

    2002-01-01

    The application of deterministic models for simulating stress corrosion cracking phenomena in Boiling Water Reactor primary coolant circuits is described. The first generation code, DAMAGE-PREDICTOR, has been used to model the radiolysis of the coolant, to estimate the electrochemical corrosion potential (ECP), and to calculate the crack growth rate (CGR) at fixed state points during reactor operation in about a dozen plants worldwide. This code has been validated in ''double-blind'' comparisons between the calculated and measured hydrogen concentration, oxygen concentration, and ECP in the recirculation system of the Leibstadt BWR in Switzerland, as well as through less formal comparisons with data from other plants. Second generation codes have now been developed, including REMAIN for simulating BWRs with internal coolant pumps and the ALERT series for modeling reactors with external pumps. One of this series, ALERT, yields the integrated damage function (IDF), which is the crack length versus time, on a component-by-component basis for a specified future operating scenario. This code therefore allows one to explore proposed future operating protocols, with the objective of identifying those that are most cost-effective and which minimizes the risk of failure of components in the coolant circuit by stress corrosion cracking. The application of this code is illustrated by exploring the benefits of partial hydrogen water chemistry (HWC) for an actual reactor, in which hydrogen is added to the feedwater over only limited periods during operation. The simulations show that the benefits, in terms of reduction in the IDFs for various components, are sensitive to when HWC was initiated in the plant life and to the length of time over which it is applied. (author)

  20. Application of damage function analysis to reactor coolant circuits

    Energy Technology Data Exchange (ETDEWEB)

    MacDonald, D.D. [Center for Electrochemical Science and Technology, Pennsylvania State Univ., University Park, PA (United States)

    2002-07-01

    The application of deterministic models for simulating stress corrosion cracking phenomena in Boiling Water Reactor primary coolant circuits is described. The first generation code, DAMAGE-PREDICTOR, has been used to model the radiolysis of the coolant, to estimate the electrochemical corrosion potential (ECP), and to calculate the crack growth rate (CGR) at fixed state points during reactor operation in about a dozen plants worldwide. This code has been validated in ''double-blind'' comparisons between the calculated and measured hydrogen concentration, oxygen concentration, and ECP in the recirculation system of the Leibstadt BWR in Switzerland, as well as through less formal comparisons with data from other plants. Second generation codes have now been developed, including REMAIN for simulating BWRs with internal coolant pumps and the ALERT series for modeling reactors with external pumps. One of this series, ALERT, yields the integrated damage function (IDF), which is the crack length versus time, on a component-by-component basis for a specified future operating scenario. This code therefore allows one to explore proposed future operating protocols, with the objective of identifying those that are most cost-effective and which minimizes the risk of failure of components in the coolant circuit by stress corrosion cracking. The application of this code is illustrated by exploring the benefits of partial hydrogen water chemistry (HWC) for an actual reactor, in which hydrogen is added to the feedwater over only limited periods during operation. The simulations show that the benefits, in terms of reduction in the IDFs for various components, are sensitive to when HWC was initiated in the plant life and to the length of time over which it is applied. (author)

  1. Models for coolant void reactivity evaluation in Candu Generation II and III+

    International Nuclear Information System (INIS)

    Popov, Alexi V.; Chambon, Richard P.; Le Tellier, Romain; Marleau, Guy; Hebert, Alain

    2008-01-01

    In the simulation of large-break loss-of-coolant accidents, homogenised cross-sections from trans- port calculations are used. These are usually computed in single cells or lattices representative for an infinite repeated pattern. Large coolant accidents in Candu, however, usually exhibit a checkerboard pattern of cooled and voided channels represented by lattices. It is reasonable, therefore, that homogenised cross-sections be produced in assemblies of lattices. This allows simulating the checkerboard voiding pat- tern and more realistically reproducing the lattice boundary conditions. The result is better simulation of the accident and more precise evaluation of coolant-void reactivity. For the present study, homogenised cross-sections are generated in a 2x2 heterogeneous assembly of four lattices for Generation II and III+ Candu designs. Results of reactivity calculations with the reactor code are compared to those using the traditional method. The difference is significant for Generation III+ Candu. (authors)

  2. The effect of channel flow pattern on internal properties distribution of a proton exchange membrane fuel cell for cathode starvation conditions

    International Nuclear Information System (INIS)

    Ko, Dong Soo; Kang, Young Min; Yang, Jang Sik; Jeong, Ji Hwan; Choi, Gyung Min; Kim, Duck Jool

    2010-01-01

    The effect of channel flow pattern on the internal properties distribution of a proton exchange membrane fuel cell (PEMFC) for cathode starvation conditions in a unit cell was investigated through numerical studies and experiments. The polarization curves of a lab-scale mixed serpentine PEMFC were measured with increasing current loads for different cell temperatures (40, 50, and 60 .deg. C) at a relative humidity of 100%. To study the local temperature on the membrane, the water content in the MEA, and the gas velocity in terms of the channel type of the PEMFC with operating characteristics, numerical studies using the es-pemfc module of STAR-CD, which have been matched to the experimental data, were conducted in detail. The water content and velocity at the cathode channel bend of the mixed serpentine channel were relatively higher than those at the single and double channels. Conversely, the local temperature and mean temperature on the membrane of a single serpentine channel were the highest among all channels. These results can be used to design the PEMFC system, the channel flow field, and the cooling device

  3. Determination of mixture coefficients in tests simulating channels of a fuel assembly in a water cooled reactor

    International Nuclear Information System (INIS)

    Ferreira, W.R.

    1983-09-01

    Here, the mixture coefficients are determined in tests which simulate two symmetric and two assymetric coolant channels of a water cooled reactor. It was studies the effects of coolant flow and of the distance among the coolant channels on these coefficients. The technique used to determine the mixture coefficients was to introduce a tracer (methylene blue) into the flow. The determination of the tracer concentration at the end of the channels was made by means of spectrophotometric analysis. (author)

  4. RETRAN analysis of inter-system LOCA within the primary coolant pump

    International Nuclear Information System (INIS)

    Gangadharan, A.; Pratt, G.F.

    1992-01-01

    One example of an inter-system loss of coolant accident is the failure of the tubing within the primary coolant pump (PCP) thermal barrier heat exchanger. Such a failure would result in the entry of primary coolant into the component cooling water (CCW) system. The primary coolant flowrate through the break would rapidly pressurize the CCW system when the relief valves are too small. The piping in the CCW system at Palisades has a low pressure rating. Failures in this system outside the containment boundary could lead to primary coolant release to the atmosphere. RETRAN-02 was used to perform a simulation of the break in the PCP integral heat exchanger. The model included a detailed nodalization of the Byron-Jackson primary coolant pump internals leading up to the CCW system relief valves. Preliminary studies show the need for increased relief capacity in the CCW system. A case was run using a larger relief valve. Critical flow in the system upstream of the relief valves maintains the pressures in those volumes above the CCW design pressure. The pressures downstream from the relief valves and outside containment will be at or below the design pressure. This paper presents the results of the transient analysis

  5. Simulation of a loss of coolant accident

    International Nuclear Information System (INIS)

    1987-06-01

    An essential component of nuclear safety activities is the analysis of postulated accidents which are taken as a design basis for a facility. This analysis is usually carried out by using complex computer codes to simulate the behaviour of the plant and to calculate vital plant parameters, which are then compared with the design limits. Since these simulations cannot be verified at the plant itself, computer codes must be validated by comparing the results of calculations with experimental data obtained in test facilities. With this objective in mind, the Central Research Institute for Physics (CRIP) of the Hungarian Academy of Sciences designed and constructed the PMK-NVH (Paks Model Circuit) test facility, a scaled down model of the WWER-440 Paks nuclear power plant. Hungary with the aim of strengthening the international co-operation on nuclear safety, made the PMK-NVH facility available to the IAEA to conduct a standard problem exercise. In this exercise, experimental data from the simulation of a 7.4% break loss of coolant accident were compared with analytical predictions of the behaviour of the facility calculated with computer codes. This document presents a complete overview of the Standard Problem Exercise, including description of the facility, the experiment, the codes and models used by the participants and a detailed intercomparison of calculated and experimental results. It is recognized that code assessment is a long process which involves many inter-related steps, therefore, no general conclusion on optimum code or best model was reached. However, the exercise was recognized as an important contributor to code validation

  6. An experimental investigation of the post-CHF enhancement factor for a prototypical ITER divertor plate with water coolant

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, T.D. [Rensselaer Polytechnic Institute, Troy, NY (United States); Watson, R.D.; McDonald, J.M. [Sandia National Lab., Albuquerque, NM (United States)] [and others

    1995-09-01

    In an off-normal event, water-cooled copper divertor plates in the International Thermonuclear Experimental Reactor (ITER) may either experience heat loads beyond their design basis, or the normal heat loads may be accompanied by low coolant pressure and velocity. The purpose of this experiment was to illustrate that during one-sided heating, as in ITER, a copper divertor plate with the proper side wall thickness, at low system pressure and velocity can absorb without failing an incident heat flux, q{sub i}, that significantly exceed the value, q{sub i}{sup CHF}, which is associated with local CHF at the wall of the coolant channel. The experiment was performed using a 30 kW electron beam test system for heating of a square cross-section divertor heat sink with a smooth circular channel of 7.63 mm diameter. The heated width, length, and wall thickness were 16, 40, and 3 mm, respectively. Stable surface temperatures were observed at incident heat fluxes greater than the local CHF point, presumably due to circumferential conduction around the thick tube walls when q{sub i}{sup CHF} was exceeded. The Post-CHF enhancement factor, {eta}, is defined as the ratio of the incident burnout heat flux, q{sub i}{sup BO}, to q{sub i}{sup CHF}. For this experiment with water at inlet conditions of 70{degrees}C, 1 m/s, and 1 MPa, q{sub i}{sup CHF} and q{sub i}{sup BO} were 600 and 1100 W/cm{sup 2}, respectively, which gave an {eta} of 1.8.

  7. WIMP search in the mono-photon channel at the international linear collider

    Energy Technology Data Exchange (ETDEWEB)

    Habermehl, Moritz [Deutsches Elektronen-Synchrotron (DESY), Notkestrasse 85, 22607 Hamburg (Germany); Universitaet Hamburg, Institut fuer Experimentalphysik, Luruper Chaussee 149, 22761 Hamburg (Germany); List, Jenny [Deutsches Elektronen-Synchrotron (DESY), Notkestrasse 85, 22607 Hamburg (Germany)

    2016-07-01

    The International Linear Collider (ILC) is a planned electron-positron collider with √(s) tunable from 250 to 500 GeV, with a possible upgrade to 1 TeV. Besides precision measurements of the Higgs boson its physics goals comprise searches for physics beyond the Standard Model, e.g. searches for Dark Matter. This collider search assumes the production of WIMPs in pairs. They are not visible in the detector but the energy carried away can be observed via an additional (''tag'') particle. Photon emission from the initial state leads to the almost model independent signature: e{sup +}e{sup -} → χχγ. As this analysis tests couplings between WIMPs and leptons it is complementary to analogues searches at the LHC. A precise study is facilitated by the clean environment of lepton colliders with small systematics of electroweak backgrounds. While the conceptual feasibility and the sensitivity reach of the ILC have been shown in the past, this talk focusses on the consequences for the detector design. The requirements for the central detector as well as for the instrumentation of the forward region are discussed in the context of the ILD detector concept.

  8. ENVIRONMENTALLY REDUCING OF COOLANTS IN METAL CUTTING

    Directory of Open Access Journals (Sweden)

    Veijo KAUPPINEN

    2012-11-01

    Full Text Available Strained environment is a global problem. In metal industries the use of coolant has become more problematic in terms of both employee health and environmental pollution. It is said that the use of coolant forms approximately 8 - 16 % of the total production costs.The traditional methods that use coolants are now obviously becoming obsolete. Hence, it is clear that using a dry cutting system has great implications for resource preservation and waste reduction. For this purpose, a new cooling system is designed for dry cutting. This paper presents the new eco-friendly cooling innovation and the benefits gained by using this method. The new cooling system relies on a unit for ionising ejected air. In order to compare the performance of using this system, cutting experiments were carried out. A series of tests were performed on a horizontal turning machine and on a horizontal machining centre.

  9. Iron crud supply device to reactor coolant

    International Nuclear Information System (INIS)

    Baba, Takao.

    1993-01-01

    In a device for supplying iron cruds into reactor coolants in a BWR type power plant, a system in which feed water containing iron cruds is supplied to the reactor coolants after once passing through an ion exchange resin is disposed. As a result, iron cruds having characteristics similar with those of naturally occurring iron cruds in the plant are obtained and they react with ionic radioactivity, to form composite oxides. Then, iron cruds having high performance of being secured to the surface of a fuel cladding tube can be supplied to the reactor coolants, thereby enabling to greatly reduce the density of reactor water ionic radioactivity. In its turn, dose rate on the surface of pipelines can be reduced, thereby enabling to reduce operators' radiation exposure dose in the plant. Further, contamination of a condensate desalting device due to iron cruds can be prevented, and further, the density of the iron cruds supplied can easily be controlled. (N.H.)

  10. Limits to fuel/coolant mixing

    International Nuclear Information System (INIS)

    Corradini, M.L.; Moses, G.A.

    1985-01-01

    The vapor explosion process involves the mixing of fuel with coolant prior to the explosion. A number of analysts have identified limits to the amount of fuel/coolant mixing that could occur within the reactor vessel following a core melt accident. Past models are reviewed and a sim plified approach is suggested to estimate the upper limit on the amount of fuel/coolant mixing pos sible. The approach uses concepts first advanced by Fauske in a different way. The results indicat that water depth is an important parameter as well as the mixing length scale D /SUB mix/ , and for large values of D /SUB mix/ the fuel mass mixed is limited to <7% of the core mass

  11. Fuel channel refilling

    International Nuclear Information System (INIS)

    Shoukri, M.; Abdul-Razzak, A.

    1992-11-01

    This report extends the work reported in document INFO-0370 on fuel channel refilling by providing analysis of the refilling tests conducted using the RD-14 and RD-14M test facilities. The analysis focuses on the general thermal-hydraulic characteristics of the facilities following various small and large inlet and outlet header breaks with emergency coolant injection. The two-fluid model thermal-hydraulic computer code CATHENA was tested against results obtained from selected experiments carried out in the two facilities. Conclusions related to the effect of break size, mode of emergency core injection, primary pump operation and parallel channels are presented. (Author) (116 figs., 17 tabs., 53 refs.)

  12. Main coolant pump testing at Ontario Hydro

    International Nuclear Information System (INIS)

    Hartlen, R.

    1991-01-01

    This article describes Ontario Hydro Research Division's experience with a computerized data acquisition and analysis system for monitoring mechanical vibration in reactor coolant pumps. The topics covered include bench-marking of the computer system and the coolant pumps, signatures of normal and malfunctioning pumps, analysis of data collected by the monitoring system, simulation of faults, and concerns that have been expressed about data interpretation, sensor types and locations, alarm/shutdown limits and confirmation of nondestructive examination testing. This presentation consists of overheads only

  13. Comparative design study of FR plants with various coolants. 1. Studies on Na coolant FR, Pb-Bi coolant FR, gas coolant FR

    International Nuclear Information System (INIS)

    Konomura, Mamoru; Shimakawa, Yoshio; Hori, Toru; Kawasaki, Nobuchika; Enuma, Yasuhiro; Kida, Masanori; Kasai, Shigeo; Ichimiya, Masakazu

    2001-01-01

    In Phase I of the Feasibility Studies on the Commercialized Fast Reactor (FR) Cycle System, plant designs on FR were performed with various coolants. This report describes the plant designs on FR with sodium, lead-bismuth, CO 2 gas and He gas coolants. A construction cost of 0.2 million yen/kWe was set up as a design goal. The result is as follows: The sodium reactor has a capability to obtain the goal, and lead-bismuth and gas reactors may satisfy the goal with further improvements. (author)

  14. On-Line Coolant Chemistry Analysis

    International Nuclear Information System (INIS)

    LM Bachman

    2006-01-01

    Impurities in the gas coolant of the space nuclear power plant (SNPP) can provide valuable indications of problems in the reactor and an overall view of system health. By monitoring the types and amounts of these impurities, much can be implied regarding the status of the reactor plant. However, a preliminary understanding of the expected impurities is important before evaluating prospective detection and monitoring systems. Currently, a spectroscopy system is judged to hold the greatest promise for monitoring the impurities of interest in the coolant because it minimizes the number of entry and exit points to the plant and provides the ability to detect impurities down to the 1 ppm level

  15. Leak detection device for reactor coolant

    International Nuclear Information System (INIS)

    Oshima, Koichiro.

    1990-01-01

    In a light water cooled reactor, if reactor coolants are leaked from pipelines in a pipeline chamber, activated products (N-16) are diffused together to an atmosphere in the pipeline chamber. N-16 is sucked from an extracting tube which is always sucking the atmosphere in the pipeline chamber to a sucking blower. Then, β-rays released from N-16 are monitored by a radiation monitor in a measuring chamber which is radiation-shielded from the pipeline chamber. Accordingly, since the radiation monitor can detect even slight leakage, the slight leakage of reactor coolants in the pipelines can be detected at an early stage. (I.N.)

  16. Reactor coolant pump for a nuclear reactor

    International Nuclear Information System (INIS)

    Burkhardt, W.; Richter, G.

    1976-01-01

    An improvement is proposed concerning the easier disengagement of the coupling at the reactor coolant pump for a nuclear reactor transporting a pressurized coolant. According to the invention the disengaging coupling consists of two parts separated by screws. At least one of the screws contains a propellent charge ananged within a bore and provided with a speed-dependent ignition device in such a way that by separation of the screws at overspeeds the coupling is disengaged. The sub-claims are concerned with the kind of ignition ot the propellent charge. (UWI) [de

  17. Ion channeling

    International Nuclear Information System (INIS)

    Erramli, H.; Blondiaux, G.

    1994-01-01

    Channeling phenomenon was predicted, many years ago, by stark. The first channeling experiments were performed in 1963 by Davies and his coworkers. Parallely Robinson and Oen have investigated this process by simulating trajectories of ions in monocrystals. This technique has been combined with many methods like Rutherford Backscattering Spectrometry (R.B.S.), Particles Induced X-rays Emission (P.I.X.E) and online Nuclear Reaction (N.R.A.) to localize trace elements in the crystal or to determine crystalline quality. To use channeling for material characterization we need data about the stopping power of the incident particle in the channeled direction. The ratios of channeled to random stopping powers of silicon for irradiation in the direction have been investigated and compared to the available theoretical results. We describe few applications of ion channeling in the field of materials characterization. Special attention is given to ion channeling combined with Charged Particle Activation Analysis (C.P.A.A.) for studying the behaviour of oxygen atoms in Czochralski silicon lattices under the influence of internal gettering and in different gaseous atmospheres. Association between ion channeling and C.P.A.A was also utilised for studying the influence of the growing conditions on concentration and position of carbon atoms at trace levels in the MOVPE Ga sub (1-x) Al sub x lattice. 6 figs., 1 tab., 32 refs. (author)

  18. Balancing the people, profit and planet dimensions in international marketing channels : a study on coordinating mechanisms in the Nile perch channel from Lake Victoria

    NARCIS (Netherlands)

    Kambewa, E.

    2007-01-01

    Keywords: people, profit, planet, marketing channels, contracts, sustainability, quality,LakeVictoria, conjointanalysis

  19. Protein kinase A-induced internalization of Slack channels from the neuronal membrane occurs by adaptor protein-2/clathrin-mediated endocytosis.

    Science.gov (United States)

    Gururaj, Sushmitha; Evely, Katherine M; Pryce, Kerri D; Li, Jun; Qu, Jun; Bhattacharjee, Arin

    2017-11-24

    The sodium-activated potassium (K Na ) channel Kcnt1 (Slack) is abundantly expressed in nociceptor (pain-sensing) neurons of the dorsal root ganglion (DRG), where they transmit the large outward conductance I KNa and arbitrate membrane excitability. Slack channel expression at the DRG membrane is necessary for their characteristic firing accommodation during maintained stimulation, and reduced membrane channel density causes hyperexcitability. We have previously shown that in a pro-inflammatory state, a decrease in membrane channel expression leading to reduced Slack-mediated I KNa expression underlies DRG neuronal sensitization. An important component of the inflammatory milieu, PKA internalizes Slack channels from the DRG membrane, reduces I KNa , and produces DRG neuronal hyperexcitability when activated in cultured primary DRG neurons. Here, we show that this PKA-induced retrograde trafficking of Slack channels also occurs in intact spinal cord slices and that it is carried out by adaptor protein-2 (AP-2) via clathrin-mediated endocytosis. We provide mass spectrometric and biochemical evidence of an association of native neuronal AP-2 adaptor proteins with Slack channels, facilitated by a dileucine motif housed in the cytoplasmic Slack C terminus that binds AP-2. By creating a competitive peptide blocker of AP-2-Slack binding, we demonstrated that this interaction is essential for clathrin recruitment to the DRG membrane, Slack channel endocytosis, and DRG neuronal hyperexcitability after PKA activation. Together, these findings uncover AP-2 and clathrin as players in Slack channel regulation. Given the significant role of Slack in nociceptive neuronal excitability, the AP-2 clathrin-mediated endocytosis trafficking mechanism may enable targeting of peripheral and possibly, central neuronal sensitization. © 2017 by The American Society for Biochemistry and Molecular Biology, Inc.

  20. Coolant cleanup system for BWR type reactor

    International Nuclear Information System (INIS)

    Kinoshita, Shoichiro; Araki, Hidefumi.

    1993-01-01

    The cleanup system of the present invention removes impurity ions and floating materials accumulated in a reactor during evaporation of coolants in the nuclear reactor. That is, coolants pass pipelines from a pressure vessel using pressure difference between a high pressure in the pressure vessel and a low pressure at the upstream of a condensate filtration/desalting device of a condensate/feed water system as a driving source, during which cations and floating materials are removed in a high temperature filtration/desalting device and coolants flow into the condensate/feedwater system. Impurities containing anions are removed here by the condensates filtration/desalting device. Then, they return to the pressure vessel while pressurized and heated by a condensate pump, a feed water pump and a feed water heater. At least pumps, a heat exchanger for heating, a filtration/desalting device for removing anions and pipelines connecting them used exclusively for the coolant cleanup system are no more necessary. (I.S.)

  1. Fission product release into the primary coolant

    International Nuclear Information System (INIS)

    Apperson, C.E.

    1977-01-01

    The analytic evaluation of steady state primary coolant activity is discussed. The reported calculations account for temperature dependent fuel failure in two particle types and arbitrary radioactive decay chains. A matrix operator technique implemented in the SUVIUS code is used to solve the simultaneous equations. Results are compared with General Atomic Company's published results

  2. Comparison of thermohydraulic characteristics in the use of various coolants

    International Nuclear Information System (INIS)

    Muramatsu, Toshiharu; Suda, Kazunori; Yamaguchi, Akira

    2000-11-01

    Numerical calculations were carried out for a free surface sloshing, a thermal stratification, a thermal striping, and a natural convection as key phenomena of in-vessel thermohydraulics in future fast reactor systems with various fluids as coolants. This numerical work was initiated based on a recognition that the fundamental characteristics of the phenomena have been unsolved quantitatively in the use of various coolants. From the analysis for the phenomena, the following results were obtained. [Free Surface Sloshing phenomena] (1) There is no remarkable difference between liquid sodium and liquid Pb-Bi in characteristics of internal flows and free surface characteristics based on Fr number. (2) The AQUA-VOF code has a potential enough to evaluate gas entrainment behavior from the free surface including the internal flow characteristics. [Thermal Stratification Phenomena] (1) On-set position of thermal entrainment process due to dynamic vortex flows was moved to downstream direction with decreasing of Ri number. On the other hand, the position in the case of CO 2 gas was shifted to upstream side with decreasing of Ri number. (2) Destruction speed of the thermal stratification interface was dependent on thermal diffusivity as fluid properties. Therefore it was concluded that an elimination method is necessary for the interface generated in CO 2 gas. [Thermal Striping Phenomena] (1) Large amplitudes of fluid temperature fluctuations was reached to down stream area in the use of CO 2 gas, due to larger fluid viscosity and smaller thermal diffusivity, compared with liquid sodium and liquid Pb-Bi cases. (2) To simulate thermal striping conditions such as amplitude and frequency of the fluid temperature fluctuations, it is necessary for coincidences of Re number for the amplitude and of velocity value for the frequency, in various coolants. [Natural Convection Phynomlena] (1) Fundamental behavior of the natural convection in various coolant follows buoyant jet

  3. Optimum coolant chemistry in BWRs

    International Nuclear Information System (INIS)

    Lin, C.C.; Cowan, R.L.; Kiss, E.

    2004-01-01

    LWR water chemistry parameters are directly or indirectly related to the plant's operational performance and for a significant amount of Operation and Maintenance (O and M) costs. Obvious impacts are the operational costs associated with water treatment, monitoring and associated radwaste generation. Less obvious is the important role water chemistry plays in the magnitude of drywell shutdown dose rates, fuel corrosion performance and, (probably most importantly) materials degradation such as from stress corrosion cracking of piping and Reactor Pressure Vessel (RPV) internal components. To improve the operational excellence of the BWR and to minimize the impact of water chemistry on O and M costs. General Electric has developed the concept of Optimum Water Chemistry (OWC). The 'best practices' and latest technology findings from the U.S., Asia and Europe are integrated into the suggested OWC Specification. This concept, together with cost effective ways to meet the requirement, are discussed. (author)

  4. Studies of loss-of-coolant and loss-of-regulation accidents

    International Nuclear Information System (INIS)

    Rogers, J.T.

    1979-10-01

    Studies of a CANDU reactor during loss of coolant with delayed emergency core cooling showed that the moderator is an effective heat sink, and that in reactors with moderator dump the calandria sprays provide effective cooling. Fuel channel melting would not occur, and a coolable geometry will be maintained. Studies on film cooling and film stability on calandria tubes and on the analysis of flow reversal in vertical feeder tubes are also reported

  5. MABEL-2: a code to analyse cladding deformation in a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Bowring, R.W.; Cooper, C.A.; Haste, T.J.

    1982-04-01

    MABEL can be used to determine the cladding deformation in a PWR during a LOCA. It takes the results of calculations from other codes to define the initial fuel condition and the transient whole core thermal-hydraulic behaviour. The use of MABEL with input data appropriate to different regions of a reactor core allows an overall picture of coolant channel blockage within the core to be obtained. (U.K.)

  6. Fuel and coolant motions following pin failure: EPIC models and the PBE-5S experiment

    International Nuclear Information System (INIS)

    Garner, P.L.; Abramson, P.B.

    1979-01-01

    The EPIC computer code has been used to analyze the post-fuel-pin-failure behavior in the PBE-5S experiment performed at Sandia Laboratories. The effects of modeling uncertainties on the calculation are examined. The calculations indicate that the majority of the piston motion observed in the test is due to the initial pressurization of the coolant channel by fuel vapor at cladding failure. A more definitive analysis requires improvements in calculational capabilities and experiment diagnostics

  7. A limited 4 Å radial displacement of the S4-S5 linker is sufficient for internal gate closing in Kv channels.

    Science.gov (United States)

    Faure, Élise; Starek, Greg; McGuire, Hugo; Bernèche, Simon; Blunck, Rikard

    2012-11-16

    Voltage-gated ion channels are responsible for the generation of action potentials in our nervous system. Conformational rearrangements in their voltage sensor domains in response to changes of the membrane potential control pore opening and thus ion conduction. Crystal structures of the open channel in combination with a wealth of biophysical data and molecular dynamics simulations led to a consensus on the voltage sensor movement. However, the coupling between voltage sensor movement and pore opening, the electromechanical coupling, occurs at the cytosolic face of the channel, from where no structural information is available yet. In particular, the question how far the cytosolic pore gate has to close to prevent ion conduction remains controversial. In cells, spectroscopic methods are hindered because labeling of internal sites remains difficult, whereas liposomes or detergent solutions containing purified ion channels lack voltage control. Here, to overcome these problems, we controlled the state of the channel by varying the lipid environment. This way, we directly measured the position of the S4-S5 linker in both the open and the closed state of a prokaryotic Kv channel (KvAP) in a lipid environment using Lanthanide-based resonance energy transfer. We were able to reconstruct the movement of the covalent link between the voltage sensor and the pore domain and used this information as restraints for molecular dynamics simulations of the closed state structure. We found that a small decrease of the pore radius of about 3-4 Å is sufficient to prevent ion permeation through the pore.

  8. Analysis of Coolant Options for Advanced Metal Cooled Nuclear Reactors

    National Research Council Canada - National Science Library

    Can, Levent

    2006-01-01

    .... The overall focus of this study is the build up of induced radioactivity in the coolant of metal cooled reactors as well as the evaluation of other physical and chemical properties of such coolants...

  9. Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Physicochemical effects

    Energy Technology Data Exchange (ETDEWEB)

    Kryk, Holger, E-mail: h.kryk@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany); Hoffmann, Wolfgang [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany); Kästner, Wolfgang; Alt, Sören; Seeliger, André; Renger, Stefan [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany)

    2014-12-15

    Highlights: • Physicochemical effects due to post-LOCA zinc corrosion in PWR were elucidated. • Decreasing solubility of corrosion products with increasing temperature was found. • Solid corrosion products may be deposited on hot surfaces and/or within hot-spots. • Corrosion products precipitating from coolant were identified as zinc borates. • Depending on coolant temperature, different types of zinc borate are formed. - Abstract: Within the framework of the reactor safety research, generic experimental investigations were carried out aiming at the physicochemical background of possible zinc corrosion product formation, which may occur inside the reactor pressure vessel during the sump circulation operation after loss-of-coolant accidents in pressurized water reactors. The contact of the boric acid containing coolant with hot-dip galvanized steel containment internals causes corrosion of the corresponding materials resulting in dissolution of the zinc coat. A retrograde solubility of zinc corrosion products with increasing temperature was observed during batch experiments of zinc corrosion in boric acid containing coolants. Thus, the formation and deposition of solid corrosion products cannot be ruled out if the coolant containing dissolved zinc is heated up during its recirculation into hot regions within the emergency cooling circuit (e.g. hot-spots in the core). Corrosion experiments at a lab-scale test facility, which included formation of corrosion products at a single heated cladding tube, proved that dissolved zinc, formed at low temperatures in boric acid solution by zinc corrosion, turns into solid deposits of zinc borates when contacting heated zircaloy surfaces during the heating of the coolant. Moreover, the temperature of formation influences the chemical composition of the zinc borates and thus the deposition and mobilization behavior of the products.

  10. Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Physicochemical effects

    International Nuclear Information System (INIS)

    Kryk, Holger; Hoffmann, Wolfgang; Kästner, Wolfgang; Alt, Sören; Seeliger, André; Renger, Stefan

    2014-01-01

    Highlights: • Physicochemical effects due to post-LOCA zinc corrosion in PWR were elucidated. • Decreasing solubility of corrosion products with increasing temperature was found. • Solid corrosion products may be deposited on hot surfaces and/or within hot-spots. • Corrosion products precipitating from coolant were identified as zinc borates. • Depending on coolant temperature, different types of zinc borate are formed. - Abstract: Within the framework of the reactor safety research, generic experimental investigations were carried out aiming at the physicochemical background of possible zinc corrosion product formation, which may occur inside the reactor pressure vessel during the sump circulation operation after loss-of-coolant accidents in pressurized water reactors. The contact of the boric acid containing coolant with hot-dip galvanized steel containment internals causes corrosion of the corresponding materials resulting in dissolution of the zinc coat. A retrograde solubility of zinc corrosion products with increasing temperature was observed during batch experiments of zinc corrosion in boric acid containing coolants. Thus, the formation and deposition of solid corrosion products cannot be ruled out if the coolant containing dissolved zinc is heated up during its recirculation into hot regions within the emergency cooling circuit (e.g. hot-spots in the core). Corrosion experiments at a lab-scale test facility, which included formation of corrosion products at a single heated cladding tube, proved that dissolved zinc, formed at low temperatures in boric acid solution by zinc corrosion, turns into solid deposits of zinc borates when contacting heated zircaloy surfaces during the heating of the coolant. Moreover, the temperature of formation influences the chemical composition of the zinc borates and thus the deposition and mobilization behavior of the products

  11. Determination of two dimensional axisymmetric finite element model for reactor coolant piping nozzles

    International Nuclear Information System (INIS)

    Choi, S. N.; Kim, H. N.; Jang, K. S.; Kim, H. J.

    2000-01-01

    The purpose of this paper is to determine a two dimensional axisymmetric model through a comparative study between a three dimensional and an axisymmetric finite element analysis of the reactor coolant piping nozzle subject to internal pressure. The finite element analysis results show that the stress adopting the axisymmetric model with the radius of equivalent spherical vessel are well agree with that adopting the three dimensional model. The radii of equivalent spherical vessel are 3.5 times and 7.3 times of the radius of the reactor coolant piping for the safety injection nozzle and for the residual heat removal nozzle, respectively

  12. Fabrication of micro-channel arrays on thin metallic sheet using internal fluid pressure: Investigations on size effects and development of design guidelines

    Energy Technology Data Exchange (ETDEWEB)

    Mahabunphachai, Sasawat [NSF I/UCR Center for Precision Forming, Department of Mechanical Engineering, Virginia Commonwealth University, Richmond, VA 23284 (United States); Department of Mechanical Engineering, University of Michigan, Ann Arbor, MI 48109 (United States); Koc, Muammer [NSF I/UCR Center for Precision Forming, Department of Mechanical Engineering, Virginia Commonwealth University, Richmond, VA 23284 (United States)

    2008-01-03

    Micro-feature (channel, protrusion, cavity, etc.) arrays on large area-thin metallic sheet alloys are increasingly needed for compact and integrated heat/mass transfer applications (such as fuel cells and fuel processors) that require high temperature resistance, corrosion resistance, good electrical/thermal conductivity, etc. The performance of these micro-feature arrays mainly affects the volume flow velocity of the reactants inside the arrays which directly controls the rate of convection mass/heat transport. The key factors that affect the flow velocity include channel size and shape, flow field pattern, flow path length, fluid pressure, etc. In this study, we investigated these micro-feature arrays from the manufacturability perspective since it is also an important factor to be considered in the design process. Internal fluid pressure (hydroforming) technique is investigated in this study with the specific goals to, first, understand if the so-called ''size effects'' (grain vs. feature size) are effective on the manufacturability of thin metallic sheet into micro-channels, and second, to establish design guidelines for the micro-channel hydroforming technique for robust mass production conditions. Thin stainless steel 304 blanks of 0.051 mm thick with three different grain sizes of 9.3, 10.6, and 17.0 {mu}m were used in hydroforming experiments to form micro-channels with the dimensions between 0.46-1.33 and 0.15-0.98 mm in width and height, respectively. Based on the experimental results, the effect of the grain size on the channel formability was found to be insignificant for the grain size range used in this study. On the other hand, the effect of the channel (feature) size was shown to dominate the overall formability. In addition, FE models of the process were developed and validated with the experimental results, then used to conduct a parametric study to establish micro-channel design guidelines. The results from the parametric

  13. Simulation of LLCB TBM in-vessel first wall coolant break into ITER vacuum vessel by using RELAP/MOD3.4

    International Nuclear Information System (INIS)

    Tony Sandeep, K.; Chaudhari, Vilas; Rajendra Kumar, E.; Dutta, Anu; Singh, R.K.

    2013-06-01

    To prove Test Blanket Module (TBM) safety in International Thermonuclear Experimental Reactor (ITER), various accident scenarios are postulated . One of the postulated initiating events to be analysed is TBM First wall (FW) coolant leak in ITER Vacuum vessel (VV). This accident has been classified as a reference event for the TBM (probability of occurrence >1 E -06 /a). The postulated accident occurs as a result of small leak of TBM FW helium into ITER vacuum vessel (VV), caused by the TBM weld failure. The ingress of this TBM FW helium into ITER plasma induces intense plasma disruption that deposits 1.8 MJ/m 2 of plasma stored thermal energy onto the TBM FW over a period of 1 sec in duration (assumption). Runaway electrons in this process are lost from plasma current channel and cause multiple TBM and ITER FW cooling tube failures within 10 cm torriodal strip. The size of the break is identified as double ended rupture of all coolant channels within this strip around the reactor. For LLCB TBM this represents failure of 4 FW channels. The size of ITER FW break is 0.02 m 2 . Consequently, a simultaneous blow down of TBM FW helium and ITER FW water occurs, injecting helium and water into VV. This pressurisation causes the activation of VV pressure suppressions system and ingress of water into VV. This pressurisation causes the VV pressure suppressions system (VVPSS) to open in an attempt to contain the pressure below the safety limit of 0.2 MPa. This report is intended to do the above accident analysis and assessment of active components of TBM using RELAP code and hence prove its safety in ITER environment. (author)

  14. Analysis of fuel behaviour after loss-of-coolant accident with the TESPA-code

    International Nuclear Information System (INIS)

    Keusenhoff, J.

    1981-01-01

    After a loss-of-coolant accident fuel rods go through a phase of high temperature and differential pressure before quenching and initiation of long term cooling. For licensing purpose the highest cladding temperature and the coolability of the core is of interest. The highest temperature is evaluated by a hot channel calculation with conservative assumptions. It gives little information about the status of the entire core. Therefore more detailed information is necessary. TESPA is a fast running code, which uses best-estimate assumptions, considers statistical uncertainties in the input parameters and calculates clad ballooning and rupture. The code is a usefull tool for calculation of channel blockage and cladding rupture

  15. Revised Mark 22 coolant temperature coefficients

    International Nuclear Information System (INIS)

    Graves, W.E.

    1987-01-01

    Coolant temperature coefficients for the Mark 22 charge published previously are non-conservative because of the neglect of a significant mechanism which has a positive contribution to reactivity. Even after correcting for this effect, dynamic tests made on a Mark VIB charge in the early 60's suggest the results are still non-conservative. This memorandum takes both of these sources of information into account in making a best estimate of the prompt (coolant plus metal) temperature coefficient. Although no safety issues arise from this work (the overall temperature coefficient still strongly contributes to reactor stability), it is obviously desirable to use best estimates for prompt coefficients in limits and other calculations

  16. Recovery studies for plutonium machining oil coolant

    International Nuclear Information System (INIS)

    Navratil, J.D.; Baldwin, C.E.

    1977-01-01

    Lathe coolant oil, contaminated with plutonium and having a carbon tetrachloride diluent, is generated in plutonium machining areas at Rocky Flats. A research program was initiated to determine the nature of plutonium in this mixture of oil and carbon tetrachloride. Appropriate methods then could be developed to remove the plutonium and to recycle the oil and carbon tetrachloride. Studies showed that the mixtures of spent oil and carbon tetrachloride contained particulate plutonium and plutonium species that are soluble in water or in oil and carbon tetrachloride. The particulate plutonium was removed by filtration; the nonfilterable plutonium was removed by adsorption on various materials. Laboratory-scale tests indicated the lathe-coolant oil mixture could be separated by distilling the carbon tetrachloride to yield recyclable products

  17. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1980-01-01

    This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the 13 N content in the containment atmosphere. 13 N is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/ 13 N+ 4 He. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium 13 N concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

  18. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1980-01-01

    This study deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process Hl+016/yields/Nl3+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m/sup -3/ and 7 kBq m/sup -3/ for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge(Li) flow detector assembly operated at elevated pressure. 8 refs

  19. Reactor coolant pressure boundary leakage detection system

    International Nuclear Information System (INIS)

    Dissing, E.; Svansson, L.

    1979-08-01

    The present paper deals with a system for monitoring the leakage of reactor coolant. This system is based primarily on the detection of the N13 content in the containment atmosphere. N13 is produced from the oxygen of the reactor water via the recoil proton nuclear process H1+016 → N13+He4. The generation is therefore independent of fuel element leakage and of the corrosion product content in the water. It is solely related to the neutron flux level in the reactor core. Typical figures for the equilibrium N13 concentration in the containment atmosphere following a 4 kg/minute coolant leakage are 5 kBq m -3 and 7 kBq m -3 for BWR and PWR respectively. These levels are readily measured with a 10 liter Ge (Li) flow detector assembly operated at elevated pressure. (Auth.)

  20. Reactor coolant pump seal leakage monitoring

    International Nuclear Information System (INIS)

    Stevens, D.M.; Spencer, J.W.; Morris, D.J.; James, W.; Shugars, H.G.

    1986-01-01

    Problems with reactor coolant pump seals have historically accounted for a large percentage of unscheduled outages. Studies performed for the Electric Power Research Institute (EPRI) have shown that the replacement of coolant pump seals has been one of the leading causes of nuclear plant unavailability over the last ten years. Failures of coolant pump seals can lead to primary coolant leakage rates of 200-500 gallons per minute into the reactor building. Airborne activity and high surface contamination levels following these failures require a major cleanup effort and increases the time and personnel exposure required to refurbish the pump seals. One of the problems in assessing seal integrity is the inability to accurately measure seal leakage. Because seal leakage flow is normally very small, it cannot be sensed directly with normal flow instrumentation, but must be inferred from several other temperature and flow measurements. In operating plants the leakage rate has been quantified with a tipping-bucket gauge, a device which indicates when one quart of water has been accumulated. The tipping-bucket gauge has been used for most rainfall-intensity monitoring. The need for a more accurate and less expensive gauge has been addressed. They have developed a drop-counter precipitation sensor has been developed and optimized. The applicability of the drop-counter device to the problem of measuring seal leakage is being investigated. If a review of system specification and known drop-counter performance indicates that this method is feasible for measuring seal leak rates, a drop-counter gauge will be fabricated and tested in the laboratory. If laboratory tests are successful the gauge will be demonstrated in a pump test loop at Ontario Hydro and evaluated under simulated plant conditions. 3 references, 2 figures

  1. Enhancing resistance to burnout via coolant chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Tu, J. P.; Dinh, T. N.; Theofanous, T. G. [Univ. of California, Santa Barbara (United States)

    2003-07-01

    Boiling Crisis (BC) on horizontal, upwards-facing copper and steel surfaces under the influence of various coolant chemistries relevant to reactor containment waters is considered. In addition to Boric Acid (BA) and TriSodium Phosphate (TSP), pure De-Ionized Water (DIW) and Tap Water (TW) are included in experiments carried out in the BETA facility. The results are related to a companion paper on the large scale ULPU facility.

  2. Evaluation of effective coolant flow rate in advanced design of the small scale VHTR core

    International Nuclear Information System (INIS)

    Fumizawa, Motoo; Suzuki, Kunihiko; Murakami, Tomoyuki.

    1988-02-01

    This report describes the evaluation of effective coolant flow rate in the advanced design of the small scale VHTR core. The analytical design study was carried out after the 2nd stage of detailed design in order to reduce the cost of construction. The summary of the analytical results are as follows: (1) Crossflow loss coefficient of flange type fuel block having 0.1 mm of sealing gap is about 100 times higher than that of dowel type block adopted in the 2nd stage of detailed design. (2) In case that coolant channel outer diameter is 52 mm and hydraulic diameter is 6 mm, the effective coolant flow rates using flange and dowel type fuel blocks are 80 % and 70 % respectively. Because the crossflow loss coefficients of dowel type are lower than that of flange type. (3) The effective coolant flow rate, when crossflow loss coefficients are distributed along with the axial direction, agrees well with that using mean value of crossflow loss coefficient i.e. 5 x 10 11 m -4 . (author)

  3. High-temperature process heat reactor with solid coolant and radiant heat exchange

    International Nuclear Information System (INIS)

    Alekseev, A.M.; Bulkin, Yu.M.; Vasil'ev, S.I.

    1984-01-01

    The high temperature graphite reactor with the solid coolant in which heat transfer is realized by radiant heat exchange is described. Neutron-physical and thermal-technological features of the reactor are considered. The reactor vessel is made of sheet carbon steel in the form of a sealed rectangular annular box. The moderator is a set of graphite blocks mounted as rows of arched laying Between the moderator rows the solid coolant annular layings made of graphite blocks with high temperature nuclear fuel in the form of coated microparticles are placed. The coolant layings are mounted onto ring movable platforms, the continuous rotation of which is realizod by special electric drives. Each part of the graphite coolant laying consecutively passes through the reactor core neutron cut-off zones and technological zone. In the core the graphite is heated up to the temperature of 1350 deg C sufficient for effective radiant heat transfer. In the neutron cut-off zone the chain reaction and further graphite heating are stopped. In the technological zone the graphite transfers the accumulated heat to the walls of technological channels in which the working medium moves. The described reactor is supposed to be used in nuclear-chemical complex for ammonia production by the method of methane steam catalytic conversion

  4. Minimizing secondary coolant blowdown in HANARO

    International Nuclear Information System (INIS)

    Park, Y. C.; Woo, J. S.; Ryu, J. S.; Cho, Y. G.; Lim, N. Y.

    2000-01-01

    There is about 80m 3 /h loss of the secondary cooling water by evaporation, windage and blowdown during the operation of HANARO, 30MW research reactor. The evaporation and the windage is necessary loss to maintain the performance of cooling tower, but the blowdown is artificial lose to get rid of the foreign material and to maintain the quality of the secondary cooling water. Therefore, minimizing the blowdown loss was studied. It was confirmed, through the relation of the number of cycle and the loss rate of secondary coolant, that the number of cycle is saturated to 12 without blowdown because of the windage loss. When the secondary coolant is treated by high Ca-hardness treatment program (the number of cycle > 10) to maintain the number of cycle around 12 without blowdown, only the turbidity exceeds the limit. By adding filtering system it was confirmed, through the relation of turbidity and filtering rate of secondary cooling water, that the turbidity is reduced below the limit (5 deg.) by 2% of filtering rate without blowdown. And it was verified, through the performance test of back-flow filtering unit, that this unit gets rid of foreign material up to 95% of the back-flow and that the water can be reused as coolant. Therefore, the secondary cooling water can be treated by the high Ca-hardness program and filter system without blowdown

  5. Reactor coolant pumps for nuclear reactors

    International Nuclear Information System (INIS)

    Harand, E.; Richter, G.; Tschoepel, G.

    1975-01-01

    A brake for the pump rotor of a main coolant pump or a shutoff member on the pump are provided in order to prevent excess speeds of the pump rotor. Such excess speeds may occur in PWR type reactors with water at a pressure below, e.g., 150 bars if there is leakage from a coolant line associated with the main coolant pump. As a brake, a centrifugal brake depending upon the pump speed or a brake ring arranged on the pump housing and acting on the pump rotor, which ring would be activated by pressure differentials in the pump, may be used. If the pressure differences between suction and pressure sockets are very small, a controlled hydraulic increase of the pressure force on the brake may also be provided. Furthermore, a turbine brake may be provided. A slide which is automatically movable in closing position along the pump rotor axis is used as a shutoff element. It is of cylindrical configuration and is arranged concentrically with the rotor axis. (DG) [de

  6. Design of automotive engine coolant hoses

    Directory of Open Access Journals (Sweden)

    Hrishikesh D BACHCHHAV

    2018-03-01

    Full Text Available In this paper, we are present the performance of engine coolant hoses (radiator hoses used in passenger cars by checking various physical behaviours such as hose leakage, hose burst, hose collapse or any mechanical damage as studied-thru design guidelines, CFD analysis and product validation testing and also check pressure drop of the hoses when engine will be running. The design term is more likely used for technical part modelling using CAD tool. Later on, we will focus on the transformation of the part design to process design. The process design term is more likely used for "tooling design" for manufacturing of the product using CAD Tool. Then inlet hose carries coolant from engine to radiator inlet tank, then coolant circulated in radiator and passed through radiator outlet tank to water pump of engine with the help of outlet hose. After that …nding any leakage, Burst, damage or collapse of hose and pressure drop of the hose with the help of design checklist, CFD Analysis and product validation testing.

  7. Numerical Investigation on the Performance of an Automotive Thermoelectric Generator with Exhaust-Module-Coolant Direct Contact

    Science.gov (United States)

    Wang, Yiping; Tang, Yulin; Deng, Yadong; Su, Chuqi

    2018-06-01

    Energy conservation and environmental protection have typically been a concern of research. Researchers have confirmed that in automotive engines, just 12-25% of the fuel energy converts into effective work and 30-40% gets wasted in the form of exhaust. Saidur et al. (Energy Policy 37:3650, 2009) and Hasanuzzaman et al. (Energy 36:233, 2011). It will be significant to enhance fuel availability and decrease environmental pollution if the waste heat in the exhaust could be recovered. Thermoelectric generators (TEGs), which can translate heat into electricity, have become a topic of interest for vehicle exhaust waste heat recovery. In conventional automotive TEGs, the thermoelectric modules (TEMs) are arranged between the exhaust tank and the coolant tank. The TEMs do not contact the hot exhaust and coolant, which leads to low heat transfer efficiency. Moreover, to provide enough packing force to keep good contact with the exhaust tank and the coolant tank, the framework required is so robust that the TEGs become too heavy. Therefore, in current study, an automotive TEG was designed which included one exhaust channel, one coolant channel and several TEMs. In the TEG, the TEMs which contacted the exhaust and coolant directly were inserted into the walls of each coolant channel. To evaluate the performance of the automotive TEG, the flow field and temperature field were computed by computational fluid dynamics (CFD). Based on the temperature distribution obtained by CFD and the performance parameters of the modules, the total power generation was obtained by some proved empirical formulas. Compared with conventional automotive TEGs, the power generation per unit volume exhaust was boosted.

  8. Numerical Investigation on the Performance of an Automotive Thermoelectric Generator with Exhaust-Module-Coolant Direct Contact

    Science.gov (United States)

    Wang, Yiping; Tang, Yulin; Deng, Yadong; Su, Chuqi

    2017-12-01

    Energy conservation and environmental protection have typically been a concern of research. Researchers have confirmed that in automotive engines, just 12-25% of the fuel energy converts into effective work and 30-40% gets wasted in the form of exhaust. Saidur et al. (Energy Policy 37:3650, 2009) and Hasanuzzaman et al. (Energy 36:233, 2011). It will be significant to enhance fuel availability and decrease environmental pollution if the waste heat in the exhaust could be recovered. Thermoelectric generators (TEGs), which can translate heat into electricity, have become a topic of interest for vehicle exhaust waste heat recovery. In conventional automotive TEGs, the thermoelectric modules (TEMs) are arranged between the exhaust tank and the coolant tank. The TEMs do not contact the hot exhaust and coolant, which leads to low heat transfer efficiency. Moreover, to provide enough packing force to keep good contact with the exhaust tank and the coolant tank, the framework required is so robust that the TEGs become too heavy. Therefore, in current study, an automotive TEG was designed which included one exhaust channel, one coolant channel and several TEMs. In the TEG, the TEMs which contacted the exhaust and coolant directly were inserted into the walls of each coolant channel. To evaluate the performance of the automotive TEG, the flow field and temperature field were computed by computational fluid dynamics (CFD). Based on the temperature distribution obtained by CFD and the performance parameters of the modules, the total power generation was obtained by some proved empirical formulas. Compared with conventional automotive TEGs, the power generation per unit volume exhaust was boosted.

  9. X-ray microbeam measurements of long-range internal stresses in commercial-purity aluminum processed by multiple passes of equal-channel angular pressing

    International Nuclear Information System (INIS)

    Phan, Thien Q.; Lee, I-Fang; Levine, Lyle E.; Tischler, Jonathan Z.; Huang, Yi; Fox, Alan G.; Langdon, Terence G.; Kassner, Michael E.

    2014-01-01

    X-ray microbeam diffraction was used to measure long-range internal stresses (LRISs) in the grain/subgrain interiors of commercial-purity aluminum processed by equal-channel angular pressing for up to eight passes. The LRIS values at +4.9° off the axial (pressing) direction show only a slight increase with increasing numbers of passes. The normalized stress remains approximately constant at ∼0.10 of the flow stress

  10. Qualification of MHD effects in dual-coolant DEMO blanket and approaches to their modelling

    International Nuclear Information System (INIS)

    Mas de les Valls, E.; Batet, L.; Medina, V. de; Fradera, J.; Sedano, L.A.

    2011-01-01

    Design refinements of vertical insulated banana-shaped liquid metal channels are being considered as a progress of conceptual design of dual-coolant liquid metal blankets (DEMO specifications). Among them: (a) optimised channel geometry and (b) improvements on flow channel inserts. Progress of channel conceptual design is conducted in parallel with underlying physics of MHD models in diverse aspects: (1) MHD models, (2) MHD turbulence, (3) LM buoyancy effects, (4) three-dimensional flows, and (5) LM/FCI/wall electrical and thermal coupling; in order to progress on common liquid metal flow characterisation, pressure drop and three-dimensional flows. The analyses are assumed as extension of those previous carried out for the DCLL blankets for new design refinements. At the present stage of the conceptual design progress, a preliminary thermofluid MHD study is of crucial interest for further design improvements and future detailed modelling. The paper overviews the ongoing modelling studies, making model refinements explicit, and anticipates some modelling results.

  11. Supercritical heat transfer in an annular channel with two-sided heaing

    International Nuclear Information System (INIS)

    Sergeev, V.V.; Remizov, O.V.; Gal'chenko, Eh.F.

    1986-01-01

    The paper deals with experimental inestigation into worsening of heat transfer at forced up flow in steam-water mixture in a vertical annular channel with two-sided heating and development of technique for calculation of supercritical heat exchange in this channel. Bench-scale experiments are carried out at high-pressure at mass rates of the coolant equal to 300-865 kg/(m 2 x s), pressure of 9.8-17.8 MPa and heat flux on the internal surface - 20-400 kW/m 2 , on the external surface - 35-450 kW/m 2 . Technique for calculation of supercritical heat exchange in channels with one- and two-sided heating is suggested. Analysis of the obtained experimental data permits to determine conditions for arising departure nucleate boiling on the internal and external surfaces and on both surfaces simultaneously. It is concluded that the suggested technique of calculation adequately reflects the effect of regime parameters of coolant flow on temperature regime of heat transferring surfaces in the supercritical area

  12. A conceptual design of the ITER upper port plug structure and its cooling channels

    International Nuclear Information System (INIS)

    Pak, S.I.; Lee, H.G.; Jung, K.J.; Walker, C.I.; Kim, D.G.; Choi, K.S.

    2008-01-01

    A study is conducted on the conceptual design of the structure and cooling channels of the upper port plug of International Thermonuclear Experimental Reactor (ITER). Modification of the earlier port plug design is made and a simple fabrication method is proposed. It is shown that the newly designed port plug can accommodate the installation of both diagnostic and electron cyclotron heating (ECH) devices. Design assessment is carried out through structural and thermo-hydraulic analyses. Results of the analyses show that the port plug structure is stable against one of the most severe plasma events and the total pressure drop of the coolant is within the allowable level

  13. Development of Coolant Radioactivity Interpretation Code

    International Nuclear Information System (INIS)

    Kim, Kiyoung; Jung, Youngsuk; Kim, Kyounghyun; Kim, Jangwook

    2013-01-01

    In Korea, the coolant radioactivity analysis has been performed by using the computer codes of foreign companies such as CADE (Westinghouse), IODYNE and CESIUM (ABB-CE). However, these computer codes are too conservative and have involved considerable errors. Furthermore, since these codes are DOS-based program, their easy operability is not satisfactory. Therefore it is required development of an enhanced analysis algorithm applying an analytical method reflecting the change of operational environments of domestic nuclear power plants and a fuel failure evaluation software considering user' conveniences. We have developed a nuclear fuel failure evaluation code able to estimate the number of failed fuel rods and the burn-up of failed fuels during nuclear power plant operation cycle. A Coolant Radio-activity Interpretation Code (CRIC) for LWR has been developed as the output of the project 'Development of Fuel Reliability Enhanced Technique' organized by Korea Institute of Energy Technology Evaluation and Planning (KETEP). The CRIC is Windows based-software able to evaluate the number of failed fuel rods and the burn-up of failed fuel region by analyzing coolant radioactivity of LWR in operation. The CRIC is based on the model of fission products release commonly known as 'three region model' (pellet region, gap region, and coolant region), and we are verifying the CRIC results based on the cases of domestic fuel failures. CRIC users are able to estimate the number of failed fuel rods, burn-up and regions of failed fuel considered enrichment and power distribution of fuel region by using operational cycle data, coolant activity data, fuel loading pattern, Cs-134/Cs-137 ratio according to burn-up and U-235 enrichment provided in the code. Due to development of the CRIC, it is secured own unique fuel failure evaluation code. And, it is expected to have the following significant meaning. This is that the code reflecting a proprietary technique for quantitatively

  14. Investigation of coolant thermal mixing within 28-element CANDU fuel bundles using the ASSERT-PV thermal hydraulics code

    International Nuclear Information System (INIS)

    Lightston, M.F.; Rock, R.

    1996-01-01

    This paper presents the results of a study of the thermal mixing of single-phase coolant in 28-element CANDU fuel bundles under steady-state conditions. The study, which is based on simulations performed using the ASSERT-PV thermal hydraulic code, consists of two main parts. In the first part the various physical mechanisms that contribute to coolant mixing are identified and their impact is isolated via ASSERT-PV simulations. The second part is concerned with development of a preliminary model suitable for use in the fuel and fuel channel code FACTAR to predict the thermal mixing that occurs between flow annuli. (author)

  15. Boundary vapor contentsin an annular channel

    International Nuclear Information System (INIS)

    Remizov, O.V.; Shurkin, N.G.; Podgornyj, K.K.; Gal'chenko, Eh.F.; Bukhteev, I.S.

    1978-01-01

    The work is aimed at the experimental investigation of the worsening of the heat transfer in an annular channel. The experiments have been carried out on the annular channel 32x28x3000 mm with the even distribution of the heat flux along the length at pressures of 6.9-19.6 MPa, flow rate of 350-1000 kg/m 2 s, and specific heat fluxes from 0.18 up to 0.6 MW/m 2 . Heating is external, oneside. Water monodistillate of the following composition has been used as a coolant: pH 9; dry residue - 0.8-1.2 mg/kg, oxygen -10-15 mg/kg. It is found out that the change character of the temperature field of the heating surface of the annular channel at the regime with the worsen of heat emission depends on the ratio of regime parameters. At pressures of 6.9-13.7 MPa and flow rate of 350-500 kg/m 2 s the channel wall temperature rises monotoneously, never reaching its maximum. With pressure rise > 13.7 MPa and mass velocity > 500 kg/m 2 s the temperature of the heat emitting surface reaches its maximum, and then slowly falls. At pressures of 6.9-11.8 MPa the boundary vapor content value within the whole range of mass velocities does not depend on the specific heat flux q. At pressures higher than 13.7 MPa and mass velocities of 350-1000 kg/m 2 s the boundary vapor content depends on q. The heating of the external or internal surface of the annular channel affects the value of the boundary vapor content within the whole range of regime parameters' change under investigation

  16. Water quality control device and water quality control method for reactor primary coolant system

    International Nuclear Information System (INIS)

    Wada, Yoichi; Ibe, Eishi; Watanabe, Atsushi.

    1995-01-01

    The present invention is suitable for preventing defects due to corrosion of structural materials in a primary coolant system of a BWR type reactor. Namely, a concentration measuring means measures the concentration of oxidative ingredients contained in a reactor water. A reducing electrode is disposed along a reactor water flow channel in the primary coolant system and reduces the oxidative ingredients. A reducing counter electrode is disposed along the reactor water flow channel in the primary coolant system, and electrically connected to the reducing electrode. The reactor structural materials are used as a reference electrode providing a reference potential to the reducing electrode and the reducing counter electrode. A potential control means controls the potential of the reducing electrode relative to the reference potential based on the signals from the concentration measuring means. A stable reference potential in a region where an effective oxygen concentration is stable can be obtained irrespective of the change of operation conditions by using the reactor structural materials disposed to a boiling region in the reactor core as a reference electrode. As a result, the water quality can be controlled at high accuracy. (I.S.)

  17. New method for NPP sodium coolant pipeline austenization

    International Nuclear Information System (INIS)

    Malashonok, V.A.; Rotshtejn, A.V.; Gotshalk, A.L.; Miryushchenko, E.F.

    1980-01-01

    Heat treatment technology is considered for pipelines intended for the NPP cooling systems employing sodium coolant. Various techniques are discussed which are used for protecting the pipeline internal surfaces against oxidation in the process of heat treatment. It is noted that the austenite formation of welded joints of steel 12Kh18N9 and steel Kh16N11M3 at temperatures of 1050 and 1100 deg C releases welding-induced stresses and reduces a possibility of local damages. Evacuation down to 1 mm Hg appears to be the most rational protective technique. The considered procedure of the pipeline heat treatment has been utilized for mounting the equipment of the BN-600 reactor at the Beloyarskaya NPP. The economic gain resulting from the use of the procedure, owing to decrease in argon consumption and reduction of labour input, makes up 150 000 roubles

  18. On-line monitoring of main coolant pump seals

    International Nuclear Information System (INIS)

    Stevens, D.M.; Spencer, J.W.; Morris, D.J.; Glass, S.W.; Sommerfield, G.A.; Harrison, D.

    1984-06-01

    The Babcock and Wilcox Company has developed and implemented a Reactor Coolant Pump Monitoring and Diagnostic System (RCPM and DS). The system has been installed at Toledo Edison Company's Davis-Besse Nuclear Power Station Unit 1. The RCPM and PS continuously monitors a number of indicators of pump performance and notifies the plant operator of out-of-tolerance conditions or pump performance trending toward out-of-tolerance conditions. Pump seal parameters being monitored include pump internal pressures, temperatures, and flow rates. Rotordynamic performanvce and plant operating conditions are also measured with a variety of dynamic sensors. This paper describes the implementation of the system and the results of on-line monitoring of four RC pumps

  19. Fuel-coolant interactions: preliminary experiments on the effect of gases dissolved in the 'coolant'

    International Nuclear Information System (INIS)

    Asher, R.C.; Davies, D.; Jones, P.G.

    1976-12-01

    A simple apparatus has been used to study fuel-coolant interactions under reasonably well controlled conditions. Preliminary experiments have used water as the 'coolant' and molten tin at 800 0 C as the 'fuel' and have investigated how the violence of the interaction is affected by dissolving gases (oxygen, nitrogen, carbon dioxide and nitrous oxide) in the water. It was found that saturating the water with carbon dioxide or nitrous oxide completely suppresses the violent interaction. Experiments in which the concentrations of these gases were varied showed that a certain critical concentration was needed; below this concentration the dissolved gas has no significant effect but above it the suppression is

  20. Early detection of coolant boiling in research reactors with MTR-type fuel

    International Nuclear Information System (INIS)

    Kozma, R.; Turkcan, E.; Verhoef, J.P.

    1992-10-01

    A reactor core monitoring system having the function of early detection of boiling in the coolant channels of research reactors with MTR-type fuel is introduced. The system is based on the on-line analysis of signals of various ex-core and in-core neutron detectors. Early detection of coolant boiling cannot be accomplished by the evaluation of the DC components of these detectors in a number of practically important cases of boiling anomaly. It is shown that the noise component of the available neutron detector signals can be used for the detection of boiling in these cases. Experiments have been carried out at a boiling setup in the research reactor HOR of the Interfaculty Reactor Institute, Technical University of Delft, The Netherlands. (author). 8 refs., 11 figs

  1. The effect of coolants on the performance of magnetic micro-refrigerators.

    Science.gov (United States)

    Silva, D J; Bordalo, B D; Pereira, A M; Ventura, J; Oliveira, J C R E; Araújo, J P

    2014-06-01

    Magnetic refrigeration is an alternative cooling technique with envisaged technological applications on micro- and opto-electronic devices. Here, we present a magnetic micro-refrigerator cooling device with embedded micro-channels and based on the magnetocaloric effect. We studied the influence of the coolant fluid in the refrigeration process by numerically simulating the heat transfer processes using the finite element method. This allowed us to calculate the cooling power of the device. Our results show that gallium is the most efficient coolant fluid and, when used with Gd5Si2Ge2, a maximum power of 11.2 W/mm3 at a working frequency of -5 kHz can be reached. However, for operation frequencies around 50 Hz, water is the most efficient fluid with a cooling power of 0.137 W/mm3.

  2. EDF PWRs primary coolant purification strategies

    International Nuclear Information System (INIS)

    Gressier, Frederic; Mascarenhas, Darren; Taunier, Stephane; Le-Calvar, Marc; Bretelle, Jean-Luc; Ranchoux, Gilles

    2012-09-01

    In order to achieve a good physico-chemical quality of the primary coolant fluid, the primary water is continuously treated by the Chemical and Volume Control System (CVCS). This system is composed of a treatment chain containing filters and ion-exchange resins. In the EDF design, an upstream filter is placed before the resin so as to prevent it from being saturated with insoluble particles. Then, the fluid passes through several resin beds (up to 3 depending on the configuration) and again through a downstream filter that prevents resin fines dissemination into the reactor coolant. Much work has been conducted in the last 5 years on the homogenisation of products and usage on French EDF NPP primary coolant treatment, while taking into account the compromise between source term reduction, liquid and solid waste, and buying and disposal costs. Two national markets have been created, and two operational documents for chemists on site have been published: a filtration guideline and an ion-exchange resin guideline. Both documents give general information about the products used, how are they characterized and selected for national market (technical requirements, standards and tests), how they should be used and what are the change-out criteria. They are also periodically updated based on feedback from sites. The positive impact on resin and filter lifetime (extension of some, limitation of others), homogenisation of products and usage will be presented. Moreover, EDF is constantly in the process of improving the current purification methods, as well as researching the use of existing and novel technologies. In this field, recent experiments on short loading of resin during reactor shutdown has been tested on site with success. In addition, work is done on silica free filters, filter consumption and filter chemical release. An overview of these optimization methods will be given. (authors)

  3. Coolant degassing device for PWR type reactors

    International Nuclear Information System (INIS)

    Kita, Kaoru; Takezawa, Kazuaki; Minemoto, Masaki.

    1982-01-01

    Purpose: To efficiently decrease the rare gas concentration in primary coolants, as well as shorten the degassing time required for the periodical inspection in the waste gas processing system of a PWR type reactor. Constitution: Usual degassing method by supplying hydrogen or nitrogen to a volume control tank is replaced with a method of utilizing a degassing tower (method of flowing down processing liquid into the filled tower from above while uprising streams from the bottom of the tower thereby degassing the gases dissolved in the liquid into the steams). The degassing tower is combined with a hydrogen separator or hydrogen recombiner to constitute a waste gas processing system. (Ikeda, J.)

  4. Microstructural characterization of primary coolant pipe steel

    International Nuclear Information System (INIS)

    Miller, M.K.; Bentley, J.

    1986-01-01

    Atom probe field-ion microscopy, analytical electron microscopy, and optical microscopy have been used to investigate the changes that occur in the microstructure of cast CF 8 primary coolant pipe stainless steel after long term thermal aging. The cast duplex microstructure consisted of austenite with 15% delta-ferrite. Investigation of the aged material revealed that the ferrite spinodally decomposed into a fine scaled network of α and α'. A fine G-phase precipitate was also observed in the ferrite. The observed degradation in mechanical properties is probably a consequence of the spinodal decomposition in the ferrite

  5. Calorimetric and reactor coolant system flow uncertainty

    International Nuclear Information System (INIS)

    Bates, L.; McLean, T.

    1991-01-01

    This paper describes a methodology for the quantification of errors associated with the determination of a feedwater flow, secondary power, and Reactor Coolant System (RCS) flow used at the Trojan Nuclear Plant to ensure compliance with regulatory requirements. The sources of error in Plant indications and process measurement are identified and tracked, using examples, through the mathematical processes necessary to calculate the uncertainty in the RCS flow measurement. An error of approximately 1.4 percent is calculated for secondary power. This error results, along with the consideration of other errors, in an uncertainty of approximately 3 percent in the RCS flow determination

  6. Reactor auxiliary cooling facility and coolant supplying method therefor

    Energy Technology Data Exchange (ETDEWEB)

    Ando, Koji; Kinoshita, Shoichiro

    1996-06-07

    A reactor auxiliary cooling facility of the present invention comprises a coolant recycling line for recycling coolants by way of a reactor auxiliary coolant pump and a cooling load, a gravitational surge tank for supplying coolants to the coolant recycling line and a supplemental water supplying line for supplying a supply the supplemental water to the tank. Then, a pressurization-type supply water surge tank is disposed for operating the coolant recycling line upon performing an initial system performance test in parallel with the gravitational surge tank. With such a constitution, the period of time required from the start of the installation of reactor auxiliary cooling facilities to the completion of the system performance test can be shortened at a reduced cost without enlarging the scale of the facility. (T.M.)

  7. Reactor auxiliary cooling facility and coolant supplying method therefor

    International Nuclear Information System (INIS)

    Ando, Koji; Kinoshita, Shoichiro.

    1996-01-01

    A reactor auxiliary cooling facility of the present invention comprises a coolant recycling line for recycling coolants by way of a reactor auxiliary coolant pump and a cooling load, a gravitational surge tank for supplying coolants to the coolant recycling line and a supplemental water supplying line for supplying a supply the supplemental water to the tank. Then, a pressurization-type supply water surge tank is disposed for operating the coolant recycling line upon performing an initial system performance test in parallel with the gravitational surge tank. With such a constitution, the period of time required from the start of the installation of reactor auxiliary cooling facilities to the completion of the system performance test can be shortened at a reduced cost without enlarging the scale of the facility. (T.M.)

  8. Effect of cross-flow direction of coolant on film cooling effectiveness with one inlet and double outlet hole injection

    Directory of Open Access Journals (Sweden)

    Guangchao Li

    2012-12-01

    Full Text Available In order to study the effect of cross-flow directions of an internal coolant on film cooling performance, the discharge coefficients and film cooling effectiveness with one inlet and double outlet hole injections were simulated. The numerical results show that two different cross-flow directions of the coolant cause the same decrease in the discharge coefficients as that in the case of supplying coolant by a plenum. The different proportion of the mass flow out of the two outlets of the film hole results in different values of the film cooling effectiveness for three different cases of coolant supplies. The film cooling effectiveness is the highest for the case of supplying coolant by the plenum. At a lower blowing ratio of 1.0, the film cooling effectiveness with coolant injection from the right entrance of the passage is higher than that from the left entrance of the passage. At a higher blowing ratio of 2.0, the opposite result is found.

  9. Effect of parameter variation of reactor coolant pump on loss of coolant accident consequence

    International Nuclear Information System (INIS)

    Dang Gaojian; Huang Daishun; Gao Yingxian; He Xiaoqiang

    2015-01-01

    In this paper, the analyses were carried out on Ling'ao nuclear power station phase II to study the consequence of the loss of coolant accident when the homologous characteristic curves and free volumes of the reactor coolant pump changed. Two different pumps used in the analysis were 100D (employed on Ling'ao nuclear power station phase II) and ANDRITZ. The thermal characteristics in the large break LOCA accident were analyzed using CATHRE GB and CONPATE4, and the reactor coolant system hydraulics load during blow-clown phase of LOCA accident was analyzed using ATHIS and FORCET. The calculated results show that the homologous characteristic curves have great effect on the thermal characteristics of reactor core during the reflood phase of the large break LOCA accident. The maximum cladding surface temperatures are quite different when the pump's homologous characteristic curves change. On the other hand, the pump's free volume changing results in the variation of the LOCA rarefaction wave propagation, and therefore, the reactor coolant system hydraulic load in LOCA accident would be different. (authors)

  10. Predicted Variations of Water Chemistry in the Primary Coolant Circuit of a Supercritical Water Reactor

    International Nuclear Information System (INIS)

    Yeh, Tsung-Kuang; Wang, Mei-Ya; Liu, Hong-Ming; Lee, Min

    2012-09-01

    In response to the demand over a higher efficiency for a nuclear power plant, various types of Generation IV nuclear reactors have been proposed. One of the new generation reactors adopts supercritical light water as the reactor coolant. While current in-service light water reactors (LWRs) bear an average thermal efficiency of 33%, the thermal efficiency of a supercritical water reactor (SCWR) could generally reach more than 44%. For LWRs, the coolants are oxidizing due to the presence of hydrogen peroxide and oxygen, and the degradation of structural materials has mainly resulted from stress corrosion cracking. Since oxygen is completely soluble in supercritical water, similar or even worse degradation phenomena are expected to appear in the structural and core components of an SCWR. To ensure proper designs of the structural components and suitable selections of the materials to meet the requirements of operation safety, it would be of great importance for the design engineers of an SCWR to be fully aware of the state of water chemistry in the primary coolant circuit (PCC). Since SCWRs are still in the stage of conceptual design and no practical data are available, a computer model was therefore developed for analyzing water chemistry variation and corrosion behavior of metallic materials in the PCC of a conceptual SCWR. In this study, a U.S. designed SCWR with a rated thermal power of 3575 MW and a coolant flow rate of 1843 kg/s was selected for investigating the variations in redox species concentration in the PCC. Our analyses indicated that the [H 2 ] and [H 2 O 2 ] at the core channel were higher than those at the other regions in the PCC of this SCWR. Due to the self-decomposition of H 2 O 2 , the core channel exhibited a lower [O 2 ] than the upper plenum. Because the middle water rod region was in parallel with the core channel region with relatively high dose rates, the [H 2 ] and [H 2 O 2 ] in this region were higher than those in the other regions

  11. Improving Income Positions of Primary Producers in International Marketing Channels: The Lake Victoria-EU Nile Perch Case

    NARCIS (Netherlands)

    Kambewa, E.; Ingenbleek, P.T.M.; Tilburg, van A.

    2008-01-01

    Although fair distribution of incomes within marketing channels and systems receives increasing attention in companies¿ corporate social responsibility policies, the marketing literature offers few insights that may be helpful to initiate projects that improve incomes of primary producers in

  12. Natural circulation in reactor coolant system

    International Nuclear Information System (INIS)

    Han, J.T.

    1987-01-01

    Reactor coolant system (RCS) natural circulation in a PWR is the buoyancy-driven coolant circulation between the core and the upper-plenum region (in-vessel circulation) with or without a countercurrent flow in the hot leg piping between the vessel and steam generators (ex-vessel circulation). This kind of multidimensional bouyancy-driven flow circulation serves as a means of transferring the heat from the core to the structures in the upper plenum, hot legs, and possibly steam generators. As a result, the RCS piping and other pressure boundaries may be heated to high temperatures at which the structural integrity is challenged. RCS natural circulation is likely to occur during the core uncovery period of the TMLB' accident in a PWR when the vessel upper plenum and hot leg are already drained and filled with steam and possibly other gaseous species. RCS natural circulation is being studied for the Surry plant during the TMLB' accident in which station blackout coincides with the loss of auxiliary feedwater and no operator actions. The effects of the multidimensional RCS natural circulation during the TMLB' accident are discussed

  13. Efficiency of water coolant for DEMO divertor

    International Nuclear Information System (INIS)

    Fetzer, Renate; Igitkhanov, Yuri; Bazylev, Boris

    2015-01-01

    Up to now, water-cooled divertor concepts have been developed for limited incident fluxes without taking into account transient power loadings. In this paper we analyzed the efficiency of water as a coolant for the particular PFC tungsten monoblock shield with a cooling tube made from Cu alloy (Cu OFHC) as a laminate adjacent to W and a low activation martensitic steel (Eurofer) as inner tube contacting the coolant. Thermal analysis is carried out by using the code MEMOS, which simulates W armour damage under the repetitive ELM heat loads. We consider cooling conditions which allow one to keep relatively high material temperatures (in the range 300–600 °C) thus minimizing Eurofer embrittlement under neutron irradiation. Expected DEMO I and DEMO II heat loads including type I ELMs are found to cause melting of the W surface during unmitigated ELMs. By mitigation of the ELMs melting of W is avoided. DEMO I operation under these conditions is save for cooling at water pressure 15.5 MPa and temperature 325 °C, while for DEMO II with mitigated ELMs the critical heat flux is exceeded and safe operation is not provided.

  14. Chemistry of liquid metal coolants and sensors

    International Nuclear Information System (INIS)

    Gnanasekaran, T.

    2015-01-01

    Liquid sodium is the coolant of choice for the current generation fast breeder reactors. When sodium contains low levels of dissolved non-metallic impurities, it is highly compatible with structural steels. When the dissolved oxygen level is high, corrosion and mass transfer in sodium-steel circuits are enhanced and this involves formation of NaxMyOz type of species (M = alloying components in steels). Experience has shown that this enhancement of corrosion in a sodium circuit with all austenitic steel structural materials would not be encountered if oxygen level in sodium is below ~ 5ppm. For understanding this observation, a complete knowledge on the phase diagrams of Na-M-O systems and the thermochemical data of all relevant NaxMyOz compounds is essential. This presentation would highlight the work carried out at IGCAR on the chemistry of liquid sodium and heavy liquid metal coolants. Work carried out on various sensors for their use in these liquid metal circuits would be described and their current status would be discussed

  15. Efficiency of water coolant for DEMO divertor

    Energy Technology Data Exchange (ETDEWEB)

    Fetzer, Renate, E-mail: renate.fetzer@kit.edu; Igitkhanov, Yuri; Bazylev, Boris

    2015-10-15

    Up to now, water-cooled divertor concepts have been developed for limited incident fluxes without taking into account transient power loadings. In this paper we analyzed the efficiency of water as a coolant for the particular PFC tungsten monoblock shield with a cooling tube made from Cu alloy (Cu OFHC) as a laminate adjacent to W and a low activation martensitic steel (Eurofer) as inner tube contacting the coolant. Thermal analysis is carried out by using the code MEMOS, which simulates W armour damage under the repetitive ELM heat loads. We consider cooling conditions which allow one to keep relatively high material temperatures (in the range 300–600 °C) thus minimizing Eurofer embrittlement under neutron irradiation. Expected DEMO I and DEMO II heat loads including type I ELMs are found to cause melting of the W surface during unmitigated ELMs. By mitigation of the ELMs melting of W is avoided. DEMO I operation under these conditions is save for cooling at water pressure 15.5 MPa and temperature 325 °C, while for DEMO II with mitigated ELMs the critical heat flux is exceeded and safe operation is not provided.

  16. Requirements of coolants in nuclear reactors

    International Nuclear Information System (INIS)

    Abass, O. A. M.

    2014-11-01

    This study discussed the purposes and types of coolants in nuclear reactors to generate electricity. The major systems and components associated with nuclear reactors are cooling system. There are two major cooling systems utilized to convert the heat generated in the fuel into electrical power. The primary system transfers the heat from the fuel to the steam generator, where the secondary system begins. The steam formed in the steam generator is transferred by the secondary system to the main turbine generator, where it s converted into electricity after passing through the low pressure turbine. There are various coolants used in nuclear reactors-light water, heavy water and liquid metal. The two major types of water-cooled reactors are pressurized water reactors (PWR) and boiling water reactors (BWR) but pressurized water reactors are more in the world. Also discusses this study the reactors and impact of the major nuclear accidents, in the April 1986 disaster at the Chernobyl nuclear power plant in Ukraine was the product operators, and in the March 2011 at the Fukushima nuclear power plant in Japan was the product of earthquake of magnitude 9.0, the accidents caused the largest uncontrolled radioactive release into the environment.(Author)

  17. Computational and Experimental Investigations of the Coolant Flow in the Cassette Fissile Core of a KLT-40S Reactor

    Science.gov (United States)

    Dmitriev, S. M.; Varentsov, A. V.; Dobrov, A. A.; Doronkov, D. V.; Pronin, A. N.; Sorokin, V. D.; Khrobostov, A. E.

    2017-07-01

    Results of experimental investigations of the local hydrodynamic and mass-exchange characteristics of a coolant flowing through the cells in the characteristic zones of a fuel assembly of a KLT-40S reactor plant downstream of a plate-type spacer grid by the method of diffusion of a gas tracer in the coolant flow with measurement of its velocity by a five-channel pneumometric probe are presented. An analysis of the concentration distribution of the tracer in the coolant flow downstream of a plate-type spacer grid in the fuel assembly of the KLT-40S reactor plant and its velocity field made it possible to obtain a detailed pattern of this flow and to determine its main mechanisms and features. Results of measurement of the hydraulic-resistance coefficient of a plate-type spacer grid depending on the Reynolds number are presented. On the basis of the experimental data obtained, recommendations for improvement of the method of calculating the flow rate of a coolant in the cells of the fissile core of a KLT-40S reactor were developed. The results of investigations of the local hydrodynamic and mass-exchange characteristics of the coolant flow in the fuel assembly of the KLT-40S reactor plant were accepted for estimating the thermal and technical reliability of the fissile cores of KLT-40S reactors and were included in the database for verification of computational hydrodynamics programs (CFD codes).

  18. Coolant make-up device for BWR type reactor

    International Nuclear Information System (INIS)

    Sasagawa, Hiroshi.

    1994-01-01

    In a coolant make-up device, an opening of a pressure equalizing pipeline in a pressure vessel is disposed in coolants above a reactor core and below a usual fluctuation range of a reactor vessel water level. Further, a float check valve is disposed to the pressure equalizing pipeline for preventing coolants in the pressure vessel flowing into the pipeline. If the water level in the pressure vessel is lowered than the setting position for the float check valve, the float drops by its own weight to open the opening of the pressure equalizing pipeline. Then, steams in the pressure vessel are flown into the pipeline, to equalize the pressure between a coolant storage tank and the pressure vessel of the reactor. Coolants in the coolant storage tank is injected to the pressure vessel by way of the water injection pipeline due to the difference of the pressure head between the water level in the coolants storage tank and the water level in the pressure vessel. If the coolants are lowered than the setting position for the float check value, the float check valve does not close unless the water level is recovered to the setting position for the float valve and, accordingly, the coolant make-up is continued. (N.H.)

  19. Core dynamics analysis for reactivity insertion and loss of coolant flow tests using the HTTR

    International Nuclear Information System (INIS)

    Takamatsu, Kuniyoshi; Nakagawa, Shigeaki; Takeda, Tetsuaki

    2007-01-01

    The High Temperature engineering Test Reactor (HTTR) is a graphite-moderated and a gas-cooled reactor with a thermal power of 30 MW and a reactor outlet coolant temperature of 950degC (SAITO, 1994). Safety demonstration tests using the HTTR are in progress to verify its inherent safety features and improve the safety technology and design methodology for High-Temperature Gas-cooled Reactors (HTGRs) (TACHIBANA 2002) (NAKAGAWA 2004). The reactivity insertion test is one of the safety demonstration tests for the HTTR. This test simulates the rapid increase in the reactor power by withdrawing the control rod without operating the reactor power control system. In addition, the loss of coolant flow tests has been conducted to simulate the rapid decrease in the reactor power by tripping one, two or all out of three gas circulators. The experimental results have revealed the inherent safety features of HTGRs, such as the negative reactivity feedback effect. The numerical analysis code, which was named ACCORD (TAKAMATSU 2006), was developed to analyze the reactor dynamics including the flow behavior in the HTTR core. We used a conventional method, namely, a one-dimensional flow channel model and reactor kinetics model with a single temperature coefficient, taking into account the temperature changes in the core. However, a slight difference between the analytical and experimental results was observed. Therefore, we have modified this code to use a model with four parallel channels and twenty temperature coefficients in the core. Furthermore, we added another analytical model of the core for calculating the heat conduction between the fuel channels and the core in the case of the loss of coolant flow tests. This paper describes the validation results for the newly developed code using the experimental results of the reactivity insertion test as well as the loss of coolant flow tests by tripping one or two out of three gas circulators. Finally, the pre-analytical result of

  20. System for mitigating consequences of loss of coolant accident at nuclear power station

    International Nuclear Information System (INIS)

    Bukrinsky, A.M.; Rzheznikov, J.V.; Shvyryaev, J.V.; Zlatin, D.A.; Kuznetsov, J.A.; Babenko, E.A.; Tatarnikov, V.P.; Lapshin, A.L.; Sanovich, V.I.

    1981-01-01

    The system according to the invention comprises a first room which accommodates a reactor plant and an active-type sprinkler means. As pressure rises in the first room due to a release of steam from the lost coolant, most of the air contained in this first room is driven out through holes provided in walls of the first room in immediate proximity to a floor of the first room, wherefrom it proceeds to a second room through channels and a basin-type condenser accommodated in the second room. The length of the channels is selected so as to form a water seal in these channels to prevent the back-flow of air from the second room to the first room and thus produce rarefaction in the first room. (author)

  1. Statistical analysis of the vibration loading of the reactor internals and fuel assemblies of reactor units type WWER-440 from deferent projects

    International Nuclear Information System (INIS)

    Ovcharov, O.; Pavelko, V.; Usanov, A.; Arkadov, G.; Dolgov, A.; Molchanov, V.; Anikeev, J.; Pljush, A.

    2006-01-01

    In this paper the following items have been presented: 1) Vibration noise instrument channels; 2) Vibration loading characteristics of control assemblies, internals and design peculiarities of internals of WWER-440 deferent projects; 3) Coolant flow rate through the reactor, reactor core, fuel assemblies and control assemblies for different projects WWER-440 and 4) Noise measurements of coolant speed per channel. The change of auto power spectrum density of absolute displacement detector signal for the last 12 years of SUS monitoring of the Kola NPP unit 2; the coherence functions groups between two SPND of the same level for the Kola NPP unit 1; the measured coolant flow rate at Paks NPP and the auto power spectrum density group of SPND signals from 11 neutron measuring channels of the Kola NPP unit 1 are given. The main factors of vibration loading of internals and fuel assemblies for Kola NPP units 1-4, Bohunice NPP units 1 and 2 and Novovoronezh NPP units 3 and 4 are also discussed

  2. Numerical investigation of the thermal and electrical performances for combined solar photovoltaic/thermal (PV/T) modules based on internally extruded fin flow channel

    Science.gov (United States)

    Deng, Y. C.; Li, Q. P.; Wang, G. J.

    2017-11-01

    A solar photovoltaic/thermal (PV/T) module based on internally extruded fin flow channel was investigated numerically in this paper. First of all, the structures of the thin plate heat exchanger and the PV/T module were presented. Then, a numerical model of the PV/T module considering solar irradiation, fluid flow and heat transfer was developed to analyze the performance of the module. Finally, the steady electrical and thermal efficiencies of the PV/T module at different inlet water temperatures and mass flow rates were achieved. These numerical results supply theory basis for practical application of the PV/T module.

  3. 14C Behaviour in PWR coolant

    International Nuclear Information System (INIS)

    Sims, Howard; Dickinson Shirley; Garbett, Keith

    2012-09-01

    Although 14 C is produced in relatively small amounts in PWR coolant, it is important to know its fate, for example whether it is released by gaseous discharge, removed by absorption on ion exchange (IX) resins or deposited on the fuel pin surfaces. 14 C can exist in a range of possible chemical forms: inorganic carbon compounds (probably mainly CO 2 ), elemental carbon, and organic compounds such as hydrocarbons. This paper presents results from a preliminary survey of the possible reactions of 14 C in PWR coolant. The main conclusions of the study are: - A combination of thermal and radiolytic reactions controls the chemistry of 14 C in reactor coolant. A simple chemical kinetic model predicts that CH 3 OH would be the initial product from radiolytic reactions of 14 C following its formation from 17 O. CH 3 OH is predicted to arise as a result of reactions of OH . with CH 4 and CH 3 , and it persists because there is no known radiation chemical reduction mechanism. - Thermodynamic considerations show that CH 3 OH can be thermally reduced to CH 4 in PWR conditions, although formation of CO 2 from small organics is the most thermodynamically favourable outcome. Such reactions could be catalysed on active nickel surfaces in the primary circuit. - Limited plant data would suggest that CH 4 is the dominant form in PWR and CO 2 in BWR. This implies that radiation chemistry may be important in determining the speciation. - Addition of acetate does not affect the amount of 14 C formed, but the addition of large amounts of stable carbon would lead to a large range of additional products, some of which would be expected to deposit on fuel pin surfaces as high molecular weight hydrocarbons. However, the subsequent thermal decomposition reactions of these products are not known. - Acetate addition may represent a small input of 12 C compared with organic material released from CVCS resins, although the importance of this may depend on whether that is predominantly soluble

  4. International Benchmark based on Pressurised Water Reactor Sub-channel and Bundle Tests. Volume III: Departure from Nucleate Boiling

    International Nuclear Information System (INIS)

    Rubin, Adam; Avramova, Maria; Velazquez-Lozada, Alexander

    2016-03-01

    This report summarised the second phase of the Nuclear Energy Agency (NEA) and the Nuclear Regulatory Commission (NRC) Benchmark Based on NUPEC PWR Sub-channel and Bundle Tests (PSBT), which was intended to provide data for the verification of Departure from Nucleate Boiling (DNB) prediction in existing thermal-hydraulics codes and provide direction in the development of future methods. This phase was composed of three exercises; Exercise 1: fluid temperature benchmark, Exercise 2: steady-state rod bundle benchmark and Exercise 3: transient rod bundle benchmark. The experimental data provided to the participants of this benchmark is from a series of void measurement tests using full-size mock-up tests for both BWRs and PWRs. These tests were performed from 1987 to 1995 by the Nuclear Power Engineering Corporation (NUPEC) in Japan and made available by the Japan Nuclear Energy Safety Organisation (JNES) for the purposes of this benchmark, which was organised by Pennsylvania State University. Nine institutions from seven countries participated in this benchmark. Nine different computer codes were used in Exercise 1, 2 and 3. Among the computer codes were porous media, sub-channel and systems thermal-hydraulic code. The improvement between FLICA-OVAP (sub-channel) and FLICA (sub-channel) was noticeable. The main difference between the two was that FLICA-OVAP implicitly assigned flow regime based on drift flux, while FLICA assumes single phase flows. In Exercises 2 and 3, the codes were generally able to predict the Departure from Nucleate Boiling (DNB) power as well as the axial location of the onset of DNB (for the steady-state cases) and the time of DNB (for the transient cases). It was noted that the codes that used the Electric-Power-Research- Institute (EPRI) Critical-Heat-Flux (CHF) correlation had the lowest mean error in Exercise 2 for the predicted DNB power

  5. Nuclear data needs for subcritical reactors with heavy-metal coolant

    International Nuclear Information System (INIS)

    Ignatyuk, A.V.

    2001-01-01

    Requests on improvement of evaluated data files for minor actinides (MA) are briefly reviewed. New evaluations of neutron cross sections for Np-237, Am-241 and Am-243 after the corresponding tests and verifications should satisfy the required accuracies of data for developing MA-burners. More difficult problems arise for curium isotopes, evaluated data of which are strongly divergent. International expertise of available evaluations could be very desirable. Needs in data improvements for perspective heavy-metal liquid coolants are outlined. (author)

  6. Integral nuclear power reactor with natural coolant circulation. Investigation of passive RHR system

    International Nuclear Information System (INIS)

    Samoilov, O.B.; Kuul, V.S.; Malamud, V.A.; Tarasov, G.I.

    1996-01-01

    The development of a small power (up to 240 MWe) integral PWR for nuclear co-generation power plants has been carried out. The distinctive features of this advanced reactor are: primary circuit arrangement in a single pressure vessel; natural coolant circulation; passive safety systems with self-activated control devices; use of a second (guard) vessel housing the reactor; favourable conditions for the most severe accident management. A passive steam condensing channel has been developed which is activated by the direct action of the primary circuit pressure without an automatic controlling action or manual intervention for emergency cooling of an integral reactor with an in-built pressurizer. In an emergency situation as pressure rises in the reactor a self-activated device blows out non-condensable gases from the condenser tube bundle and returns them in the steam-condensing mode of the operation with the returing primary coolant condensate into the reactor. The thermo-physical test facility is constructed and the experimental development of the steam-condensing channels is performed aiming at the verification of mathematical models for these channels operation in integral reactors both at loss-of-heat removal and LOCA accidents. (orig.)

  7. Tritium transport modeling at system level for the EUROfusion dual coolant lithium-lead breeding blanket

    Science.gov (United States)

    Urgorri, F. R.; Moreno, C.; Carella, E.; Rapisarda, D.; Fernández-Berceruelo, I.; Palermo, I.; Ibarra, A.

    2017-11-01

    The dual coolant lithium lead (DCLL) breeding blanket is one of the four breeder blanket concepts under consideration within the framework of EUROfusion consortium activities. The aim of this work is to develop a model that can dynamically track tritium concentrations and fluxes along each part of the DCLL blanket and the ancillary systems associated to it at any time. Because of tritium nature, the phenomena of diffusion, dissociation, recombination and solubilisation have been modeled in order to describe the interaction between the lead-lithium channels, the structural material, the flow channel inserts and the helium channels that are present in the breeding blanket. Results have been obtained for a pulsed generation scenario for DEMO. The tritium inventory in different parts of the blanket, the permeation rates from the breeder to the secondary coolant and the amount of tritium extracted from the lead-lithium loop have been computed. Results present an oscillating behavior around mean values. The obtained average permeation rate from the liquid metal to the helium is 1.66 mg h-1 while the mean tritium inventory in the whole system is 417 mg. Besides the reference case results, parametric studies of the lead-lithium mass flow rate, the tritium extraction efficiency and the tritium solubility in lead-lithium have been performed showing the reaction of the system to the variation of these parameters.

  8. Coolant mixing in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, T; Grunwald, G

    1998-10-01

    The behavior of PWRs during cold water or boron dilution transients is strongly influenced by the distribution of coolant temperature and boron concentration at the core inlet. This distribution is the needed input to 3-dimensional neutron kinetics to calculate the power distribution in the core. It mainly depends on how the plugs of cold or unborated water formed in a single loop are mixed in the downcomer and in the lower plenum. To simulate such mixture phenomena requires the application of 3-dimensional CFD (computational fluid dynamics) codes. The results of the simulation have to be validated against mixture experiments at scaled facilities. Therefore, in the framework of a research project funded by BMBF, the institute creates a 1:5 mixture facility representing first the geometry of a German pressurized water reactor and later the European Pressurized Water Reactor (EPR) geometry. The calculations are based on the CFD Code CFX-4. (orig.)

  9. Automated surveillance of reactor coolant pump performance

    International Nuclear Information System (INIS)

    Gross, K.C.; Singer, R.M.; Humenik, K.E.

    1992-01-01

    An artificial intelligence based expert system has been developed for continuous surveillance and diagnosis of centrifugal-type reactor coolant pump (RCP) performance and operability. The expert system continuously monitors digitized signals from a variety of physical variables (speed, vibration level, motor power, discharge pressure) associated with RCP performance for annunciation of the incipience or onset of off-normal operation. The system employs an extremely sensitive pattern-recognition technique, the sequential probability ratio test (SPRT) for rapid identification of pump operability degradation. The sequential statistical analysis of the signal noise has been shown to provide the theoretically shortest sampling time to detect disturbances and thus has the potential of providing incipient fault detection information to operators sufficiently early to avoid forced plant shutdowns. The sensitivity and response time of the expert system are analyzed in this paper using monte carlo simulation techniques

  10. Reactor coolant system and containment aqueous chemistry

    International Nuclear Information System (INIS)

    Torgerson, D.F.

    1986-01-01

    Fission products released from fuel during reactor accidents can be subject to a variety of environments that will affect their ultimate behavior. In the reactor coolant system (RCS), for example, neutral or reducing steam conditions, radiation, and surfaces could all have an effect on fission product retention and chemistry. Furthermore, if water is encountered in the RCS, the high temperature aqueous chemistry of fission products must be assessed to determine the quantity and chemical form of fission products released to the containment building. In the containment building, aqueous chemistry will determine the longer-term release of volatile fission products to the containment atmosphere. Over the past few years, the principles of physical chemistry have been rigorously applied to the various chemical conditions described above. This paper reviews the current state of knowledge and discusses the future directions of chemistry research relating to the behavior of fission products in the RCS and containment

  11. MHD considerations for poloidal-toroidal coolant ducts of self-cooled blankets

    International Nuclear Information System (INIS)

    Hua, T.Q.; Walker, J.S.

    1990-01-01

    Magnetohydrodynamic flows of liquid metals through sharp elbow ducts with rectangular cross sections and with thin conducting walls in the presence of strong uniform magnetic fields are examined. The geometries simulate the poloidaltoroidal coolant channels in fusion tokamak blankets. Analysis for obtaining the three-dimensional numerical solutions are described. Results for pressure drop, velocity profiles and flow distribution are predicted for the upcoming joint ANL/KfK sharp elbow experiment. Results from a parametric study using fusion relevant parameters to investigate the three-dimensional pressure drop are presented for possible applications to blanket designs. 10 refs., 9 refs

  12. Coolant rate distribution in horizontal steam generator under natural circulation

    International Nuclear Information System (INIS)

    Blagovechtchenski, A.; Leontieva, V.; Mitrioukhin, A.

    1997-01-01

    In the presentation the major factors determining the conditions of NCC (Natural Coolant Circulation) in the primary circuit and in particular conditions of coolant rate distribution on the horizontal tubes of PGV-1000 in NPP with VVER-1000 under NCC are considered

  13. Automatic coolant flow control device for a nuclear reactor assembly

    Science.gov (United States)

    Hutter, Ernest

    1986-01-01

    A device which controls coolant flow through a nuclear reactor assembly comprises a baffle means at the exit end of said assembly having a plurality of orifices, and a bimetallic member in operative relation to the baffle means such that at increased temperatures said bimetallic member deforms to unblock some of said orifices and allow increased coolant flow therethrough.

  14. Coolant rate distribution in horizontal steam generator under natural circulation

    Energy Technology Data Exchange (ETDEWEB)

    Blagovechtchenski, A.; Leontieva, V.; Mitrioukhin, A. [St. Petersburg State Technical Univ. (Russian Federation)

    1997-12-31

    In the presentation the major factors determining the conditions of NCC (Natural Coolant Circulation) in the primary circuit and in particular conditions of coolant rate distribution on the horizontal tubes of PGV-1000 in NPP with VVER-1000 under NCC are considered. 5 refs.

  15. Fuel coolant interaction experiment by direct electrical heating method

    International Nuclear Information System (INIS)

    Takeda, Tsuneo; Hirano, Kenmei

    1979-01-01

    In the PCM (Power Cooling Mismatch) experiments, the FCI (Fuel Coolant Interaction) test is one of necessary tests in order to predict various phenomena that occur during PCM in the core. A direct electrical heating method is used for the FCI tests for fuel pellet temperature of over 1000 0 C. Therefore, preheating is required before initiating the direct electrical heating. The fuel pin used in the FCI tests is typical LWR fuel element, which is surrounded by coolant water. It is undersirable to heat up the coolant water during preheating of the fuel pin. Therefore, a zirconia (ZrO 2 ) pellet which is similar to a UO 2 pellet in physical and chemical properties is used. Electric property (electric conductivity) of ZrO 2 is particularly suitable for direct electrical heating as in the case of UO 2 . In this experiment, ZrO 2 pellet (melting point 2500 0 C) melting was achieved by use of both preheating and direct electrical heating. Temperature changes of coolant and fuel surface, as well as the pressure change of coolant water, were measured. The molten fuel interacted with the coolant and generated shock waves. A portion of this molten fuel fragmented into small particles during this interaction. The peak pressure of the observed shock wave was about 35 bars. The damaged fuel pin was photographed after disassembly. This report shows the measured coolant pressure changes and the coolant temperature changes, as well as photographs of damaged fuel pin and fuel fragments. (author)

  16. Coolant rate distribution in horizontal steam generator under natural circulation

    Energy Technology Data Exchange (ETDEWEB)

    Blagovechtchenski, A; Leontieva, V; Mitrioukhin, A [St. Petersburg State Technical Univ. (Russian Federation)

    1998-12-31

    In the presentation the major factors determining the conditions of NCC (Natural Coolant Circulation) in the primary circuit and in particular conditions of coolant rate distribution on the horizontal tubes of PGV-1000 in NPP with VVER-1000 under NCC are considered. 5 refs.

  17. Stress Analysis of Fuel Rod under Axial Coolant Flow

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Hai Lan; Lee, Young Shin; Lee, Hyun Seung [Chungnam National University, Daejeon (Korea, Republic of); Park, Num Kyu; Jeon, Kyung Rok [Kerea Nuclear Fuel., Daejeon (Korea, Republic of)

    2010-05-15

    A pressurized water reactor(PWR) fuel assembly, is a typical bundle structure, which uses light water as a coolant in most commercial nuclear power plants. Fuel rods that have a very slender and long clad are supported by fuel assembly which consists of several spacer grids. A coolant is a fluid which flows through device to prevent its overheating, transferring the heat produced by the device to other devices that use or dissipate it. But at the same time, the coolant flow will bring out the fluid induced vibration(FIV) of fuel rods and even damaged the fuel rod. This study has been conducted to investigate the flow characteristics and nuclear reactor fuel rod stress under effect of coolant. Fluid structure interaction(FSI) analysis on nuclear reactor fuel rod was performed. Fluid analysis of the coolant which flow along the axial direction and structural analysis under effect of flow velocity were carried out under different output flow velocity conditions

  18. Method of charging instruments into liquid metal coolant

    International Nuclear Information System (INIS)

    Yamazaki, Hiroshi

    1980-01-01

    Purpose: To alleviate the thermal shock of a reactor charging machine when charging the machine into liquid metal coolant after the machine is preheated in cover gas. Method: When a reactor fueling machine reaches at the lowermost portion the position immediately above liquid metal coolant surface level, the machine is stopped moving down. The reactor fueling machine is heated at the lowermost portion by thermal radiation from the surface of the liquid metal coolant. After the machine is thus preheated in cover gas, it is again steadily moved down by a winch and charged into the liquid metal coolant. Therefore, the thermal shock of the machine becomes low when charging the machine into the liquid metal coolant to eliminate the damage and deformation at the machine. (Yoshihara, H.)

  19. Stress Analysis of Fuel Rod under Axial Coolant Flow

    International Nuclear Information System (INIS)

    Jin, Hai Lan; Lee, Young Shin; Lee, Hyun Seung; Park, Num Kyu; Jeon, Kyung Rok

    2010-01-01

    A pressurized water reactor(PWR) fuel assembly, is a typical bundle structure, which uses light water as a coolant in most commercial nuclear power plants. Fuel rods that have a very slender and long clad are supported by fuel assembly which consists of several spacer grids. A coolant is a fluid which flows through device to prevent its overheating, transferring the heat produced by the device to other devices that use or dissipate it. But at the same time, the coolant flow will bring out the fluid induced vibration(FIV) of fuel rods and even damaged the fuel rod. This study has been conducted to investigate the flow characteristics and nuclear reactor fuel rod stress under effect of coolant. Fluid structure interaction(FSI) analysis on nuclear reactor fuel rod was performed. Fluid analysis of the coolant which flow along the axial direction and structural analysis under effect of flow velocity were carried out under different output flow velocity conditions

  20. Device for preventing coolant in a reactor from being lost

    International Nuclear Information System (INIS)

    Maruyama, Hiromi; Matsumoto, Tomoyuki.

    1975-01-01

    Object: To prevent all of coolant from being lost from the core at the time of failure in rupture of pipe in a recirculation system to cool the core with the coolant remained within the reactor. Structure: A valve, which will be closed when a water level of the coolant within the core is in a level less than a predetermined level, is provided on a recirculating water outlet nozzle in a pressure vessel to thereby prevent the coolant from being lost when the pipe is broken, thus cooling the core by means of reduced-pressure boiling of coolant remained within the core and boiling due to heat, and restraining core reactivity by means of void produced at that time. (Kamimura, M.)

  1. The effect of coolant quantity on local fuel–coolant interactions in a molten pool

    International Nuclear Information System (INIS)

    Cheng, Songbai; Matsuba, Ken-ichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Tohru; Tobita, Yoshiharu

    2015-01-01

    Highlights: • We investigate local fuel–coolant interactions in a molten pool. • As water volume increases, limited pressurization and mechanical energy observed. • Only a part of water is evaporated and responsible for the pressurization. - Abstract: Studies on local fuel–coolant interactions (FCI) in a molten pool are important for severe accident analyses of sodium-cooled fast reactors (SFRs). Motivated by providing some evidence for understanding this interaction, in this study several experimental tests, with comparatively larger difference in coolant volumes, were conducted by delivering a given quantity of water into a simulated molten fuel pool (formed with a low-melting-point alloy). Interaction characteristics including the pressure-buildup as well as mechanical energy release and its conversion efficiency are evaluated and compared. It is found that as water quantity increases, a limited pressure-buildup and the resultant mechanical energy release are observable. The performed analyses also suggest that only a part of water is probably vaporized during local FCIs and responsible for the pressurization and mechanical energy release, especially for those cases with much larger water volumes

  2. Technical meeting on 'Primary coolant pipe rupture event in liquid metal cooled fast reactors'. Working material

    International Nuclear Information System (INIS)

    2003-01-01

    In Liquid Metal cooled Fast Reactors (LMFR) or in accelerator driven sub-critical systems (ADS) with LMFR like sub-critical cores, the primary coolant pipes (PCP) connect the primary coolant pumps to the grid plate. A rupture in one of these pipes could cause significant loss of coolant flow to the core with severe consequences. In loop type reactors, all primary pipelines are provided with double envelopes and inter-space coolant leak monitoring systems that permit leak detection before break. Thus, the PCP rupture event can be placed in the beyond design basis event (BDBE) category. Such an arrangement is difficult to incorporate for pool type reactors, and hence it could be argued that the PCP rupture event needs to be analysed in detail as a design basis event (DBE, category 4 event). The primary coolant pipes are made of ductile austenitic stainless steel material and operate at temperatures of the cold pool and at comparatively low pressures. For such low stressed piping with negligible creep and embrittlement effects, it is of interest to discuss under what design provisions, for pool type reactors, the guillotine rupture of PCP could be placed in the BDBE category. The topical Technical Meeting (TM) on 'Primary Coolant Pipe Rupture Event in Liquid Metal Cooled Reactors' was called to enable the specialists to present the philosophy and analyses applied on this topic in the various Member States for different LMFRs. The scope of the Technical Meeting was to provide a global forum for information exchange on the philosophy applied in the various participating Member States and the analyses performed for different LMFRs with regard to the primary coolant pipe rupture event. More specifically, the objectives of the Technical Meeting were to review the safety philosophy for the PCP rupture event in pool type LMFR, to assess the structural reliability of the PCP and the probability of rupture under different conditions (with/without in-service inspection), to

  3. Evaluation of the channelized Hotelling observer with an internal-noise model in a train-test paradigm for cardiac SPECT defect detection

    International Nuclear Information System (INIS)

    Brankov, Jovan G

    2013-01-01

    The channelized Hotelling observer (CHO) has become a widely used approach for evaluating medical image quality, acting as a surrogate for human observers in early-stage research on assessment and optimization of imaging devices and algorithms. The CHO is typically used to measure lesion detectability. Its popularity stems from experiments showing that the CHO's detection performance can correlate well with that of human observers. In some cases, CHO performance overestimates human performance; to counteract this effect, an internal-noise model is introduced, which allows the CHO to be tuned to match human-observer performance. Typically, this tuning is achieved using example data obtained from human observers. We argue that this internal-noise tuning step is essentially a model training exercise; therefore, just as in supervised learning, it is essential to test the CHO with an internal-noise model on a set of data that is distinct from that used to tune (train) the model. Furthermore, we argue that, if the CHO is to provide useful insights about new imaging algorithms or devices, the test data should reflect such potential differences from the training data; it is not sufficient simply to use new noise realizations of the same imaging method. Motivated by these considerations, the novelty of this paper is the use of new model selection criteria to evaluate ten established internal-noise models, utilizing four different channel models, in a train-test approach. Though not the focus of the paper, a new internal-noise model is also proposed that outperformed the ten established models in the cases tested. The results, using cardiac perfusion SPECT data, show that the proposed train-test approach is necessary, as judged by the newly proposed model selection criteria, to avoid spurious conclusions. The results also demonstrate that, in some models, the optimal internal-noise parameter is very sensitive to the choice of training data; therefore, these models are prone

  4. Evaluation of the channelized Hotelling observer with an internal-noise model in a train-test paradigm for cardiac SPECT defect detection.

    Science.gov (United States)

    Brankov, Jovan G

    2013-10-21

    The channelized Hotelling observer (CHO) has become a widely used approach for evaluating medical image quality, acting as a surrogate for human observers in early-stage research on assessment and optimization of imaging devices and algorithms. The CHO is typically used to measure lesion detectability. Its popularity stems from experiments showing that the CHO's detection performance can correlate well with that of human observers. In some cases, CHO performance overestimates human performance; to counteract this effect, an internal-noise model is introduced, which allows the CHO to be tuned to match human-observer performance. Typically, this tuning is achieved using example data obtained from human observers. We argue that this internal-noise tuning step is essentially a model training exercise; therefore, just as in supervised learning, it is essential to test the CHO with an internal-noise model on a set of data that is distinct from that used to tune (train) the model. Furthermore, we argue that, if the CHO is to provide useful insights about new imaging algorithms or devices, the test data should reflect such potential differences from the training data; it is not sufficient simply to use new noise realizations of the same imaging method. Motivated by these considerations, the novelty of this paper is the use of new model selection criteria to evaluate ten established internal-noise models, utilizing four different channel models, in a train-test approach. Though not the focus of the paper, a new internal-noise model is also proposed that outperformed the ten established models in the cases tested. The results, using cardiac perfusion SPECT data, show that the proposed train-test approach is necessary, as judged by the newly proposed model selection criteria, to avoid spurious conclusions. The results also demonstrate that, in some models, the optimal internal-noise parameter is very sensitive to the choice of training data; therefore, these models are prone

  5. International Conference on Legal Medicine and Forensic Sciences 8th Cross Channel Conference- Forensic Sciences Recent Developments

    Directory of Open Access Journals (Sweden)

    Yasemin Günay Balcı

    2004-12-01

    Full Text Available International Conrefence on Legal Medicine and Forensic Sciences, 8th Cross Channel Conference (Uluslararası Adli Tıp ve Adli Bilimler Üzerine Konferans, 20-24 Nisan 2004 tarihleri arasında Belçika-Brugge'de yapıldı. Konferans ana başlığı “Adli Bilimler-Son Gelişmeler” şeklinde idi. Konferansı organize eden kuruluşlar; -\tBelçika Adli Tıp Topluluğu, -\tFransa Adli Tıp ve Kriminoloji Topluluğu, -\tFlollanda Adli Tıp Topluluğu, -\tIngiltere Adli Tabipler Birliği idi. Kongrenin ilk iki günü “Çalıştay (Workshop” şeklinde toplantılar vardı. Çalıştay konuları; -\tAntropoloji-Yanmış Kemikler ve Adli Antropolojide Mikroskobik Çalışma -\tAdli Tıp ve Adli Bilimlerde Kalite Güvencesi -\tMultidisipliner Ekip içinde Adli Diş Hekimliği -\tFasial rekonstrüksiyon ve Restorasyon idi. Bunlardan “Adli Tıp ve Adli Bilimlerde Kalite Güvencesi” başlıklı çalıştay katılım azlığı nedeniyle iptal edilmişti. “Antropoloji-Yanmış Kemikler ve Adli Antropolojide Mikroskobik Çalışma” konulu çalıştay uygulamalı olup, katılımcılar kemik dokusundan mikroskobik inceleme için manuel olarak, ucuz ve kısa sürede préparât hazırlamayı öğrendiler. Tüm katılanlar kendi hazırladıkları en az 3 preparatla ülkelerine ya da birimlerine döndüler. Yine katılımcılar elde ettikleri bu beceri ile kendi çalıştıkları birimlerde, oldukça ucuz maliyetle mikroskobik inceleme için kemik preparatı hazırlama laboratuvarı oluşturabilecek durumda olmanın sevincini yaşadılar. Kolay ve oldukça güzel sonuç alıcı bu yöntem kimliği belirsiz kemiklerde kemik yenilenme durumuna göre yaş tayini gibi adli kemik incelemeleri yanı sıra, örneğin ortopedi ameliyatlarında tümör özelliğinin belirlenmesi gibi kemikten acil mikroskobik inceleme gereken durumlarda da kullanılabilir. “Multididipliner Ekip içinde Adli Diş Hekimliği” konulu çahştayda adli diş hekimliğine yakla

  6. Evaluation of alternate secondary (and tertiary) coolants for the molten-salt breeder reactor

    International Nuclear Information System (INIS)

    Kelmers, A.D.; Baes, C.F.; Bettis, E.S.; Brynestad, J.; Cantor, S.; Engel, J.R.; Grimes, W.R.; McCoy, H.E.; Meyer, A.S.

    1976-04-01

    The three most promising coolant selections for an MSBR have been identified and evaluated in detail from the many coolants considered for application either as a secondary coolant in 1000-MW(e) MSBR configurations using only one coolant, or as secondary and tertiary coolants in an MSBR dual coolant configuration employing two different coolants. These are, as single secondary coolants: (1) a ternary sodium--lithium--beryllium fluoride melt; (2) the sodium fluoroborate--sodium fluoride eutectic melt, the present reference design secondary coolant. In the case of the dual coolant configuration, the preferred system is molten lithium--beryllium fluoride (Li 2 BeF 4 ) as the secondary coolant and helium gas as the tertiary coolant

  7. EXPERIMENTAL STUDY OF LOCAL HYDRODYNAMICS AND MASS EXCHANGE PROCESSES OF COOLANT IN FUEL ASSEMBLIES OF PRESSURIZED WATER REACTORS

    Directory of Open Access Journals (Sweden)

    S. M. Dmitriev

    2016-01-01

    Full Text Available The results of experimental studies of local hydrodynamics and mass exchange of coolant flow behind spacer and mixing grids of different structural versions that were developed for fuel assemblies of domestic and foreign nuclear reactors are presented in the article. In order to carry out the study the models of the following fuel assemblies have been fabricated: FA for VVER and VBER, FA-KVADRAT for PWR-reactor and FA for KLT-40C reactor. All the models have been fabricated with a full geometrical similarity with full-scale fuel assemblies. The study was carried out by simulating the flow of coolant in a core by air on an aerodynamic test rig. In order to measure local hydrodynamic characteristics of coolant flow five-channel Pitot probes were used that enable to measure the velocity vector in a point by its three components. The tracerpropane method was used for studying mass transfer processes. Flow hydrodynamics was studied by measuring cross-section velocities of coolant flow and coolant rates according to the model cells. The investigation of mass exchange processes consisted of a study of concentration distribution for tracer in experimental model, in determination of attenuation lengths of mass transfer processes behind mixing grids, in calculating of inter-cellar mass exchange coefficient. The database on coolant flow in fuel assemblies for different types of reactors had been accumulated that formed the basis of the engineering substantiation of reactor cores designs. The recommendations on choice of optimal versions of mixing grids have been taken into consideration by implementers of the JSC “OKBM Afrikantov” when creating commissioned fuel assemblies. The results of the study are used for verification of CFD-codes and CFD programs of detailed cell-by-cell calculation of reactor cores in order to decrease conservatism for substantiation of thermal-mechanical reliability.

  8. Analysis of fuel rod behaviour within a rod bundle of a pressurized water reactor under the conditions of a loss of coolant accident (LOCA) using probabilistic methodology

    International Nuclear Information System (INIS)

    Sengpiel, W.

    1980-12-01

    The assessment of fuel rod behaviour under PWR LOCA conditions aims at the evaluation of the peak cladding temperatures and the (final) maximum circumferential cladding strains. Moreover, the estimation of the amount of possible coolant channel blockages within a rod bundle is of special interest, as large coplanar clad strains of adjacent rods may result in strong local reductions of coolant channel areas. Coolant channel blockages of large radial extent may impair the long-term coolability of the corresponding rods. A model has been developed to describe these accident consequences using probabilistic methodology. This model is applied to study the behaviour of fuel rods under accident conditions following the double-ended pipe rupture between collant pump and pressure vessel in the primary system of a 1300 MW(el)-PWR. Specifically a rod bundle is considered consisting of 236 fuel rods, that is subjected to severe thermal and mechanical loading. The results obtained indicate that plastic clad deformations with circumferential clad strains of more than 30% cannot be excluded for hot rods of the reference bundle. However, coplanar coolant channel blockages of significant extent seem to be probable within that bundle only under certain boundary conditions which are assumed to be pessimistic. (orig./RW) [de

  9. Coolant mixing in pressurized water reactors. Proceedings

    International Nuclear Information System (INIS)

    Hoehne, T.; Grunwald, G.; Rohde, U.

    1998-10-01

    For the analysis of boron dilution transients and main steam like break scenarios the modelling of the coolant mixing inside the reactor vessel is important. The reactivity insertion due to overcooling or deboration depends strongly on the coolant temperature and boron concentration. The three-dimensional flow distribution in the downcomer and the lower plenum of PWR's was calculated with a computational fluid dynamics (CFD) code (CFX-4). Calculations were performed for the PWR's of SIEMENS KWU, Westinghouse and VVER-440 / V-230 type. The following important factors were identified: exact representation of the cold leg inlet region (bend radii etc.), extension of the downcomer below the inlet region at the PWR Konvoi, obstruction of the flow by the outlet nozzles penetrating the downcomer, etc. The k-ε turbulence model was used. Construction elements like perforated plates in the lower plenum have large influence on the velocity field. It is impossible to model all the orifices in the perforated plates. A porous region model was used to simulate perforated plates and the core. The porous medium is added with additional body forces to simulate the pressure drop through perforated plates in the VVER-440. For the PWR Konvoi the whole core was modelled with porous media parameters. The velocity fields of the PWR Konvoi calculated for the case of operation of all four main circulation pumps show a good agreement with experimental results. The CFD-calculation especially confirms the back flow areas below the inlet nozzles. The downcomer flow of the Russian VVER-440 has no recirculation areas under normal operation conditions. By CFD calculations for the downcomer and the lower plenum an analytical mixing model used in the reactor dynamic code DYN3D was verified. The measurements, the analytical model and the CFD-calculations provided very well agreeing results particularly for the inlet region. The difficulties of analytical solutions and the uncertainties of turbulence

  10. International nuclear safety center database on material properties

    International Nuclear Information System (INIS)

    Fink, J.K.

    1996-01-01

    International nuclear safety center database on the following material properties is described: fuel, cladding,absorbers, moderators, structural materials, coolants, concretes, liquid mixtures, uranium dioxide

  11. Experimental determination of average turbulent heat transfer and friction factor in stator internal rib-roughened cooling channels.

    Science.gov (United States)

    Battisti, L; Baggio, P

    2001-05-01

    In gas turbine cooling design, techniques for heat extraction from the surfaces exposed to the hot stream are based on the increase of the inner heat transfer areas and on the promotion of the turbulence of the cooling flow. This is currently obtained by casting periodic ribs on one or more sides of the serpentine passages into the core of the blade. Fluid dynamic and thermal behaviour of the cooling flow have been extensively investigated by means of experimental facilities and many papers dealing with this subject have appeared in the latest years. The evaluation of the average value of the heat transfer coefficient most of the time is inferred from local measurements obtained by various experimental techniques. Moreover the great majority of these studies are not concerned with the overall average heat transfer coefficient for the combined ribs and region between them, but do focus just on one of them. This paper presents an attempt to collect information about the average Nusselt number inside a straight ribbed duct. Series of measurements have been performed in steady state eliminating the error sources inherently connected with transient methods. A low speed wind tunnel, operating in steady state flow, has been built to simulate the actual flow condition occurring in a rectilinear blade cooling channel. A straight square channel with 20 transverse ribs on two sides has been tested for Re of about 3 x 10(4), 4.5 x 10(4) and 6 x 10(4). The ribbed wall test section is electrically heated and the heat removed by a stationary flow of known thermal and fluid dynamic characteristics.

  12. SLAC divertor channel entrance thermal stress analysis

    International Nuclear Information System (INIS)

    Johnson, G.L.; Stein, W.; Lu, S.C.; Riddle, R.A.

    1985-01-01

    X-ray beams emerging from the new SLAC electron-positron storage ring (PEP) impinge on the entrance to tangential divertor channels causing highly localized heating in the channel structure. Analyses were completed to determine the temperatures and thermally-induced stresses due to this heating. These parts are cooled with water flowing axially over them at 30 0 C. The current design and operating conditions should result in the entrance to the new divertor channel operating at a peak temperature of 123 0 C with a peak thermal stress at 91% of yield. Any micro-cracks that form due to thermally-induced stresses should not propagate to the coolant wall nor form a path for the coolant to leak into the storage ring vacuum. 34 figs., 4 tabs

  13. Device for preventing coolant outflow in a reactor

    International Nuclear Information System (INIS)

    Nemoto, Kiyomitsu; Mochizuki, Keiichi.

    1975-01-01

    Object: To prevent outflow of coolant from a reactor vessel even in an occurrence of leaking trouble at a low position in a primary cooling system or the like in the reactor vessel. Structure: An inlet at the foremost end of a coolant inlet pipe inserted into a reactor vessel is arranged at a level lower than a core, and a check valve is positioned at a level higher than the core in a rising portion of the inlet. In normal condition, the check valve is pushed up by discharge pressure of a main circulating pump and remains closed, and hence, producing no flow loss of coolant, sodium. However, when a trouble such as rupture occurs at the lower position in the primary cooling system, the attractive force for allowing the coolant to back-flow outside the reactor vessel and the load force of the coolant within the reactor vessel cause the check valve to actuate, as a consequence of which a liquid level of the coolant downwardly moves to the position of the check valve to intake the cover gases into a gas intake, thereby cutting off a flow passage of the coolant to stop outflow thereof. (Kamimura, M.)

  14. Development of lead-bismuth coolant technology for nuclear device

    International Nuclear Information System (INIS)

    Kamata, Kin-ya; Kitano, Teruaki; Ono, Mikinori

    2004-01-01

    Liquid lead-bismuth is a promising material as a future fast reactor coolant or an intensive neutron source material for accelerator driven transmutation system (ADS). To develop nuclear plants and their installations using lead-bismuth coolant for practical use, both coolant technologies, inhabitation process of steels and quality control of coolant, and total operation system for liquid lead-bismuth plants are required. Based on the experience of liquid metal coolant, Mitsui Engineering and Shipbuilding Co., Ltd. (MES) has completed the liquid lead-bismuth forced circulation loop and has acquired various engineering data on main components including economizer. As a result of tis operation, MES has developed key technologies of lead-bismuth coolant such as controlling of oxygen content in lead-bismuth and a purification of lead-bismuth coolant. MES participated in the national project, ''The Development of Accelerator Driven Transmutation System'', together with JAERI (Japan Atomic Energy Research Institute) and started corrosion test for beam window of ADS. (author)

  15. Liquid metal coolant disposal from UKAEA reactors at Dounreay

    International Nuclear Information System (INIS)

    Adam, E.R.

    1997-01-01

    As part of the United Kingdom's Fast Reactor Development programme two reactors were built and operated at Dounreay in the North of Scotland. DFR (Dounreay Fast Reactor) was operated from 1959-1977 and PFR (Prototype Fast Reactor) was operated from 1974-1994. Both reactors are currently undergoing Stage 1 Decommissioning and are installing plant to dispose of the bulk coolant (DFR ∼ 60 tonne; PFR ∼ 1500 tonne). The coolant (NaK) remaining at DFR is mainly in the primary circuit which contains in excess of 500 TBq of Cs137. Disposal of 40 tonnes of secondary coolant has already been carried out. The paper will describe the processes used to dispose of this secondary circuit coolant and how it is intended the remaining primary circuit coolant will be handled. The programme to process the primary coolant will also be described which involves the conversion of the liquid metal to caustic and its decontamination. No PFR coolant Na has been disposed off to date. The paper will describe the current decommissioning programme activities relating to liquid metal disposal and treatment describing the materials to be disposed of and the issue of decontamination of the effluents. (author)

  16. Q-factor of coolant flow in the primary circuit of NPP with pressurised water reactors

    International Nuclear Information System (INIS)

    Proskuryakov, K.N.; Belikov, S.O.; Novikov, K.S.

    2011-01-01

    Systems of preoperational vibration dynamic monitoring in of WWER are presented. The results of measurements during commission of NPP with WWER are presented. The paper provides the result of the research, that estimation of coolant fluctuations caused by pulse perturbation of pressure in the primary circuit NPP. It is shown that results could be received at known value of a Q - factor of acoustical oscillatory system only. The research demonstrates the results of dependence of the sound speed from the mass steam content in the coolant flow thru reactor core. The worked out results can be used for identification of the reasons of abnormal growth of level of vibrations of fuel assembly, fuel rod, equipment and internals, and for forecasting the operation conditions which provide of vibration - acoustical resonances in the primary loop equipment. (author)

  17. Interfacing systems loss of coolant accident (ISLOCA) pressure capacity methodology and Davis-Besse results

    International Nuclear Information System (INIS)

    Wesley, D.A.

    1991-01-01

    A loss of coolant accident resulting from the overpressurization by reactor coolant fluid of a system designed for low-pressure, low-temperature service has been identified as a potential contributor to nuclear power plant risk. In this paper, the methodology developed to assess the probability of failure as a function of internal pressure is presented, and sample results developed for the controlling failure modes and locations of four fluid systems at the Davis-Besse Plant are shown. Included in this evaluation are the tanks, heat exchangers, filters, pumps, valves, and flanged connections for each system. The variability in the probability of failure is included, and the estimated leak rates or leak areas are given for the controlling modes of failure. For this evaluation, all failures are based on quasistatic pressures since the probability of dynamic effects resulting from such causes as water hammer have been initially judged to be negligible for the Davis-Besse plant ISLOCA

  18. The premixing and propagation phases of fuel-coolant interactions: a review of recent experimental studies and code developments

    International Nuclear Information System (INIS)

    Antariksawan, A.R.; Moriyama, Kiyofumi; Park, Hyun-sun; Maruyama, Yu; Yang, Yanhua; Sugimoto, Jun

    1998-09-01

    A vapor explosion (or an energetic fuel-coolant interactions, FCIs) is a process in which hot liquid (fuel) transfers its internal energy to colder, more volatile liquid (coolant); thus the coolant vaporizes at high pressure and expands and does works on its surroundings. Traditionally, the energetic fuel-coolant interactions could be distinguished in subsequent stages: premixing (or coarse mixing), triggering, propagation and expansion. Realizing that better and realistic prediction of fuel-coolant interaction consequences will be available understanding the phenomenology in the premixing and propagation stages, many experimental and analytical studies have been performed during more than two decades. A lot of important achievements are obtained during the time. However, some fundamental aspects are still not clear enough; thus the works are directed to that direction. In conjunction, the model/code development is pursuit. This is aimed to provide a scaling tool to bridge the experimental results to the real geometries, e.g. reactor pressure vessel, reactor containment. The present review intends to collect the available information on the recent works performed to study the premixing and propagation phases. (author). 97 refs

  19. The premixing and propagation phases of fuel-coolant interactions: a review of recent experimental studies and code developments

    Energy Technology Data Exchange (ETDEWEB)

    Antariksawan, A.R. [Reactor Safety Technology Research Center of BATAN (Indonesia); Moriyama, Kiyofumi; Park, Hyun-sun; Maruyama, Yu; Yang, Yanhua; Sugimoto, Jun

    1998-09-01

    A vapor explosion (or an energetic fuel-coolant interactions, FCIs) is a process in which hot liquid (fuel) transfers its internal energy to colder, more volatile liquid (coolant); thus the coolant vaporizes at high pressure and expands and does works on its surroundings. Traditionally, the energetic fuel-coolant interactions could be distinguished in subsequent stages: premixing (or coarse mixing), triggering, propagation and expansion. Realizing that better and realistic prediction of fuel-coolant interaction consequences will be available understanding the phenomenology in the premixing and propagation stages, many experimental and analytical studies have been performed during more than two decades. A lot of important achievements are obtained during the time. However, some fundamental aspects are still not clear enough; thus the works are directed to that direction. In conjunction, the model/code development is pursuit. This is aimed to provide a scaling tool to bridge the experimental results to the real geometries, e.g. reactor pressure vessel, reactor containment. The present review intends to collect the available information on the recent works performed to study the premixing and propagation phases. (author). 97 refs.

  20. Power supplyer for reactor coolant recycling pump

    International Nuclear Information System (INIS)

    Nara, Hiroshi; Okinaka, Yo.

    1991-01-01

    The present invention concerns a variable voltage/variable frequency static power source (static power source) used as a power source for a coolants recycling pump motor of a nuclear power plant. That is, during lower power operation such as start up or shutdown in which stoppage of the power source gives less effect to a reactor core, power is supplied from a power system, a main power generator connected thereto or a high voltage bus in the plant or a common high voltage bus to the static power source. However, during rated power operation, power is supplied from the output of an axially power generator connected with a main power generator having an extremely great inertia moment to the static power device. With such a constitution, the static power device is not stopped by the lowering of the voltage due to a thunderbolt falling accident or the like to a power-distribution line suddenly occurred in the power system. Accordingly, reactor core flowrate is free from rapid decrease caused by the reduction of rotation speed of the recycling pump. Accordingly, disadvantgages upon operation control in the reactor core is not caused. (I.S.)

  1. Characterization of reactor coolant by XRF

    Energy Technology Data Exchange (ETDEWEB)

    Legreid, G.; Beverskog, B. [OECD Halden Reactor Project (Norway)

    2002-07-01

    The analyzes of membrane filters is of utmost importance in characterizing the coolant chemistry in nuclear power plants. Traditional analyzes of filters includes oxidative digestion followed by instrumental analyzes. XRF (X-ray Fluorescence spectrometry) can analyze without digestion of the filters. The method is much faster and demands only a cutting step as sample preparation. By use of XRF the analytical laboratory at the Halden Reactor Project will get increased capacity, which makes it possible to analyze more samples and improve the characterization of the water. The method has shown to give more stable results than other methods in use, and has proved to have good precision. New calibration methods have been developed and tested successfully against other methods. A round robin test, attending seven laboratories from nuclear power plants, was initiated by the Halden Project to verify the instrument. The test of standard cation exchange filters showed that conventional filter digestion results in too low values. The XRF methodology shows very good agreement with the standard values. The round robin test for particle filters could not confirm that filter digestion results in too low values. This was mainly due to lack of standard particle filters and large scatter in the reported data. (author)

  2. Characterization of reactor coolant by XRF

    International Nuclear Information System (INIS)

    Legreid, G.; Beverskog, B.

    2002-01-01

    The analyzes of membrane filters is of utmost importance in characterizing the coolant chemistry in nuclear power plants. Traditional analyzes of filters includes oxidative digestion followed by instrumental analyzes. XRF (X-ray Fluorescence spectrometry) can analyze without digestion of the filters. The method is much faster and demands only a cutting step as sample preparation. By use of XRF the analytical laboratory at the Halden Reactor Project will get increased capacity, which makes it possible to analyze more samples and improve the characterization of the water. The method has shown to give more stable results than other methods in use, and has proved to have good precision. New calibration methods have been developed and tested successfully against other methods. A round robin test, attending seven laboratories from nuclear power plants, was initiated by the Halden Project to verify the instrument. The test of standard cation exchange filters showed that conventional filter digestion results in too low values. The XRF methodology shows very good agreement with the standard values. The round robin test for particle filters could not confirm that filter digestion results in too low values. This was mainly due to lack of standard particle filters and large scatter in the reported data. (author)

  3. Numerical model simulation of atmospheric coolant plumes

    International Nuclear Information System (INIS)

    Gaillard, P.

    1980-01-01

    The effect of humid atmospheric coolants on the atmosphere is simulated by means of a three-dimensional numerical model. The atmosphere is defined by its natural vertical profiles of horizontal velocity, temperature, pressure and relative humidity. Effluent discharge is characterised by its vertical velocity and the temperature of air satured with water vapour. The subject of investigation is the area in the vicinity of the point of discharge, with due allowance for the wake effect of the tower and buildings and, where application, wind veer with altitude. The model equations express the conservation relationships for mometum, energy, total mass and water mass, for an incompressible fluid behaving in accordance with the Boussinesq assumptions. Condensation is represented by a simple thermodynamic model, and turbulent fluxes are simulated by introduction of turbulent viscosity and diffusivity data based on in-situ and experimental water model measurements. The three-dimensional problem expressed in terms of the primitive variables (u, v, w, p) is governed by an elliptic equation system which is solved numerically by application of an explicit time-marching algorithm in order to predict the steady-flow velocity distribution, temperature, water vapour concentration and the liquid-water concentration defining the visible plume. Windstill conditions are simulated by a program processing the elliptic equations in an axisymmetrical revolution coordinate system. The calculated visible plumes are compared with plumes observed on site with a view to validate the models [fr

  4. Speed control device for coolant recycling pump

    International Nuclear Information System (INIS)

    Kageyama, Takao.

    1992-01-01

    The present invention intends to increase a margin relative of the oscillations of neutron fluxes when the temperature of feedwater is lowered in a compulsory recycling type BWR reactor. That is, when the operation point represented by a reactor thermal power and a reactor core inlet flow rate is in a state approximate to an oscillation limit of the reactor power, the device of the present invention controls the recycling pump speed in the increasing direction depending on the lowering range of the feedwater temperature from a stationary state. With such a constitution, even if the reactor power is in the operation region near the oscillation limit in the BWR type reactor and a feedwater heating loss is caused, the speed of the coolant recycling pump is increased by 10% at the maximum depending on the extent of the reduction of the feedwater temperature, so that the oscillation of the reactor power can be prevented from lasting for a long period of time even if a reactivity external disturbance should occur in the reactor. (I.S.)

  5. Reactor coolant pump monitoring and diagnostic system

    International Nuclear Information System (INIS)

    Singer, R.M.; Gross, K.C.; Walsh, M.; Humenik, K.E.

    1990-01-01

    In order to reliably and safely operate a nuclear power plant, it is necessary to continuously monitor the performance of numerous subsystems to confirm that the plant state is within its prescribed limits. An important function of a properly designed monitoring system is the detection of incipient faults in all subsystems (with the avoidance of false alarms) coupled with an information system that provides the operators with fault diagnosis, prognosis of fault progression and recommended (either automatic or prescriptive) corrective action. In this paper, such a system is described that has been applied to reactor coolant pumps. This system includes a sensitive pattern-recognition technique based upon the sequential probability ratio test (SPRT) that detects incipient faults from validated signals, an expert system embodying knowledge bases on pump and sensor performance, extensive hypertext files containing operating and emergency procedures as well as pump and sensor information and a graphical interface providing the operator with easily perceived information on the location and character of the fault as well as recommended corrective action. This system is in the prototype stage and is currently being validated utilizing data from a liquid-metal cooled fast reactor (EBR-II). 3 refs., 4 figs

  6. Alternative protections for loss of coolant accidents

    International Nuclear Information System (INIS)

    Estevez, E.A.

    1997-01-01

    One way to mitigate a small loss of coolant accident (LOCA) is by depressurizing the primary system, in order to turn the accident into a sequence where water is fed to a low pressure system. It can be achieved by two different ways: by incorporating a valve system (ADS - Automatic Depressurization System) to the design, which helps to diminish the pressure, obtaining a bigger LOCA, or by extracting heat from the system. Our analysis is centered in integrated reactors. The first characterization performed was on CAREM reactor. The idea was then to observe its behavior with LOCAs for different thermal power relations, water volume and rupture area. A simple depressurization model is presented, which enables us to find the parameter relationships which characterize this process, from which some particular cases will arise. ADS implementation is then analyzed, giving the criteria for the triggering time. A study on its reliability and the probability of a spurious opening is made, taking into account independent and dependent failures. An analysis on heat extraction as alternative for depressurizing is also made. Finally, the different reasons to choose between ADS or heat extraction as alternative are given, and the meaning of the parameters found are discussed. An alternative to classify LOCAs, instead of the traditional classification, by fracture size, is suggested. (author)

  7. Method of decontaminating primary coolant circuits

    International Nuclear Information System (INIS)

    Ishibashi, Masaru; Sumi, Masao.

    1981-01-01

    Purpose: To eliminate hard contaminated layers as well as soft contaminated layers without injuring substrate materials, upon decontamination of radiation contaminated portions in equipments and pipeways constituting primary coolant circuits. Constitution: High pressure water from a high pressure pump is jetted out from the nozzle of a spray gun to the radiation contaminated portions in equipments, for example, to the surface of water chamber in a vapor evaporator. High pressure pure water or aqueous boric acid is jetted out from the periphery and boric oxide particles (of about 1 - 100 μ particle size) are jetted out from the center of the nozzle of the spray gun. The particles (blasting material) jetted out together with the high pressure water impinge on the contaminated surfaces to remove the contaminated layers. Upon impingement, the high pressure water acts as the shock absorber for the blasting material and, after the impingement, it flows down to the bottom of the water chamber, and the blasting material is dissolved in the high pressure water. (Horiuchi, T.)

  8. Steam as turbine blade coolant: Experimental data generation

    Energy Technology Data Exchange (ETDEWEB)

    Wilmsen, B.; Engeda, A.; Lloyd, J.R. [Michigan State Univ., East Lansing, MI (United States)

    1995-10-01

    Steam as a coolant is a possible option to cool blades in high temperature gas turbines. However, to quantify steam as a coolant, there exists practically no experimental data. This work deals with an attempt to generate such data and with the design of an experimental setup used for the purpose. Initially, in order to guide the direction of experiments, a preliminary theoretical and empirical prediction of the expected experimental data is performed and is presented here. This initial analysis also compares the coolant properties of steam and air.

  9. Evaluation of primary coolant leaks and assessment of detection methods

    International Nuclear Information System (INIS)

    Cassette, P.; Giroux, C.; Roche, H.; Seveon, J.J.

    1984-11-01

    A review of French PWR situation concerning primary coolant leaks is presented, including a description of operating technical specifications, of the collecting system of primary coolant leakage into the containment and of the detection methods. It is mainly based on a compilation over three years, 1981 to 1983, of almost all occurred leaks, their natures, causes, consequences and methods used for their detection. By analysing these data it is possible to evaluate the efficiency of the primary coolant leak detection system and the problems raised by the compliance with the criteria defined in the operating technical specifications

  10. Evaluation of primary coolant leaks and assessment of detection methods

    International Nuclear Information System (INIS)

    Cassette, P.; Giroux, C.; Roche, H.; Seveon, J.J.

    1986-01-01

    A review of the French PWR situation concerning primary coolant leaks is presented, including a description of operating technical specifications, of the collecting system of primary coolant leakage into the containment and of the detection methods. It is mainly based on a compilation over three years, 1981 to 1983, of almost all actual leaks, their natures, causes, consequences and methods used for their detection. By analysing these data it is possible to evaluate the efficiency of the primary coolant leak detection system and the problems raised by compliance with the criteria defined in the operating technical specifications

  11. Evaluation of filtration and distillation methods for recycling automotive coolant

    International Nuclear Information System (INIS)

    Randall, P.M.; Gavaskar, A.R.

    1992-01-01

    Government regulations and high waste disposal cost of spent automotive coolant have driven the vehicle maintenance industry to explore on-site recycling. The USEPA in cooperation with the New Jersey Department of Environmental Protection (NJDEP) and the New Jersey Department of Transportation (NJDOT) evaluated two commercially available technologies that have potential for reducing the volume of spent automotive coolant. The objective of this study was to evaluate the quality of the recycled coolant, the pollution prevention potential, and the economic feasibility of the technologies

  12. Low-activation lead coolant for advanced small modular NPP

    International Nuclear Information System (INIS)

    Khorasanov, G.L.; Ivanov, A.P.; Blokhin, A.I.

    2001-01-01

    The purpose of the paper is in studying perspectives of a new heavy liquid metal coolant for a small fast reactor (FR) concept. To reduce the post irradiation activity of the coolant the using of lead isotope, Pb-206, instead of natural lead, Pb-nat, is offered. In this case the accumulation of such hazardous radionuclides, as Po-210, Bi-208, Bi-207, essentially decreases. The interval of the lead-206 coolant cost which does not exceed 20% of the overall FR cost is estimated. The possibility of lead-206 obtaining for FR needs with the centrifugal separation technique is pointed out. (author)

  13. Analysis of multi-dimensional and countercurrent effects in a BWR loss-of-coolant accident

    International Nuclear Information System (INIS)

    Shiralkar, B.S.; Dix, G.E.; Alamgir, M.

    1989-01-01

    The presence of parallel enclosed channels in a BWR provides opportunities for multiple flow regimes in co-current and countercurrent flow under Loss-of-Coolant Accident (LOCA) conditions. To address and understand these phenomena, an integrated experimental and analytical study has been conducted. The primary experimental facility was the Steam Sector Test Facility (SSTF) which simulated a full scale 30deg sector of a BWR/6 reactor vessel. Both steady-state separate effects tests and integral transients with vessel blowdown and refill were performed. The present of multi-dimensional and parallel channel effects was found to be very beneficial to BWR LOCA performance. The best estimate TRAC-BWR computer code was extended as part of this study by incorporation of a phenomenological upper plenum mixing model. TRAC-BWR was applied to the analysis of these full scale experiments. Excellent predictions of phenomena and experimental trends were achieved. (orig.)

  14. Analysis of multidimensional and countercurrent effects in a BWR loss-of-coolant accident

    International Nuclear Information System (INIS)

    Shiralkar, B.S.; Dix, G.E.; Alamgir, M.

    1991-01-01

    The presence of parallel enclosed channels in a boiling water reactor (BWR) provides opportunities for multiple flow regimes in cocurrent and countercurrent flow under loss-of-coolant accident (LOCA) conditions. To address and understand these phenomena, an integrated experimental and analytical study has been conducted. The primary experimental facility was the steam sector test facility (SSFT), which simulated a full scale 30deg sector of a BWR/6 reactor vessel. Both steady-state separate effects tests an integral transients with vessel vlowdown and refill were performed. The presence of multidimensional and parallel-channel effects was found to be very beneficial to BWR LOCA performance. The best estimate TRAC-BWR computer code was extended as part of this study by incorporation of a phenomenological upper plenum mixing model. TRAC-BWR was applied to the analysis of these full scale experiments. Excellent predictions of phenomena and experimental trends were achieved. (orig.)

  15. Design of the solid target structure and the study on the coolant flow distribution in the solid target using the 2-dimensional flow analysis

    International Nuclear Information System (INIS)

    Haga, Katsuhiro; Terada, Atsuhiko; Ishikura, Shuichi; Teshigawara, Makoto; Kinoshita, Hidetaka; Kobayashi, Kaoru; Kaminaga, Masaki; Hino, Ryutaro; Susuki, Akira

    1999-11-01

    A solid target cooled by heavy water is presently under development under the Neutron Science Research Project of the Japan Atomic Energy Research Institute (JAERI). Target plates of several millimeters thickness made of heavy metal are used as the spallation target material and they are put face to face in a row with one to two millimeters gaps in between though which heavy water flows, as the coolant. Based on the design criteria regarding the target plate cooling, the volume percentage of the coolant, and the thermal stress produced in the target plates, we conducted thermal and hydraulic analysis with a one dimensional target plate model. We choosed tungsten as the target material, and decided on various target plate thicknesses. We then calculated the temperature and the thermal stress in the target plates using a two dimensional model, and confirmed the validity of the target plate thicknesses. Based on these analytical results, we proposed a target structure in which forty target plates are divided into six groups and each group is cooled using a single pass of coolant. In order to investigate the relationship between the distribution of the coolant flow, the pressure drop, and the coolant velocity, we conducted a hydraulic analysis using the general purpose hydraulic analysis code. As a result, we realized that an uniform coolant flow distribution can be achieved under a wide range of flow velocity conditions in the target plate cooling channels from 1 m/s to 10 m/s. The pressure drop along the coolant path was 0.09 MPa and 0.17 MPa when the coolant flow velocity was 5 m/s and 7 m/s respectively, which is required to cool the 1.5 MW and 2.5 MW solid targets. (author)

  16. Numerical simulation on coolant flow and heat transfer in core

    International Nuclear Information System (INIS)

    Yao Zhaohui; Wang Xuefang; Shen Mengyu

    1997-01-01

    To simulate the coolant flow and the heat transfer characteristics of a core, a computer code, THAPMA (Thermal Hydraulic Analysis Porous Medium Analysis) has been developed. In THAPMA code, conservation equations are based on a porous-medium formulation, which uses four parameters, i.e, volume porosity, directional surface porosity, distributed resistance, and distributed heat source (sink), to model the effects of fuel rods and other internal solid structures on flow and heat transfer. Because the scheme and the solution are very important in accuracy and speed of calculation, a new difference scheme (WSUC) has been used in the energy equation, and a modified PISO solution method have been employed to simulate the steady/transient states. The code has been proved reliable and can effectively solve the transient state problem by several numerical tests. According to the design of Qinshan NPP-II, the flow and heat transfer phenomena in reactor core have been numerically simulated. The distributions of the velocity and the temperature can provide a theoretical basis for core design and safety analysis

  17. Transient behaviour of main coolant pump in nuclear power plants

    International Nuclear Information System (INIS)

    Delja, A.

    1986-01-01

    A basic concept of PWR reactor coolant pump thermo-hydraulic modelling in transient and accident operational condition is presented. The reactor coolant pump is a component of the nuclear steam supply system which forces the coolant through the reactor and steam generator, maintaining design heat transfer condition. The pump operating conditions have strong influence on the flow and thermal behaviour of NSSS, both in the stationary and nonstationary conditions. A mathematical model of the reactor coolant pump is formed by using dimensionless homologous relations in the four-quadrant regimes: normal pump, turbine, dissipation and reversed flow. Since in some operational regimes flow of mixture, liquid and steam may occur, the model has additional correction members for two-phase homologous relations. Modular concept has been used in developing computer program. The verification is performed on the simulation loss of offsite power transient and obtained results are presented. (author)

  18. Transient two-phase performance of LOFT reactor coolant pumps

    International Nuclear Information System (INIS)

    Chen, T.H.; Modro, S.M.

    1983-01-01

    Performance characteristics of Loss-of-Fluid Test (LOFT) reactor coolant pumps under transient two-phase flow conditions were obtained based on the analysis of two large and small break loss-of-coolant experiments conducted at the LOFT facility. Emphasis is placed on the evaluation of the transient two-phase flow effects on the LOFT reactor coolant pump performance during the first quadrant operation. The measured pump characteristics are presented as functions of pump void fraction which was determined based on the measured density. The calculated pump characteristics such as pump head, torque (or hydraulic torque), and efficiency are also determined as functions of pump void fractions. The importance of accurate modeling of the reactor coolant pump performance under two-phase conditions is addressed. The analytical pump model, currently used in most reactor analysis codes to predict transient two-phase pump behavior, is assessed

  19. Flow rate control systems for coolants for BWR type reactors

    International Nuclear Information System (INIS)

    Igarashi, Yoko; Kato, Naoyoshi.

    1981-01-01

    Purpose: To increase spontaneous recycling flow rate of coolants in BWR type reactors when the water level in the reactor decreases, by communicating a downcomer with a lower plenum. Constitution: An opening is provided to the back plate disposed at the lower end of a reactor core shroud for communicating a downcomer with a lower plenum, and an ON-OFF valve actuated by an operation rod is provided to the opening. When abnormal water level or pressure in the reactor is detected by a level metal or pressure meter, the operation rod is driven to open the ON-OFF valve, whereby coolants fed from a jet pump partially flows through the opening to increase the spontaneous recycling flow rate of the coolants. This can increase the spontaneous recycling flow rate of the coolants upon spontaneous recycling operation, thereby maintaining the reactor safety and the fuel soundness. (Moriyama, K.)

  20. Effect of removal of a central thimble on coolant flow distribution in a research reactor fuel element

    International Nuclear Information System (INIS)

    Green, W.J.

    1977-01-01

    Using two twice full-size models of a HIFAR research reactor fuel element, experiments have been performed to determine how the flow distribution of coolant gas through the element in a transfer flask is affected by removal of the central instrumentation thimble. With the thimble present, experimental flow results agree with theoretical predictions. Over the range of total flowrates considered, mass flow apportioning among the five annular channels was independent of annular channel Reynolds number (in the range 3500 to 10,500) and ranged between 13% and 27% of the total flowrate. For the case with the thimble removed, interesting experimental flow characteristics were obtained which could not have been predicted. Flow apportioning among the annular channels was found to be uniquely dependent upon total flowrate and ranged between 3% and 8% for the experimental conditions investigated (annular channel Reynolds numbers in the range 800 to 4000). (Author)

  1. Fusion-reactor blanket and coolant material compatibility

    International Nuclear Information System (INIS)

    Jeppson, D.W.; Keough, R.F.

    1981-01-01

    Fusion reactor blanket and coolant compatibility tests are being conducted to aid in the selection and design of safe blanket and coolant systems for future fusion reactors. Results of scoping compatibility tests to date are reported for blanket material and water interactions at near operating temperatures. These tests indicate the quantitative hydrogen release, the maximum temperature and pressures produced and the rates of interactions for selected blanket materials

  2. Review on research of small break loss of coolant accident

    International Nuclear Information System (INIS)

    Bo Jinhai; Wang Fei

    1998-01-01

    The Small Break Loss of Coolant Accident (SBLOCA) and its research art-of -work are reviewed. A typical SBLOCA process in Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) and the influence of break size, break location and reactor coolant pump on the process are described. The existing papers are classified in two categories: experimental and numerical modeling, with the primary experimental apparatuses in the world listed and the research works on SBLOCA summarized

  3. Laboratory simulation of rod-to-rod mechanical interactions during postulated loss-of-coolant accidents in a PWR involving cladding oxidation

    International Nuclear Information System (INIS)

    Hindle, E.D.; Haste, T.J.; Harrison, W.R.

    1987-01-01

    Creep deformation of Zircaloy cladding in postulated PWR loss-of-coolant accidents may lead to rod-to-rod mechanical interactions. Tests have been performed in the electrically heated FOURSQUARE rig at 750 0 C and 850 0 C in steam to investigate this effect. Conservatisms inherent in a simple 'square with rounded corners' coolant channel blockage model have been quantified; about 5-10% flow area may remain even at strains which in ideal circumstances would give total blockage. Reduction of average burst strains produced by an oxide layer (up to 13 μm) has been demonstrated, resulting from strain concentration at oxide cracks. (author)

  4. Comparative analysis of coolants for FBR of future nuclear power

    International Nuclear Information System (INIS)

    Toshinsky, G.I.; Grigoryev, O.G.; Pylchenkov, E.H.; Skorikov, D.E.; Komkova, O.I.

    2001-01-01

    Selection of a fast reactor (FR) coolant for future nuclear reactors is a complex task that has not a single solution. Safety requirements are expected to grow in the future. The requirements to FR are reconsidered. Gradual transition from the FR as a builder up of plutonium to the FR as an economically effective energy source, is taking place. Among all types of coolants viable for FR, LMC (light molten salt coolants) cover the most complete range of requirements to advanced reactors and have a complete database. Sodium and lead-bismuth coolant (LBC) are selected because there is a complete package of technologies for their handling. Heavy liquid metal coolant (HLMC), being at a disadvantage of heat transfer rate in relation to sodium, make it possible to give the inherent safety properties to the reactor and, as a result, to simplify essentially the reactor design and its safety systems. This results in capital and costs reduction. Neutronic characteristics of HLMC cooled reactors make possible to transmute their own minor actinides (MA) safely, and LBC cooled reactors are able to transmute LWR'MA with high safety characteristics. Basing on the comparison carried out, it can be concluded, that both LBC and sodium are perspective coolants for future FR

  5. Comparative analysis of coolants for FBR of future nuclear power

    Energy Technology Data Exchange (ETDEWEB)

    Toshinsky, G.I.; Grigoryev, O.G.; Pylchenkov, E.H.; Skorikov, D.E.; Komkova, O.I. [State Scientific Center of Russian Federation, Institute for Physics and Power Engineering named after Academician A.I. Leipusky, Kaluga Region (Russian Federation)

    2001-07-01

    Selection of a fast reactor (FR) coolant for future nuclear reactors is a complex task that has not a single solution. Safety requirements are expected to grow in the future. The requirements to FR are reconsidered. Gradual transition from the FR as a builder up of plutonium to the FR as an economically effective energy source, is taking place. Among all types of coolants viable for FR, LMC (light molten salt coolants) cover the most complete range of requirements to advanced reactors and have a complete database. Sodium and lead-bismuth coolant (LBC) are selected because there is a complete package of technologies for their handling. Heavy liquid metal coolant (HLMC), being at a disadvantage of heat transfer rate in relation to sodium, make it possible to give the inherent safety properties to the reactor and, as a result, to simplify essentially the reactor design and its safety systems. This results in capital and costs reduction. Neutronic characteristics of HLMC cooled reactors make possible to transmute their own minor actinides (MA) safely, and LBC cooled reactors are able to transmute LWR'MA with high safety characteristics. Basing on the comparison carried out, it can be concluded, that both LBC and sodium are perspective coolants for future FR.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  7. Technological status of reactor coolant pumps in generation III+ pressurized nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Brecht, Bernhard; Bross, Stephan [KSB Aktiengesellschaft, Frankenthal (Germany)

    2016-05-15

    KSB has been developing and producing pumps for thermal power plants for nearly 90 years. Consequently, KSB also started to develop and manufacture pumps for all kinds of nuclear power plants from the very beginning of the civil use of nuclear energy. This is especially true for reactor coolant pumps for pressurized water reactors. For the generation of advanced evolutionary reactors (Generation III+ reactors), KSB developed an advanced shaft seal system which is also able to fulfill the requirements of station blackout conditions. The tests in the KSB test rigs, which were successfully completed in December 2015, proved the full functionality of the new design. For generation III+ passive plant reactors KSB developed a new reactor coolant pump type called RUV, which is based on the experience of classic reactor coolant pumps and reactor internal pumps. It is a very compact, hermetically sealed vertical pump-motor unit with a wet winding motor. A full scale prototype successfully passed the 1st stage qualification test program in October 2015.

  8. Mechanistic study of fuel freezing, channel plugging, and continued coolability during fast reactor overpower excursions

    International Nuclear Information System (INIS)

    Wong, K.W.; Catton, I.; Kastenberg, W.E.

    1977-07-01

    A mechanistic model is presented which describes events following fuel pin failure which may lead to in-channel fuel plate-out. The thermal and hydraulic effects of the plate-out fuel are also evaluated. Given the amount and particle size of the fuel injected into the coolant channel during fuel pin failure, and the initial conditions of the interaction zone, the physical states of the fuel particles and the coolant in the interaction zone can be determined. The trajectories of the fuel particles in the coolant channel are determined by assuming a slip factor between the local tangential velocities of the coolant and the fuel particles. The time and distance after which a fuel particle hits a wire wrap are then determined and the impact stresses induced in the thin solid fuel crust can be evaluated

  9. Simulation of steam explosion in stratified melt-coolant configuration

    International Nuclear Information System (INIS)

    Leskovar, Matjaž; Centrih, Vasilij; Uršič, Mitja

    2016-01-01

    Highlights: • Strong steam explosions may develop spontaneously in stratified configurations. • Considerable melt-coolant premixed layer formed in subcooled water with hot melts. • Analysis with MC3D code provided insight into stratified steam explosion phenomenon. • Up to 25% of poured melt was mixed with water and available for steam explosion. • Better instrumented experiments needed to determine dominant mixing process. - Abstract: A steam explosion is an energetic fuel coolant interaction process, which may occur during a severe reactor accident when the molten core comes into contact with the coolant water. In nuclear reactor safety analyses steam explosions are primarily considered in melt jet-coolant pool configurations where sufficiently deep coolant pool conditions provide complete jet breakup and efficient premixture formation. Stratified melt-coolant configurations, i.e. a molten melt layer below a coolant layer, were up to now believed as being unable to generate strong explosive interactions. Based on the hypothesis that there are no interfacial instabilities in a stratified configuration it was assumed that the amount of melt in the premixture is insufficient to produce strong explosions. However, the recently performed experiments in the PULiMS and SES (KTH, Sweden) facilities with oxidic corium simulants revealed that strong steam explosions may develop spontaneously also in stratified melt-coolant configurations, where with high temperature melts and subcooled water conditions a considerable melt-coolant premixed layer is formed. In the article, the performed study of steam explosions in a stratified melt-coolant configuration in PULiMS like conditions is presented. The goal of this analytical work is to supplement the experimental activities within the PULiMS research program by addressing the key questions, especially regarding the explosivity of the formed premixed layer and the mechanisms responsible for the melt-water mixing. To

  10. Lower internals for pressurized water reactor

    International Nuclear Information System (INIS)

    Chevereau, G.; Babin, M.

    1989-01-01

    The lower internals for PWR has a separating plate mounted beneath its lower core plate and defining a distribution chamber with it, peripheral mechanical connectors joining the plates separated by coolant passage and apertures in the separation plate connected to a coolant pipe [fr

  11. Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

    Science.gov (United States)

    Jones, Robert D.

    1978-01-01

    A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level.

  12. Apparatus for controlling coolant level in a liquid-metal-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Jones, R.D.

    1978-01-01

    A liquid-metal-cooled fast-breeder reactor which has a thermal liner spaced inwardly of the pressure vessel and includes means for passing bypass coolant through the annulus between the thermal liner and the pressure vessel to insulate the pressure vessel from hot outlet coolant includes control ports in the thermal liner a short distance below the normal operating coolant level in the reactor and an overflow nozzle in the pressure vessel below the control ports connected to an overflow line including a portion at an elevation such that overflow coolant flow is established when the coolant level in the reactor is above the top of the coolant ports. When no makeup coolant is added, bypass flow is inwardly through the control ports and there is no overflow; when makeup coolant is being added, coolant flow through the overflow line will maintain the coolant level

  13. Experimental approach to investigate the dynamics of mixing coolant flow in complex geometry using PIV and PLIF techniques

    Directory of Open Access Journals (Sweden)

    Hutli Ezddin

    2015-01-01

    Full Text Available The aim of this work is to investigate experimentally the increase of mixing phenomenon in a coolant flow in order to improve the heat transfer, the economical operation and the structural integrity of Light Water Reactors-Pressurized Water Reactors (LWRs-PWRs. Thus the parameters related to the heat transfer process in the system will be investigated. Data from a set of experiments, obtained by using high precision measurement techniques, Particle Image Velocimetry and Planar Laser-Induced Fluorescence (PIV and PLIF, respectively are to improve the basic understanding of turbulent mixing phenomenon and to provide data for CFD code validation. The coolant mixing phenomenon in the head part of a fuel assembly which includes spacer grids has been investigated (the fuel simulator has half-length of a VVER 440 reactor fuel. The two-dimensional velocity vector and temperature fields in the area of interest are obtained by PIV and PLIF technique, respectively. The measurements of the turbulent flow in the regular tube channel around the thermocouple proved that there is rotation and asymmetry in the coolant flow caused by the mixing grid and the geometrical asymmetry of the fuel bundle. Both PIV and PLIF results showed that at the level of the core exit thermocouple the coolant is homogeneous. The discrepancies that could exist between the outlet average temperature of the coolant and the temperature at in-core thermocouple were clarified. Results of the applied techniques showed that both of them can be used as good provider for data base and to validate CFD results.

  14. SSYST-1. A computer code system to analyse the fuel rod behaviour during a loss of coolant accident

    International Nuclear Information System (INIS)

    Gulden, W.

    1977-08-01

    The modules of the SSYST program system allow the detailed analysis of an LWR fuel rod in the course of a postulated loss-of-coolant accident. They provide a tool for considering the interaction between the heat conduction in the fuel rod, heat transfer in the gap, fuel and cladding tube deformation, pressure in the coolant, as well as thermal and fluid dynamics in the cooling channel and for calculating the time and location of ballooning and rod failure, respectively. They can be used both to precalculate the behaviour of fuel rods during LWR accidents and in support of the design of experiments. Depending on the problem to be solved, the individual modules can be easily combined. (orig.) [de

  15. Experimental determination of heat transfer critical conditions in water forced convection at low pressure in a circular channel

    International Nuclear Information System (INIS)

    Fernandes, M.P.

    1973-02-01

    An experimental determination was made of heat transfer critical conditions in a circular channel, uniformly heated, and internally cooled by water in ascending forced convection, under a pressure slightly above atmospheric pressure. Measurements were made of water flow, pressure, electric power temperature and heating, and a systematic analysis was made of the system's parameters. The values obtained for the heat critical flux are circa 50% lower than those predicted by Becker and Biasi and this is accounted to flowing instabilities of thermo-hydrodynamic nature. It is suggested that the flowing channels of circuits aiming at the study of the boiling crisis phenomenon be expanded in its upper extremity, and that the coolant circulation be kept through a pump with a pressure X flow characteristic as vertical as possible

  16. Liquid metal coolant flow rate regulation

    International Nuclear Information System (INIS)

    Vitkovskij, I.V.; Glukhikh, V.A.; Kirillov, I.R.; Smirnov, A.M.

    1981-01-01

    Some aspects of fast reactor and experimental bench operation related to liquid metal flow rate regulation are considered. Requirements to the devices for the flow rate regulation are formulated. A new type of these devices namely magnetohydrodynamic (MHD) throttles is described. Structural peculiarities of MHD throttles of different types are described as well. It is noted that the MHD throttles with a screw channel have the best energy mass indices. On the basis of the comparison of the MHD throttles with mechanical valves it is concluded that the MHD throttles described are useful for regulating the flow rates of any working media. Smoothness and accuracy of the flow rate regulation by the throttles are determined by the electric control circuit and may be practically anyone. The total coefficient of hydraulic losses in the throttle channel in the absence of a magnetic field is ten and more times lesser than in completely open mechanical valve. Electromagnetic time constant of the MHD throttles does not exceed several tenths of a second [ru

  17. The effect of constraint on fuel-coolant interactions in a confined geometry

    Energy Technology Data Exchange (ETDEWEB)

    Park, H.; Corradini, M.L. [Univ. of Wisconsin, Madison, WI (United States)

    1995-09-01

    A Fuel-Coolant Interaction (FCI or vapor explosion) is the phenomena in which a hot liquid rapidly transfers its internal energy into a surrounding colder and more volatile liquid. The energetics of such a complex multi-phase and multi-component phenomenon is partially determined by the surrounding boundary conditions. As one of the boundary conditions, we studied the effect of constraint on FCIs. The WFCI-D series of experiments were performed specifically to observe this effect. The results from these and our previous WFCI tests as well as those of other investigators are compared.

  18. Characteristic study of a novel compact Solar Thermal Facade (STF) with internally extruded pin–fin flow channel for building integration

    International Nuclear Information System (INIS)

    Shen, Jingchun; Zhang, Xingxing; Yang, Tong; Tang, Llewellyn; Cheshmehzangi, Ali; Wu, Yupeng; Huang, Guiqin; Zhong, Dan; Xu, Peng; Liu, Shengchun

    2016-01-01

    Highlights: • A novel compact STF with internally extruded pin–fin flow channel is proposed. • Thermal performance of the STF is characterized in different operation modes. • This STF is with simple structure, low cost and high feasibility in building design. • Such STF can achieve better thermal performance as conventional ones. - Abstract: The fully building integrated Solar Thermal Facade (STF) systems can become potential solutions for aesthetics architectural design, as well as for the enhancement of energy efficiency and reduction of operational cost in the contemporary built environment. As a result, this article introduces a novel compact STF with internally extruded pin–fin flow channel that is particularly suitable for the building integration. A dedicated simulation model was developed on basis of the heat transfer and the flow mechanics. A prototype of this STF was fabricated and then it was tested under a series of controlled environmental conditions. The experimental validation illustrated a good agreement with the simulation results, indicating the established model was able to predict the STF’s thermal performance at a reasonable accuracy (i.e. mean deviation of less than 5.46%). The impacts of several operational parameters, i.e. equivalent solar radiation, air temperature, air velocity, water mass flow rate and inlet water temperature, on the STF thermal performance were then discussed respectively. Given the baseline testing condition, the collector efficiency factor F′ is almost 0.9930, leading to a relatively high nominal thermal efficiency at about 63.21%, which demonstrates such STF, with simpler structure, lower cost and higher feasibility in architectural design, can achieve an equivalent or better thermal performance than recent bionic STF or the conventional ones. It is also concluded that the thermal efficiency varies proportionally with solar radiation, air temperature, and mass flow rate of water, but oppositely to air

  19. Primary coolant recycling device for FBR type reactor

    International Nuclear Information System (INIS)

    Kanbe, Mitsuru; Tokiwai, Moriyasu

    1998-01-01

    A primary coolants (liquid sodium) recycling device comprises a plurality of recycling pumps. The recycling pumps are operated while using, as a power source, electric power generated by a thermoelectric power generation system by utilizing heat stored in the coolants. The thermoelectric power generation system comprises a thermo-electric conversion module, heat collecting heat pipes as a high temperature side heat conduction means and heat dissipating pipes as a low temperature side heat conduction means. The heat of coolants is transferred to the surface of the high temperature side of each thermo-electric conversion elements of the thermal power generation system by the heat collecting heat pipes. The heat on the low temperature side of each of the thermo-electric conversion elements is removed by the heat dissipating pipes. Accordingly, temperature difference is caused between both surfaces of the thermo-electric conversion elements. Even upon loss of a main power source due to stoppage of electricity, electric power is generated by utilizing heat of coolants, so that the recycling pumps circulate coolants to cool a reactor core continuously. (I.N.)

  20. Analyses of Decrease in Reactor Coolant Flow Rate in SMART

    International Nuclear Information System (INIS)

    Kim, Hyung Rae; Bae, Kyoo Hwan; Choi, Suhn

    2011-01-01

    SMART is a small integral reactor, which is under development at KAERI to get the standard design approval by the end of 2011. SMART works like a pressurized light-water reactor in principle though it is more compact than large commercial reactors. SMART houses major components such as steam generators, a pressurizer, and reactor coolant pumps inside the reactor pressure vessel. Due to its compact design, SMART adopts a canned-motor type reactor coolant pump which has much smaller rotational inertia than the ones used in commercial reactors. As a consequence, the reactor coolant pump has very short coastdown time and reactor coolant flow rate decreases more severely compared to commercial reactors. The transients initiated by reduction of reactor coolant flow rate have been analyzed to ensure that SMART can be safely shutdown on such transients. The design basis events in this category are complete loss of flow, single pump locked rotor with loss of offsite power, and single pump shaft break with loss of offsite power

  1. Full reactor coolant system chemical decontamination qualification programs

    Energy Technology Data Exchange (ETDEWEB)

    Miller, P.E. [Westinghouse Electric Corp., Pittsburgh, PA (United States)

    1995-03-01

    Corrosion and wear products are found throughout the reactor coolant system (RCS), or primary loop, of a PWR power plant. These products circulate with the primary coolant through the reactor where they may become activated. An oxide layer including these activated products forms on the surfaces of the RCS (including the fuel elements). The amount of radioactivity deposited on the different surface varies and depends primarily on the corrosion rate of the materials concerned, the amount of cobalt in the coolant and the chemistry of the coolant. The oxide layer, commonly called crud, on the surfaces of nuclear plant systems leads to personnel radiation exposure. The level of the radiation fields from the crud increases with time from initial plant startup and typically levels off after 4 to 6 cycles of plant operation. Thereafter, significant personnel radiation exposure may be incurred whenever major maintenance is performed. Personnel exposure is highest during refueling outages when routine maintenance on major plant components, such as steam generators and reactor coolant pumps, is performed. Administrative controls are established at nuclear plants to minimize the exposure incurred by an individual and the plant workers as a whole.

  2. The installation welding of pressure water reactor coolant piping

    International Nuclear Information System (INIS)

    Deng Feng

    2010-01-01

    Large pressure water reactor nuclear power plants are constructing in our country. There are three symmetry standard loops in reactor coolant system. Each loop possesses a steam generator and a primary poop, in which one of the loops is equipped with a pressurizer. These components are connected with reactor pressure vessel by installation welding of the coolant piping. The integrity of reactor coolant pressure boundary is the second barrier to protect the radioactive substance from release to outside, so the safe operation of nuclear power plant is closely related to the quality of coolant piping installation welding. The heavy tube with super low carbon content austenitic stainless steel is selected for coolant piping. This kind of material has good welding behavior, but the poor thermal conductivity, the big liner expansion coefficient and the big welding deformation will cause bigger welding stress. To reduce the welding deformation, to control the dimension precision, to reduce the residual stress and to ensure the welding quality the installation sequence should be properly designed and the welding technology should be properly controlled. (authors)

  3. International Union of Basic and Clinical Pharmacology. C. Nomenclature and Properties of Calcium-Activated and Sodium-Activated Potassium Channels.

    Science.gov (United States)

    Kaczmarek, Leonard K; Aldrich, Richard W; Chandy, K George; Grissmer, Stephan; Wei, Aguan D; Wulff, Heike

    2017-01-01

    A subset of potassium channels is regulated primarily by changes in the cytoplasmic concentration of ions, including calcium, sodium, chloride, and protons. The eight members of this subfamily were originally all designated as calcium-activated channels. More recent studies have clarified the gating mechanisms for these channels and have documented that not all members are sensitive to calcium. This article describes the molecular relationships between these channels and provides an introduction to their functional properties. It also introduces a new nomenclature that differentiates between calcium- and sodium-activated potassium channels. Copyright © 2016 by The American Society for Pharmacology and Experimental Therapeutics.

  4. The analysis of coolant-velocity distribution in plat-typed fuel element using CFD method for RSG-GAS research reactor

    International Nuclear Information System (INIS)

    Muhammad Subekti; Darwis Isnaini; Endiah Puji Hastuti

    2013-01-01

    The measurement experiment for coolant-velocity distribution in the subchannel of fuel element of RSG-GAS research reactor is difficult to be carried out due to too narrow channel and subchannel placed inside the fuel element. Hence, the calculation is required to predict the coolant-velocity distribution inside subchannel to confirm that the handle presence does not ruin the velocity distribution into every subchannel. This calculation utilizes CFD method, which respect to 3-dimension interior. Moreover, the calculation of coolant-velocity distribution inside subchannel was not ever carried out. The research object is to investigate the distribution of coolant-velocity in plat-typed fuel element using 3-dimension CFD method for RSG-GAS research reactor. This research is required as a part of the development of thermalhydraulic design of fuel element for innovative research reactor as well. The modeling uses ½ model in Gambit software and calculation uses turbulence equation in FLUENT 6.3 software. Calculation result of 3D coolant-velocity in subchannel using CFD method is lower about 4.06 % than 1D calculation result due to 1D calculation obeys handle availability. (author)

  5. Some experimental justifications of constructions of nuclear reactors with the use of solid coolant

    International Nuclear Information System (INIS)

    Deniskin, V.; Nalivaev, V.; Fedik, I.; Vishnevski, U.; Dmitriev, A.

    2003-01-01

    was 0.5 to 1.3 mm and they were of different quality. The best results showed particles from the batch 5 which were carbon pyrocoated ones with gas-core settling. At present we have received the data on heat transfer flow of the solid coolant. The data depend on temperature of a heating wall (up to 800 deg. C), the velocity of a continuous particle flow (the velocity for our geometry is up to 0.22 m/s), gas environ (helium, argon, vacuum), diameter and degree of particle perfection. The heat transfer coefficient reached the value of 800-1000 MW/m 2 K in the channel with round cross-section diameter of 10mm without turbulence. The coefficient was reached in the best conditions (helium, velocity of 0.2 m/s, temperature of a wall 800 deg. C, a batch of the carbon pyrocoated particles). The tests of particles of the given batch were conducted during 1000 hours at temperature of 300-400 deg. C and with flow velocity of ∼ 0,1 m/s and they didn't reveal any destruction. It was recorded only a small change of geometrical average sizes and feeble increase of a surface roughness of particles within the limits of 1-3 microns. This is the result of surface corrosion and mechanical wearing. The results of the researches formed an invention formula and the patent was issued in Russia in 05.10.2001. The formula and description of the invention give the description of requirements, which allow to conduct the reliable uniform motion of particles with a high heat transfer coefficient, range of diameters of particles, degree of particle nonsphericity, carbon pyrocoating, and also characteristic specific devices necessary for construction of such a reactor. The results obtained were used to conduct the estimations of parameters of nuclear power reactors. The parameters of a possible nuclear reactor are given below (Power - 3750 MW, electrical power - 1500 MW). At that average electrical power will be ∼ 8.65 MW/m 3 in the core, and maximum will be 15.5 MW/m 3 . The basic advantages of

  6. Compensation of equipment housing elements of reactor units with heavy liquid metal coolant vessel temperature deformations

    International Nuclear Information System (INIS)

    Lebedevich, V.; Ahmetshin, M.; Mendes, D.; Kaveshnikov, S.; Vinogradov, A.

    2015-01-01

    In Russia a lot of different versions of fast reactors (FRs) are investigated and one of these is FR cooled by liquid lead and liquid lead-bismuth alloy. In this poster we are interested by FR with concrete vessel; its components are placed in cavities inside the vessel, and connected by a channel system. During the installation the equipment components are placed in several equipment housings. Between these housings there are cavities with coolant. The alignment of the housings should be provided. It can be broken by irregular concrete vessel heating during FR starting or other transition regimes. Our goal is to suggest a list of designing steps to compensate temperature deformations of equipment housing elements. A simplified model of equipment housing was suggested. It consists of two cylinders - tunnels in the concrete vessel, separated by a cavity filled by coolant and inert gas. The bottom part was considered as heated to 420 C. degrees while in the top part temperature decreased to 45 C. degrees (on the concrete surface). According to this data, results show that temperature gradient leads to a concrete layer dislocation of about 12.5 mm, which can lead to damage and breaking alignment. We propose the following solution to compensate for temperature deformation: -) to chisel out part of the upper top of the insulating concrete; -) to install an adequate misalignment of equipment housing elements preliminary; and -) to use a torsion system like a piston-type device for providing additional strength in order to compensate deformation and vibrations

  7. Analysis of Loss-of-Coolant Accidents in the NIST Research Reactor - Early Phase

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Joo S.; Diamond, David

    2016-12-06

    A study of the fuel temperature during the early phase of a loss-of-coolant accident (LOCA) in the NIST research reactor (NBSR) was completed. Previous studies had been reported in the preliminary safety analysis report for the conversion of the NBSR from high-enriched uranium (HEU) fuel to low-enriched (LEU) fuel. Those studies had focused on the most vulnerable LOCA situation, namely, a double-ended guillotine break in the time period after reactor trip when water is drained from either the coolant channels inside the fuel elements or the region outside the fuel elements. The current study fills in a gap in the analysis which is the early phase of the event when there may still be water present but the reactor is at power or immediately after reactor trip and pumps have tripped. The calculations were done, for both the current HEU-fueled core and the proposed LEU core, with the TRACE thermal-hydraulic systems code. Several break locations and different break sizes were considered. In all cases the increase in the clad (or fuel meat) temperature was relatively small so that a large margin to the temperature threshold for blistering (the Safety Limit for the NBSR) remained.

  8. Radiation leakage monitoring method and device from primary to secondary coolant systems in nuclear reactor

    International Nuclear Information System (INIS)

    Tajiri, Yoshiaki; Umehara, Toshihiro; Yamada, Masataka.

    1993-01-01

    The present invention monitors radiation leaked from any one of primary cooling systems to secondary cooling systems in a plurality of steam generators. That is, radiation monitoring means each corresponding to steam each generators are disposed to the upstream of a position where main steam pipes are joined. With such a constitution, since the detection object of each of radiation monitoring means is secondary coolants before mixing with secondary coolants of other secondary loops or dilution, lowering of detection accuracy can be avoided. Except for the abnormal case, that is, a case neither of radiation leakage nor of background change, the device is adapted as a convenient measuring system only with calculation performance. Once abnormality occurs, a loop having a value exceeding a standard value is identified by a single channel analyzer function. The amount of radiation leakage from the steam generator belonging to the specified loop is monitored quantitatively by a multichannel analyzer function. According to the method of the present invention, since specific spectrum analysis is conducted upon occurrence of abnormality, presence of radiation leakage and the scale thereof can be judged rapidly. (I.S.)

  9. Influence of coolant motion on structure of hardened steel element

    Directory of Open Access Journals (Sweden)

    A. Kulawik

    2008-08-01

    Full Text Available Presented paper is focused on volumetric hardening process using liquid low melting point metal as a coolant. Effect of convective motion of the coolant on material structure after hardening is investigated. Comparison with results obtained for model neglecting motion of liquid is executed. Mathematical and numerical model based on Finite Element Metod is described. Characteristic Based Split (CBS method is used to uncouple velocities and pressure and finally to solve Navier-Stokes equation. Petrov-Galerkin formulation is employed to stabilize convective term in heat transport equation. Phase transformations model is created on the basis of Johnson-Mehl and Avrami laws. Continuous cooling diagram (CTPc for C45 steel is exploited in presented model of phase transformations. Temporary temperatures, phases participation, thermal and structural strains in hardening element and coolant velocities are shown and discussed.

  10. Actively controlling coolant-cooled cold plate configuration

    Science.gov (United States)

    Chainer, Timothy J.; Parida, Pritish R.

    2015-07-28

    A method is provided to facilitate active control of thermal and fluid dynamic performance of a coolant-cooled cold plate. The method includes: monitoring a variable associated with at least one of the coolant-cooled cold plate or one or more electronic components being cooled by the cold plate; and dynamically varying, based on the monitored variable, a physical configuration of the cold plate. By dynamically varying the physical configuration, the thermal and fluid dynamic performance of the cold plate are adjusted to, for example, optimally cool the one or more electronic components, and at the same time, reduce cooling power consumption used in cooling the electronic component(s). The physical configuration can be adjusted by providing one or more adjustable plates within the coolant-cooled cold plate, the positioning of which may be adjusted based on the monitored variable.

  11. Sloshing of coolant in a seismically isolated reactor

    International Nuclear Information System (INIS)

    Wu, T.S.; Guildys, J.; Seidensticker, R.W.

    1988-01-01

    During a seismic event, the liquid coolant inside the reactor vessel has sloshing motion which is a low-frequency phenomenon. In a reactor system incorporated with seismic isolation, the isolation frequency usually is also very low. There is concern on the potential amplification of sloshing motion of the liquid coolant. This study investigates the effects of seismic isolation on the sloshing of liquid coolant inside the reactor vessel of a liquid metal cooled reactor. Based on a synthetic ground motion whose response spectra envelop those specified by the NRC Regulator Guide 1.60, it is found that the maximum sloshing wave height increases from 18 in. to almost 30 in. when the system is seismically isolated. Since higher sloshing wave may introduce severe impact forces and thermal shocks to the reactor closure and other components within the reactor vessel, adequate design considerations should be made either to suppress the wave height or to reduce the effects caused by high waves

  12. Fracture mechanics evaluation for at typical PWR primary coolant pipe

    International Nuclear Information System (INIS)

    Tanaka, T.; Shimizu, S.; Ogata, Y.

    1997-01-01

    For the primary coolant piping of PWRs in Japan, cast duplex stainless steel which is excellent in terms of strength, corrosion resistance, and weldability has conventionally been used. The cast duplex stainless steel contains the ferrite phase in the austenite matrix and thermal aging after long term service is known to change its material characteristics. It is considered appropriate to apply the methodology of elastic plastic fracture mechanics for an evaluation of the integrity of the primary coolant piping after thermal aging. Therefore we evaluated the integrity of the primary coolant piping for an initial PWR plant in Japan by means of elastic plastic fracture mechanics. The evaluation results show that the crack will not grow into an unstable fracture and the integrity of the piping will be secured, even when such through wall crack length is assumed to equal the fatigue crack growth length for a service period of up to 60 years

  13. Radioactivity analysis of KAMINI reactor coolant from regulatory perspectives

    International Nuclear Information System (INIS)

    Srinivasan, T.K.; Sulthan, Bajeer; Sarangapani, R.; Jose, M.T.; Venkatraman, B.; Thilagam, L.

    2016-01-01

    KAMINI (a 30kWt) research reactor is operated for neutron radiography of fuel subassemblies and pyro devices and activation analysis of various samples. The reactor is fueled by 233 U and DM water is used as the coolant. During reactor operation, fission product noble gasses (FPNGs) such as 85m Kr, 87 Kr, 88 Kr, 135 Xe, 135m Xe and 138 Xe are detected in the coolant water. In order to detect clad failure, the water is sampled during reactor operation at regular intervals as per the technical specifications. In the present work, analysis of measured activities in coolant samples collected during reactor operation at 25 kWt are presented and compared with computed values obtained using ORIGEN (Isotope Generation) code

  14. Method of eliminating cruds in the primary coolants of reactors

    International Nuclear Information System (INIS)

    Tamura, Takaaki.

    1984-01-01

    Purpose: To eliminate cruds in the primary coolants by using rind of onions or peanuts. Method: Since cruds contained in the reactor primary coolants increase the radioactive exposure to reactor operators, they have been intended to remove by ion exchange resins. In this invention, rind of onions or peanuts are crushed into an adequate particle size and packed into an absorption column instead of ion exchange resins into which primary coolants are circulated. The powderous onions or peanuts rind contain glucoside such as cosmosiin and has an effect of cationic exchanger, they satisfactorily catch heavy metals such as Fe and Cu. They have an excellent filtering effect even under a high pH condition and are excellent in economical point of view. They can be decrease the volume of the absorption column, reduce their devolume after use through corrosion and easily subjected to waste procession through oxidizing combustion in liquid. (Nakamoto, H.)

  15. Physical model and calculation code for fuel coolant interactions

    International Nuclear Information System (INIS)

    Goldammer, H.; Kottowski, H.

    1976-01-01

    A physical model is proposed to describe fuel coolant interactions in shock-tube geometry. According to the experimental results, an interaction model which divides each cycle into three phases is proposed. The first phase is the fuel-coolant-contact, the second one is the ejection and recently of the coolant, and the third phase is the impact and fragmentation. Physical background of these phases are illustrated in the first part of this paper. Mathematical expressions of the model are exposed in the second part. A principal feature of the computational method is the consistent application of the fourier-equation throughout the whole interaction process. The results of some calculations, performed for different conditions are compiled in attached figures. (Aoki, K.)

  16. Fracture mechanics evaluation for at typical PWR primary coolant pipe

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, T. [Kansai Electric Power Company, Osaka (Japan); Shimizu, S.; Ogata, Y. [Mitsubishi Heavy Industries, Ltd., Kobe (Japan)

    1997-04-01

    For the primary coolant piping of PWRs in Japan, cast duplex stainless steel which is excellent in terms of strength, corrosion resistance, and weldability has conventionally been used. The cast duplex stainless steel contains the ferrite phase in the austenite matrix and thermal aging after long term service is known to change its material characteristics. It is considered appropriate to apply the methodology of elastic plastic fracture mechanics for an evaluation of the integrity of the primary coolant piping after thermal aging. Therefore we evaluated the integrity of the primary coolant piping for an initial PWR plant in Japan by means of elastic plastic fracture mechanics. The evaluation results show that the crack will not grow into an unstable fracture and the integrity of the piping will be secured, even when such through wall crack length is assumed to equal the fatigue crack growth length for a service period of up to 60 years.

  17. Coolant circuit water chemistry of the Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Tilky, Peter; Doma, Arpad

    1985-01-01

    The numerous advantages of the proper selection of water chemistry parameters including low corrosion rate of the structural materials, hence the low-level activity build-up, depositions, radiation doses were emphasized. Major characteristics of water chemistry applied to the primary coolant of pressurized water reactors including neutral, slightly basic and strong basic ones are discussed. Boric acid is widely used to control reactivity. Primary coolant water chemistry of WWER type reactors which is based on the addition of ammonia and potassium hydroxide to boric acid is compared with that of other reactors. The demineralization of the total condensate of the steam turbines became a general trend in the water chemistry of the secondary coolant circuits. (V.N.)

  18. LOFT advanced densitometer for nuclear loss-of-coolant experiments

    International Nuclear Information System (INIS)

    Johnson, L.O.; Lassahn, G.D.; Wood, D.B.

    1979-01-01

    A ''nuclear hardened'' gamma densitometer, a device which uses radiation attenuation to measure fluid density in the presence of a background radiation field, is described. Data from the nuclear hardened gamma densitometer are acquired by time sampling the coolant fluid piping and fluid attenuated source energy spectrum. The data are used to calculate transient coolant fluid cross sectional average density to analyze transient mass flow and other thermal-hydraulic characteristics during the Loss-of-Fluid Test (LOFT) loss-of-coolant experiments. The nuclear hardened gamma densitometer uses a pulse height analysis or energy discrimination, pulse counting technique which makes separation of the gamma radiation source signal from the reactor generated gamma radiation background noise signal possible by processing discrete pulses which retain their pulse amplitude information

  19. Experimental interaction of magma and “dirty” coolants

    Science.gov (United States)

    Schipper, C. Ian; White, James D. L.; Zimanowski, Bernd; Büttner, Ralf; Sonder, Ingo; Schmid, Andrea

    2011-03-01

    The presence of water at volcanic vents can have dramatic effects on fragmentation and eruption dynamics, but little is known about how the presence of particulate matter in external water will further alter eruptions. Volcanic edifices are inherently “dirty” places, where particulate matter of multiple origins and grainsizes typically abounds. We present the results of experiments designed to simulate non-explosive interactions between molten basalt and various “coolants,” ranging from homogeneous suspensions of 0 to 30 mass% bentonite clay in pure water, to heterogeneous and/or stratified suspensions including bentonite, sand, synthetic glass beads and/or naturally-sorted pumice. Four types of data are used to characterise the interactions: (1) visual/video observations; (2) grainsize and morphology of resulting particles; (3) heat-transfer data from a network of eight thermocouples; and (4) acoustic data from three force sensors. In homogeneous coolants with ~20% sediment, heat transfer is by forced convection and conduction, and thermal granulation is less efficient, resulting in fewer blocky particles, larger grainsizes, and weaker acoustic signals. Many particles are droplet-shaped or/and “vesicular,” containing bubbles filled with coolant. Both of these particle types indicate significant hydrodynamic magma-coolant mingling, and many of them are rewelded into compound particles. The addition of coarse material to heterogeneous suspensions further slows heat transfer thus reducing thermal granulation, and variable interlocking of large particles prevents efficient hydrodynamic mingling. This results primarily in rewelded melt piles and inefficient distribution of melt and heat throughout the coolant volume. Our results indicate that even modest concentrations of sediment in water will significantly limit heat transfer during non-explosive magma-water interactions. At high concentrations, the dramatic reduction in cooling efficiency and increase in

  20. Technical committee meeting on material-coolant interactions and material movement and relocation in liquid metal fast reactors

    International Nuclear Information System (INIS)

    1994-01-01

    The Technical Committee Meeting on Material-Coolant Interactions and Material Movement and Relocation in Liquid Metal Fast Reactors was sponsored by the International Working Group on Fast Reactors (IWGFR), International Atomic Energy Agency (IAEA) and hosted by PNC, on behalf of the Japanese government. A broad range of technical subjects was discussed in the TCM, covering entire aspects of material motion and interactions relevant to the safety of LMFRs. Recent achievement and current status in research and development in this area were presented including European out-of-pile test of molten material movement and relocation; molten material-sodium interaction; molten fuel-coolant interaction; core disruptive accidents; sodium boiling; post accident material relocation, heat removal and relevant experiments already performed or planned

  1. Technical committee meeting on material-coolant interactions and material movement and relocation in liquid metal fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-07-01

    The Technical Committee Meeting on Material-Coolant Interactions and Material Movement and Relocation in Liquid Metal Fast Reactors was sponsored by the International Working Group on Fast Reactors (IWGFR), International Atomic Energy Agency (IAEA) and hosted by PNC, on behalf of the Japanese government. A broad range of technical subjects was discussed in the TCM, covering entire aspects of material motion and interactions relevant to the safety of LMFRs. Recent achievement and current status in research and development in this area were presented including European out-of-pile test of molten material movement and relocation; molten material-sodium interaction; molten fuel-coolant interaction; core disruptive accidents; sodium boiling; post accident material relocation, heat removal and relevant experiments already performed or planned.

  2. Integrated Fuel-Coolant Interaction (IFCI 6.0) code

    International Nuclear Information System (INIS)

    Davis, F.J.; Young, M.F.

    1994-04-01

    The integrated Fuel-Coolant interaction (IFCI) computer code is being developed at Sandia National Laboratories to investigate the fuel-coolant interaction (FCI) problem at large scale using a two-dimensional, four-field hydrodynamic framework and physically based models. IFCI will be capable of treating all major FCI processes in an integrated manner. This document is a product of the effort to generate a stand-alone version of IFCI, IFCI 6.0. The User's Manual describes in detail the hydrodynamic method and physical models used in IFCI 6.0. Appendix A is an input manual, provided for the creation of working decks

  3. Reactor coolant pump shaft seal behavior during blackout conditions

    International Nuclear Information System (INIS)

    Mings, W.J.

    1985-01-01

    The United States Nuclear Regulatory Commission has classified the problem of reactor coolant pump seal failures as an unresolved safety issue. This decision was made in large part due to experimental results obtained from a research program developed to study shaft seal performance during station blackout and reported in this paper. Testing and analysis indicated a potential for pump seal failure under postulated blackout conditions leading to a loss of primary coolant with a concomitant danger of core uncovery. The work to date has not answered all the concerns regarding shaft seal failure but it has helped scope the problem and focus future research needed to completely resolve this issue

  4. Effects of different rod spacers (helical types) on coolant crossmixing

    International Nuclear Information System (INIS)

    Zhukov, A.V.; Sviridenko, E.Ya.; Matyukhin, N.M.; Rymkevich, K.S.; Ushakov, P.A.

    1981-11-01

    The results of investigations (electromagnetic measuring method) on coolant cross mixing in rod clusters with spiral wire spacers with different winding directions, with alternating unfinned and finned rods (case 'fin to rod'), as well as in rod clusters with much space between the rods, (case 'fin to fin') are reported. The local fluid dynamics parameters (distribution of the transversal and longitudinal velocity component) that define the physical processes of the coolant exchange in the rod clusters with helical spacers are explained. The investigation results for different helical spacer types are compared with each other. (orig.) [de

  5. The 1994 loss of coolant incident at Pickering NGS

    Energy Technology Data Exchange (ETDEWEB)

    Charlebois, P R; Clarke, T R; Goodman, R M; McEwan, W F [Ontario Hydro, Pickering, ON (Canada). Pickering Generating Station; Cuttler, J M [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

    1996-12-31

    Fracture of the rubber diaphragm in a liquid relief valve initiated events leading to a loss of coolant in Unit 2, on December 10. The valve failed open, filling the bleed condenser. The reactor shut itself down. When pressure recovered, two spring-loaded safety relief valves opened and one of them chattered. The shock and pulsations cracked the inlet pipe to the chattering valve, and the subsequent loss of coolant triggered the emergency core cooling system. The incident was terminated by operator action. No abnormal radioactivity was released. The four reactor units of Pickering A remained shut down until the corrective actions were completed in April/May 1995. (author). 4 figs.

  6. LWR and HTGR coolant dynamics: the containment of severe accidents

    International Nuclear Information System (INIS)

    Theofanous, T.G.; Gherson, P.; Nourbakhsh, H.P.; Hu, K.; Iyer, K.; Viskanta, R.; Lommers, L.

    1983-07-01

    This is the final report of a project containing three major tasks. Task I deals with the fundamental aspects of energetic fuel/coolant interactions (steam explosions) as they pertain to LWR core melt accidents. Task II deals with the applied aspects of LWR core melt accident sequences and mechanisms important to containment response, and includes consideration of energetic fuel/coolant interaction events, as well as non-explosive ones, corium material disposition and eventual coolability, and containment pressurization phenomena. Finally, Task III is concerned with HTGR loss of forced circulation accidents. This report is organized into three major parts corresponding to these three tasks respectively

  7. Impedance calculations for power cables to primary coolant pump motors

    International Nuclear Information System (INIS)

    Hegerhorst, K.B.

    1977-01-01

    The LOFT primary system motor generator sets are located in Room B-239 and are connected to the primary coolant pumps by means of a power cable. The calculated average impedance of this cable is 0.005323 ohms per unit resistance and 0.006025 ohms per unit reactance based on 369.6 kVA and 480 volts. The report was written to show the development of power cable parameters that are to be used in the SICLOPS (Simulation of LOFT Reactor Coolant Loop Pumping System) digital computer program as written in LTR 1142-16 and also used in the pump coastdowns for the FSAR Analysis

  8. Steam generator for a pressurized-water coolant nuclear reactor

    International Nuclear Information System (INIS)

    Schroeder, H.J.; Berger, W.

    1975-01-01

    A description is given of a steam generator which has a vertical cylindrical housing having a steam output outlet, a horizontal tube sheet closing the lower end of this housing, and an inverted U-shaped tube bundle inside of the housing and having vertical inlet and outlet legs with their ends mounted in the tube sheet. Beneath the tube sheet there are inlet and outlet manifolds for the respective ends of the tube bundle so that pressurized-water coolant from a pressurized-water coolant nuclear reactor can be circulated through the tube bundle

  9. Probabilistic analyses of failure in reactor coolant piping

    International Nuclear Information System (INIS)

    Holman, G.S.

    1984-01-01

    LLNL is performing probabilistic reliability analyses of PWR and BWR reactor coolant piping for the NRC Office of Nuclear Regulatory Research. Specifically, LLNL is estimating the probability of a double-ended guillotine break (DEGB) in the reactor coolant loop piping in PWR plants, and in the main stream, feedwater, and recirculation piping of BWR plants. In estimating the probability of DEGB, LLNL considers two causes of pipe break: pipe fracture due to the growth of cracks at welded joints (direct DEGB), and pipe rupture indirectly caused by the seismically-induced failure of critical supports or equipment (indirect DEGB)

  10. Control rod drive mechanism stator loss of coolant test

    International Nuclear Information System (INIS)

    Besel, L.; Ibatuan, R.

    1977-04-01

    This report documents the stator loss of coolant test conducted at HEDL on the lead unit Control Rod Drive Mechanism (CRDM) in February, 1977. The purpose of the test was to demonstrate scram capability of the CRDM with an uncooled stator and to obtain a time versus temperature curve of an uncooled stator under power. Brief descriptions of the test, hardware used, and results obtained are presented in the report. The test demonstrated that the CRDM could be successfully scrammed with no anomalies in both the two-phase and three-phase stator winding hold conditions after the respective equilibrium stator temperatures had been obtained with no stator coolant

  11. Test facility for investigation of heat transfer of promising coolants for the nuclear power industry

    Science.gov (United States)

    Belyaev, I. A.; Sviridov, V. G.; Batenin, V. M.; Biryukov, D. A.; Nikitina, I. S.; Manchkha, S. P.; Pyatnitskaya, N. Yu.; Razuvanov, N. G.; Sviridov, E. V.

    2017-11-01

    The results are presented of experimental investigations into liquid metal heat transfer performed by the joint research group consisting of specialist in heat transfer and hydrodynamics from NIU MPEI and JIHT RAS. The program of experiments has been prepared considering the concept of development of the nuclear power industry in Russia. This concept calls for, in addition to extensive application of water-cooled, water-moderated (VVER-type) power reactors and BN-type sodium cooled fast reactors, development of the new generation of BREST-type reactors, fusion power reactors, and thermonuclear neutron sources. The basic coolants for these nuclear power installations will be heavy liquid metals, such as lead and lithium-lead alloy. The team of specialists from NRU MPEI and JIHT RAS commissioned a new RK-3 mercury MHD-test facility. The major components of this test facility are a unique electrical magnet constructed at Budker Nuclear Physics Institute and a pressurized liquid metal circuit. The test facility is designed for investigating upward and downward liquid metal flows in channels of various cross-sections in a transverse magnetic field. A probe procedure will be used for experimental investigation into heat transfer and hydrodynamics as well as for measuring temperature, velocity, and flow parameter fluctuations. It is generally adopted that liquid metals are the best coolants for the Tokamak reactors. However, alternative coolants should be sought for. As an alternative to liquid metal coolants, molten salts, such as fluorides of lithium and beryllium (so-called FLiBes) or fluorides of alkali metals (so-called FLiNaK) doped with uranium fluoride, can be used. That is why the team of specialists from NRU MPEI and JIHT RAS, in parallel with development of a mercury MHD test facility, is designing a test facility for simulating molten salt heat transfer and hydrodynamics. Since development of this test facility requires numerical predictions and verification

  12. Oxidation of SiC cladding under Loss of Coolant Accident (LOCA) conditions in LWRs

    International Nuclear Information System (INIS)

    Lee, Y.; Yue, C.; Arnold, R. P.; McKrell, T. J.; Kazimi, M. S.

    2012-01-01

    An experimental assessment of Silicon Carbide (SiC) cladding oxidation rate in steam under conditions representative of Loss of Coolant Accidents (LOCA) in light water reactors (LWRs) was conducted. SiC oxidation tests were performed with monolithic alpha phase tubular samples in a vertical quartz tube at a steam temperature of 1140 deg. C and steam velocity range of 1 to 10 m/sec, at atmospheric pressure. Linear weight loss of SiC samples due to boundary layer controlled reaction of silica scale (SiO 2 volatilization) was experimentally observed. The weight loss rate increased with increasing steam flow rate. Over the range of test conditions, SiC oxidation rates were shown to be about 3 orders of magnitude lower than the oxidation rates of zircaloy 4. A SiC volatilization correlation for developing laminar flow in a vertical channel is formulated. (authors)

  13. Safeguarding of emergency core cooling in case of loss-of-coolant accidents with insulation material release

    International Nuclear Information System (INIS)

    Pointner, W.; Broecker, A.

    2012-01-01

    The report on safeguarding of emergency core cooling in case of loss-of-coolant accidents with insulation material release covers the following issues: assessment of the relevant status for PWR, evaluation of the national and international (USA, Canada, France) status, actualization of recommendations, transferability from PWR to BWR. Generic studies on the core cooling capability in case of insulation material release in BWR-type reactors were evaluated.

  14. THERMOSS: a thermohydraulic model of flow stagnation in a horizontal fuel channel

    International Nuclear Information System (INIS)

    Gulshani, P.; Caplan, M.Z.; Spinks, N.J.

    1984-01-01

    Following a postulated inlet-side small break in the CANDU reactor, emergency coolant is injected to refull the horizontal fuel channels and remove the decay heat. As part of the accident analysis, the effects of loss of forced circulation during the accident are predicted. A break size exists for which, at the end of pump rundown, the break force balances the natural circulation force and the channel flow is reduced to near zero. The subcooled, stagnant channel condition is referred to as the standing-start condition. Subsequently, the channel coolant boils and stratifies. Eventually the steam flow from the channel heats up the endfitting to the saturation temperature and reaches the vertical feeder. The resulting buoyancy-induced flow then refills the channel. One dimensional, two-fluid conservation equations are solved in closed form to predict the duration of stagnation. In this calculation the channel water level is an important intermediate variable because it determines the amount of steam production

  15. Coolant flow monitoring in a PWR core using noise analysis

    International Nuclear Information System (INIS)

    Kostic, Lj.

    1992-01-01

    Experimental investigations of the neutron and temperature noise field have been performed in the 1350 MW PWR nuclear power plant. Evaluation in the low frequency range, where both feedback effects and different thermohydraulics phenomena are dominant, succeeded in measuring the coolant velocity. This is important for determination and localization of essential deviations and possible anomalies. (author)

  16. SMART core power control method by coolant temperature variation

    International Nuclear Information System (INIS)

    Lee, Chung Chan; Cho, Byung Oh

    2001-08-01

    SMART is a soluble boron-free integral type pressurized water reactor. Its moderator temperature coefficient (MTC) is strongly negative throughout the cycle. The purpose of this report is how to utilize the primary coolant temperature as a second reactivity control system using the strong negative MTC. The reactivity components associated with reactor power change are Doppler reactivity due to fuel temperature change, moderator temperature reactivity and xenon reactivity. Doppler reactivity and moderator temperature reactivity take effects almost as soon as reactor power changes. On the other hand, xenon reactivity change takes more than several hours to reach an equilibrium state. Therefore, coolant temperature at equilibrium state is chosen as the reference temperature. The power dependent reference temperature line is limited above 50% power not to affect adversely in reactor safety. To compensate transient xenon reactivity, coolant temperature operating range is expanded. The suggested coolant temperature operation range requires minimum control rod motion for 50% power change. For smaller power changes such as 25% power change, it is not necessary to move control rods to assure that fuel design limits are not exceeded

  17. Nanofluid as coolant for grinding process: An overview

    Science.gov (United States)

    Kananathan, J.; Samykano, M.; Sudhakar, K.; Subramaniam, S. R.; Selavamani, S. K.; Manoj Kumar, Nallapaneni; Keng, Ngui Wai; Kadirgama, K.; Hamzah, W. A. W.; Harun, W. S. W.

    2018-04-01

    This paper reviews the recent progress and applications of nanoparticles in lubricants as a coolant (cutting fluid) for grinding process. The role of grinding machining in manufacturing and the importance of lubrication fluids during material removal are discussed. In grinding process, coolants are used to improve the surface finish, wheel wear, flush the chips and to reduce the work-piece thermal deformation. The conventional cooling technique, i.e., flood cooling delivers a large amount of fluid and mist which hazardous to the environment and humans. Industries are actively looking for possible ways to reduce the volume of coolants used in metal removing operations due to the economical and ecological impacts. Thus as an alternative, an advanced cooling technique known as Minimum Quantity Lubrication (MQL) has been introduced to the enhance the surface finish, minimize the cost, to reduce the environmental impacts and to reduce the metal cutting fluid consumptions. Nanofluid is a new-fangled class of fluids engineered by dispersing nanometre-size solid particles into base fluids such as water, lubrication oils to further improve the properties of the lubricant or coolant. In addition to advanced cooling technique review, this paper also reviews the application of various nanoparticles and their performance in grinding operations. The performance of nanoparticles related to the cutting forces, surface finish, tool wear, and temperature at the cutting zone are briefly reviewed. The study reveals that the excellent properties of the nanofluid can be beneficial in cooling and lubricating application in the manufacturing process.

  18. Numerical experimentation on convective coolant flow in Ghana ...

    African Journals Online (AJOL)

    Numerical experiments on one dimensional convective coolant flow during steady state operation of the Ghana Research Reactor-1 (GHARR-I) were performed to determine the thermal hydraulic parameters of temperature, density and flow rate. The computational domain was the reactor vessel, including the reactor core.

  19. Reactor coolant and associated systems in nuclear power plants

    International Nuclear Information System (INIS)

    1986-01-01

    This Safety Guide outlines the design requirements for the reactor coolant and associated systems (RCAS) and the features required in order to achieve their safety functions. It covers design considerations for various reactor types and encompasses the safety aspects of the functions of the RCAS both during normal operation and following postulated initiating events, and to some extent also for decommissioning

  20. SALLY - a computer programme for the calculation of the dynamical behaviour of a cooling channel

    International Nuclear Information System (INIS)

    Reiche, C.; Ziegenbein, D.

    1978-12-01

    This report describes the mathematical-physical fundamentals and provides a user's manual as well as some test examples for the programme SALLY. SALLY is suitable for the calculation of the dynamical behaviour of a cooling channel under following conditions: point-kinetics, heat transfer in the fuel in radial direction only, reactivity feedback, simple model of the first cooling cycle. Reactivity, coolant inlet temperature or coolant velocity can be disturbed. (author)

  1. Coolant controls of a PEM fuel cell system

    Science.gov (United States)

    Ahn, Jong-Woo; Choe, Song-Yul

    When operating the polymer electrolyte membrane (PEM) fuel cell stack, temperatures in the stack continuously change as the load current varies. The temperature directly affects the rate of chemical reactions and transport of water and reactants. Elevated temperature increases the mobility of water vapor, which reduces the ohmic over-potential in the membrane and eases removal of water produced. Adversely, the high temperature might impose thermal stress on the membrane and cathode catalyst and cause degradation. Conversely, excessive supply of coolants lowers the temperature in the stack and reduces the rate of the chemical reactions and water activity. Corresponding parasitic power dissipated at the electrical coolant pump increases and overall efficiency of the power system drops. Therefore, proper design of a control for the coolant flow plays an important role in ensuring highly reliable and efficient operations of the fuel cell system. Herein, we propose a new temperature control strategy based on a thermal circuit. The proposed thermal circuit consists of a bypass valve, a radiator with a fan, a reservoir and a coolant pump, while a blower and inlet and outlet manifolds are components of the air supply system. Classic proportional and integral (PI) controllers and a state feedback control for the thermal circuit were used in the design. In addition, the heat source term, which is dependent upon the load current, was feed-forwarded to the closed loop and the temperature effects on the air flow rate were minimized. The dynamics and performance of the designed controllers were evaluated and analyzed by computer simulations using developed dynamic fuel cell system models, where a multi-step current and an experimental current profile measured at the federal urban driving schedule (FUDS) were applied. The results show that the proposed control strategy cannot only suppress a temperature rise in the catalyst layer and prevent oxygen starvation, but also reduce the

  2. Organic coolants and their applications to fusion reactors

    International Nuclear Information System (INIS)

    Gierszewski, P.; Hollies, B.

    1986-08-01

    Organic coolants offer a unique set of characteristics for fusion applications. Their advantages include high-temperature (670 K or 400 degrees C) but low-pressure (2 MPa) operation, limited reactivity with lithium and lithium-lead, reduced corrosion and activation, good heat-transfer capabilities, no magnetohydrodynamic (MHD) effects, and an operating temperature range that extends to room temperature. The major disadvantages are decomposition and flammability. However, organic coolants have been extensively studied in Canada, including nineteen years with an operating 60-MW organic-cooled reactor. Proper attention to design and coolant chemistry controlled these potential problems to acceptable levels. This experience provides an extensive data base for design under fusion conditions. The organic fluid characteristics are described in sufficient detail to allow fusion system designers to evaluate organic coolants for specific applications. To illustrate and assess the potential applications, analyses are presented for organic-cooled blankets, first walls, high heat flux components and thermal power cycles. Designs are identified that take advantage of organic coolant features, yet have fluid decomposition related costs that are a small fraction of the overall cost of electricity. For example, organic-cooled first walls make lithium/ferritic steel blankets possible in high-field, high-surface-heat-flux tokamaks, and organic-cooled limiters (up to about 8 MW/m 2 surface heating) are a safer alternative to water cooling for liquid metal blanket concept. Organics can also be used in intermediate heat exchanger loops to provide efficient heat transfer with low reactivity and a large tritium barrier. 55 refs

  3. Responses to Small Break Loss of Coolant Accidents for SMART

    International Nuclear Information System (INIS)

    Bae, Kyoo Hwan; Kim, Hee C.; Chang, Moon H.; Zee, Sung Q.; Kim, Si-Hwan; Lee, Un-Chul

    2004-01-01

    The SMART NSSS adopts the design characteristics of containing most of the primary circuit components, such as the reactor core, main coolant pumps (MCPs), steam generators (SGs), and N 2 gas pressurizer (PZR) in a single leak-tight Reactor Pressure Vessel (RPV) with a relatively large ratio of the primary coolant inventory to the core power compared to the conventional loop-type PWR. Due to these design characteristics, the SMART can fundamentally eliminate the possibility of Large Break Loss of Coolant Accidents (LBLOCAs), improve the natural circulation capability, and assure a sufficient time to mitigate the possibility of core uncover. Also, SMART adopts inherent safety improving features and passive engineered safety systems such as the substantially large negative moderator temperature coefficients, passive residual heat removal system, emergency core cooling system, and a steel-made leak-tight Safeguard Vessel (SV) housing the RPV. This paper presents the results of the safety analyses using a MARS/SMR code for the instantaneous guillotine ruptures of the major pipelines penetrating the RPV. The analysis results, employing conservative initial/boundary conditions and assumptions, show that the safety systems of the SMART basic design adequately remove the core decay heat without causing core uncover for all the cases of the Small Break Loss of Coolant Accidents (SBLOCAs). The sensitivity study results with variable SV conditions show that the reduced SV net free volume can shorten the time for reaching the thermal and mechanical equilibrium condition between the RPV and SV. Under these boundary conditions, the primary system inventory loss can be minimized and the core remains covered for a longer period of time without any makeup of the coolant. (authors)

  4. Fuel-Coolant Interactions: Visualization and Mixing Measurements

    International Nuclear Information System (INIS)

    Loewen, Eric P.; Bonazza, Riccardo; Corradini, Michael L.; Johannesen, Robert E.

    2002-01-01

    Dynamic X-ray imaging of fuel-coolant interactions (FCI), including quantitative measurement of fuel-coolant volume fractions and length scales, has been accomplished with a novel imaging system at the Nuclear Safety Research Center at the University of Wisconsin, Madison. The imaging system consists of visible-light high-speed digital video, low-energy X-ray digital imaging, and high-energy X-ray digital imaging subsystems. The data provide information concerning the melt jet velocity, melt jet configuration, melt volume fractions, void fractions, and spatial and temporal quantification of premixing length scales for a model fuel-coolant system of molten lead poured into a water pool (fuel temperatures 500 to 1000 K; jet diameters 10 to 30 mm; coolant temperatures 20 to 90 deg. C). Overall results indicate that the FCI has three general regions of behavior, with the high fuel-coolant temperature region similar to what might be expected under severe accident conditions. It was observed that the melt jet leading edge has the highest void fraction and readily fragments into discrete masses, which then subsequently subdivide into smaller masses of length scales <10 mm. The intact jet penetrates <3 to 5 jet length/jet diameter before this breakup occurs into discrete masses, which continue to subdivide. Hydrodynamic instabilities can be visually identified at the leading edge and along the jet column with an interfacial region that consists of melt, vapor, and water. This interface region was observed to grow in size as the water pool temperature was increased, indicating mixing enhancement by boiling processes

  5. Coolant voiding analysis following SGTR for an HLMC reactor

    International Nuclear Information System (INIS)

    Farmer, M.T.; Spencer, B.W.; Sienicki, J.J.

    2000-01-01

    Concepts are under development at Argonne National Laboratory for a small, modular, proliferation-resistant nuclear power steam supply system. Of primary interest here is the simplified system design, featuring steam generators that are directly immersed in the lead-bismuth eutectic (LBE) coolant of the primary system. To support the safety case for this design approach, model development and analysis of transient coolant voiding during a postulated guillotine-type steam generator tube rupture event has been carried out. For the current design, the blowdown will occur from the steam generator shell into the ruptured 12.7-mm-inside-diameter tube through which the LBE coolant passes. The steam will expand biaxially in the tube, with a portion of the flow vented upward to eventually expand into the cover-gas region, while the balance of the flow is vented downward as a jet into the surrounding downward-flowing LBE. Coolant freezing is not an issue in this case because of high feedwater temperature in relation to the freezing point of the LBE. The specific objectives of the current work are to (a) determine the penetration behavior of the steam jet into the lower cold-leg region, (b) characterize the resultant void behavior in terms of coherent bubble versus breakup into a size distribution of small bubbles, and (c) characterize the motion of the bubbles with regard to rise to the cover-gas region (via the liner-to-coolant vessel gap) versus downward transport with the flowing LBE and subsequent upflow through the core to the cover-gas region

  6. Improvements of primary coolant shutdown chemistry and reactor coolant system cleanup

    International Nuclear Information System (INIS)

    Gaudard, G.; Gilles, B.; Mesnage, F.; Cattant, F.

    2002-01-01

    In the framework of a radiation exposure management program entitled >, EDF aims at decreasing the mass dosimetry of nuclear power plants workers. So, the annual dose per unit, which has improved from 2.44 m.Sv in 1991 to 1.08 in 2000, should target 0.8 mSv in the year 2005 term in order to meet the results of the best nuclear operators. One of the guidelines for irradiation source term reduction is the optimization of operation parameters, including reactor coolant system (RCS) chemistry in operation, RCS shutdown chemistry and RCS cleanup improvement. This paper presents the EDF strategy for the shutdown and start up RCS chemistry optimization. All the shutdown modes have been reviewed and for each of them, the chemical specifications will be fine tuned. A survey of some US PWRs shutdown practices has been conducted for an acid and reducing shutdown chemistry implementation test at one EDF unit. This survey shows that deviating from the EPRI recommended practice for acid and reducing shutdown chemistry is possible and that critical path impact can be minimized. The paper also presents some investigations about soluble and insoluble species behavior and characterization; the study focuses here on 110m Ag, 122 Sb, 124 Sb and iodine contamination. Concerning RCS cleanup improvement, the paper presents two studies. The first one highlights some limited design modifications that are either underway or planned, for an increased flow rate during the most critical periods of the shutdown. The second one focuses on the strategy EDF envisions for filters and resins selection criteria. Matching the study on contaminants behavior with the study of filters and resins selection criteria should allow improving the cleanup efficiency. (authors)

  7. ANALYSIS OF THE IMPACT PROPERTIES OF THE COOLANT RECOVERY SYSTEM HEAT LOSSES OF COMBINED COMPRESSOR-POWER PLANT ON ITS CHARACTERISTICS

    Directory of Open Access Journals (Sweden)

    Yusha V.L.

    2012-12-01

    Full Text Available The paper presents results of theoretical analysis of the effectiveness of an ideal thermodynamic cycle internal combustion engine combined with an external utilization of exhaust heat. The influence of the properties of the coolant circuit of utilization on its operational parameters and characteristics of the power plant.

  8. Modular Porous Plate Sublimator /MPPS/ requires only water supply for coolant

    Science.gov (United States)

    Rathbun, R. J.

    1966-01-01

    Modular porous plate sublimators, provided for each location where heat must be dissipated, conserve the battery power of a space vehicle by eliminating the coolant pump. The sublimator requires only a water supply for coolant.

  9. Performance investigation of an automotive car radiator operated with nanofluid-based coolants (nanofluid as a coolant in a radiator)

    International Nuclear Information System (INIS)

    Leong, K.Y.; Saidur, R.; Kazi, S.N.; Mamun, A.H.

    2010-01-01

    Water and ethylene glycol as conventional coolants have been widely used in an automotive car radiator for many years. These heat transfer fluids offer low thermal conductivity. With the advancement of nanotechnology, the new generation of heat transfer fluids called, 'nanofluids' have been developed and researchers found that these fluids offer higher thermal conductivity compared to that of conventional coolants. This study focused on the application of ethylene glycol based copper nanofluids in an automotive cooling system. Relevant input data, nanofluid properties and empirical correlations were obtained from literatures to investigate the heat transfer enhancement of an automotive car radiator operated with nanofluid-based coolants. It was observed that, overall heat transfer coefficient and heat transfer rate in engine cooling system increased with the usage of nanofluids (with ethylene glycol the basefluid) compared to ethylene glycol (i.e. basefluid) alone. It is observed that, about 3.8% of heat transfer enhancement could be achieved with the addition of 2% copper particles in a basefluid at the Reynolds number of 6000 and 5000 for air and coolant respectively. In addition, the reduction of air frontal area was estimated.

  10. On the domestic fuel channel for BWR

    International Nuclear Information System (INIS)

    Fukada, Hiroshi

    1979-01-01

    Kobe Steel Ltd. started the domestic manufacture of fuel channel boxes for BWRs in 1967, and entered the actual production stage four years after that. Since 1976, the mass production system was adopted with the increase of the demand. The requirements about the surface contamination and the dimensional accuracy over whole length are very strict in the fuel channel boxes, moreover, special consideration must be given so as to prevent the deformation in use. The unique working methods such as electron beam welding, high temperature press forming and so on are employed in Kobe Steel Ltd. to satisfy such strict requirements, therefore the quality of the produced fuel channel boxes is superior to imported ones. At present, the fuel channel boxes domestically made by Kobe Steel Ltd. are used for almost all BWRs in Japan. The functions of fuel channel boxes are to flow boiling coolant uniformly upward, to guide control rods, and to increase the rigidity of fuel assembly. The fuel channel boxes are the square tubes of zircaloy 4 of 134.06 mm inside width, 2.03 mm thickness, and 4118 or 4239 mm length. The progress of the development and the features of the fuel channel boxes and the manufacturing processes are described. Zircaloy plates are formed into channels, and two channels are electron beam-welded after the edge preparation, to make a box. Ultrasonic examination and stress relief treatment are applied, and clips and spacers are welded. (Kako, I.)

  11. Safety Analysis of the US Dual Coolant Liquid Lead-Lithium ITER Test Blanket Module

    Energy Technology Data Exchange (ETDEWEB)

    Merrill, Brad; Reyes, Susana; Sawan, Mohamed; Wong, Clement

    2006-07-01

    The US is proposing a prototype of a dual coolant liquid lead-lithium (DCLL) DEMO blanket concept for testing in the International Thermonuclear Experimental Reactor (ITER) as an ITER Test Blanket Module (TBM). Because safety considerations are an integral part of the design process to ensure that this TBM does not adversely impact the safety of ITER, a safety assessment has been conducted for this TBM and its ancillary systems as requested by the ITER project. Four events were selected by the ITER International Team (IT) to address specific reactor safety concerns, such as VV pressurization, confinement building pressure build-up, TBM decay heat removal capability, tritium and activation products release from the TBM system, and hydrogen and heat production from chemical reactions. This paper summarizes the results of this safety assessment conducted with the MELCOR computer code.

  12. Prediction of Hydraulic Performance of a Scaled-Down Model of SMART Reactor Coolant Pump

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Sun Guk; Park, Jin Seok; Yu, Je Yong; Lee, Won Jae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-08-15

    An analysis was conducted to predict the hydraulic performance of a reactor coolant pump (RCP) of SMART at the off-design as well as design points. In order to reduce the analysis time efficiently, a single passage containing an impeller and a diffuser was considered as the computational domain. A stage scheme was used to perform a circumferential averaging of the flux on the impeller-diffuser interface. The pressure difference between the inlet and outlet of the pump was determined and was used to compute the head, efficiency, and break horse power (BHP) of a scaled-down model under conditions of steady-state incompressible flow. The predicted curves of the hydraulic performance of an RCP were similar to the typical characteristic curves of a conventional mixed-flow pump. The complex internal fluid flow of a pump, including the internal recirculation loss due to reverse flow, was observed at a low flow rate.

  13. Large-Eddy Simulation of Turbulent Flow and Heat Transfer in a Mildly Expanded Channel of IFMIF High Flux Test Module

    International Nuclear Information System (INIS)

    Shinji Ebara; Takehiko Yokomine; Akihiko Shimizu

    2006-01-01

    During irradiation test periods in the International Fusion Material Irradiation Facility (IFMIF), irradiated materials must be maintained at constant temperatures because irradiation characteristics of materials have a large dependency on temperature. In the high flux test module of the IFMIF, required performances for temperature control using gas-cooling and heater-heating are especially stringent because available space for temperature control is remarkably restricted due to very small irradiation volume of about 0.5 l. We proposed an alternative design of the test module with advantages of temperature monitoring and temperature uniformity in specimens. This design employs a rectangular duct as the vessel to pack capsules housing specimens compactly into the small irradiation volume. In the vessel the coolant flows between the capsules and vessel wall. In the basic design, both thickness of a vessel wall and a width of cooling channel are considered as 1.0 mm. Since inside the vessel gaseous helium of several atmospheric pressure flows as a coolant and a low vacuum environment is kept outside the vessel for safety requirements and thermal stress is foreseen to appear due to nuclear heating of the vessel itself, the vessel wall is considered to deform readily and this leads expansion of the cooling channels. It is also considered that a slight expansion of the vessel can have severe influence on the cooling performance due to the initial narrow channel width of 1.0 mm. Therefore, it is necessary to estimate cooling performances for the coolant flowing in the deformed channel. We conduct a finite element analysis of turbulent heat transfer in a mildly expanded channel using large-eddy simulation in this study. In a numerical system, fluid is enclosed by three-dimensionally expanded vessel wall and flat capsule wall, and flows into the system with a fully developed velocity profile. In this study, we focus not only on the cooling performances but also on change in

  14. Loss-of-coolant accident analysis of the Savannah River new production reactor design

    International Nuclear Information System (INIS)

    Maloney, K.J.; Pryor, R.J.

    1990-11-01

    This document contains the loss-of-coolant accident analysis of the representative design for the Savannah River heavy water new production reactor. Included in this document are descriptions of the primary system, reactor vessel, and loss-of-coolant accident computer input models, the results of the cold leg and hot leg loss-of-coolant accident analyses, and the results of sensitivity calculations for the cold leg loss-of-coolant accident. 5 refs., 50 figs., 4 tabs

  15. A comparative neutronic analysis of KALIMER breeder core using Na or Pb-Bi coolant

    International Nuclear Information System (INIS)

    Yoo, J. W.; Kim, S. J.; Kim, Y. I.

    2000-01-01

    A comparative neutronic study has been conducted on KALIMER breeder core according to the replacement of sodium coolant by Pb-Bi coolant. Since the atomic weight of Pb and Bi is about 9 times heavier than that of Na, the energy loss by neutron colliding with Pb-Bi nucleus will be very small. Therefore, the reactor with Pb-Bi coolant will have a harder neutron spectrum than that with Na coolant. Consequently, the breeding ratio and burnup reactivity swing is expected to be enhanced. In addition, when Pb-Bi coolant is voided, a negative coolant void coefficient can be obtained by the net effects of smaller spectrum hardening and large neutron leakage. As a result, the breeding ratio was increased from 1.18 to 1.23 and burnup reactivity swing was reduced from 631 pcm to 150 pcm. When the coolant in the whole region of active core is voided, the coolant void coefficient was found to be -539 and -264 pcm at BOEC and EOEC, respectively. In the local voided case, the smaller coolant void coefficient was obtained than that of Na coolant. Accordingly, the use of Pb-Bi coolant in KALIMER gives an advantage of higher breeding ratio, smaller burnup reactivity swing and negative coolant void coefficient without any significant degradation of nuclear performance

  16. Method and apparatus for suppressing water-solid overpressurization of coolant in nuclear reactor power apparatus

    International Nuclear Information System (INIS)

    Aanstad, O.J.; Sklencar, A.M.

    1983-01-01

    A reactor-coolant relief valve is opened for increase in mass influx if the rate of change of coolant pressure exceeds a setpoint during a predetermined interval, if, during this interval, the coolant temperature is less than a setpoint and if the level of the fluid in the pressurizer is above a predetermined setpoint (water-solid state). (author)

  17. Uncertainty analysis for hot channel

    International Nuclear Information System (INIS)

    Panka, I.; Kereszturi, A.

    2006-01-01

    The fulfillment of the safety analysis acceptance criteria is usually evaluated by separate hot channel calculations using the results of neutronic or/and thermo hydraulic system calculations. In case of an ATWS event (inadvertent withdrawal of control assembly), according to the analysis, a number of fuel rods are experiencing DNB for a longer time and must be regarded as failed. Their number must be determined for a further evaluation of the radiological consequences. In the deterministic approach, the global power history must be multiplied by different hot channel factors (kx) taking into account the radial power peaking factors for each fuel pin. If DNB occurs it is necessary to perform a few number of hot channel calculations to determine the limiting kx leading just to DNB and fuel failure (the conservative DNBR limit is 1.33). Knowing the pin power distribution from the core design calculation, the number of failed fuel pins can be calculated. The above procedure can be performed by conservative assumptions (e.g. conservative input parameters in the hot channel calculations), as well. In case of hot channel uncertainty analysis, the relevant input parameters (k x, mass flow, inlet temperature of the coolant, pin average burnup, initial gap size, selection of power history influencing the gap conductance value) of hot channel calculations and the DNBR limit are varied considering the respective uncertainties. An uncertainty analysis methodology was elaborated combining the response surface method with the one sided tolerance limit method of Wilks. The results of deterministic and uncertainty hot channel calculations are compared regarding to the number of failed fuel rods, max. temperature of the clad surface and max. temperature of the fuel (Authors)

  18. Feasibility investigations on a novel micro-manufacturing process for fabrication of fuel cell bipolar plates: Internal pressure-assisted embossing of micro-channels with in-die mechanical bonding

    Energy Technology Data Exchange (ETDEWEB)

    Koc, Muammer [NSF I/UCR Center for Precision Forming (CPF), Department of Mechanical Engineering, Virginia Commonwealth University (VCU), Richmond, VA (United States); Mahabunphachai, Sasawat [NSF I/UCR Center for Precision Forming (CPF), Department of Mechanical Engineering, Virginia Commonwealth University (VCU), Richmond, VA (United States); Department of Mechanical Engineering, University of Michigan, Ann Arbor, MI (United States)

    2007-10-25

    In this paper, we present the results of our studies on conceptual design and feasibility experiments towards development of a novel hybrid manufacturing process to fabricate fuel cell bipolar plates that consists of multi-array micro-channels on a large surface area. The premises of this hybrid micro-manufacturing process stem from the use of an internal pressure-assisted embossing process (cold or warm) combined with mechanical bonding of double bipolar plates in a single-die and single-step operation. Such combined use of hydraulic and mechanical forming forces and in-process bonding will (a) enable integrated forming of micro-channels on both surfaces (as anode and cathode flow fields) and at the middle (as cooling channels), (b) reduce the process steps, (c) reduce variation in dimensional tolerances and surface finish, (d) increase the product quality, (e) increase the performance of fuel cell by optimizing flow-field designs and ensuring consistent contact resistance, and (f) reduce the overall stack cost. This paper explains two experimental investigations that were performed to characterize and evaluate the feasibility of the conceptualized manufacturing process. The first investigation involved hydroforming of micro-channels using thin sheet metals of SS304 with a thickness of 51 {mu}m. The width of the channels ranged from 0.46 to 1.33 mm and the height range was between 0.15 and 0.98 mm. Our feasibility experiments resulted in that different aspect ratios of micro-channels could be fabricated using internal pressure in a controllable manner although there is a limit to very sharp channel shapes (i.e., high aspect ratios with narrow channels). The second investigation was on the feasibility of mechanical bonding of thin sheet metal blanks. The effects of different process and material variables on the bond quality were studied. Successful bonding of various metal blanks (Ni201, Al3003, and SS304) was obtained. The experimental results from both

  19. International

    International Nuclear Information System (INIS)

    Anon.

    1997-01-01

    This rubric reports on 10 short notes about international economical facts about nuclear power: Electricite de France (EdF) and its assistance and management contracts with Eastern Europe countries (Poland, Hungary, Bulgaria); Transnuclear Inc. company (a 100% Cogema daughter company) acquired the US Vectra Technologies company; the construction of the Khumo nuclear power plant in Northern Korea plays in favour of the reconciliation between Northern and Southern Korea; the delivery of two VVER 1000 Russian reactors to China; the enforcement of the cooperation agreement between Euratom and Argentina; Japan requested for the financing of a Russian fast breeder reactor; Russia has planned to sell a floating barge-type nuclear power plant to Indonesia; the control of the Swedish reactor vessels of Sydkraft AB company committed to Tractebel (Belgium); the renewal of the nuclear cooperation agreement between Swiss and USA; the call for bids from the Turkish TEAS electric power company for the building of the Akkuyu nuclear power plant answered by three candidates: Atomic Energy of Canada Limited (AECL), Westinghouse (US) and the French-German NPI company. (J.S.)

  20. Preservation of meandering river channels in uniformly aggrading channel belts

    NARCIS (Netherlands)

    Lageweg, W.I. van de; Schuurman, F.; Cohen, K.M.; Dijk, W.M. van; Shimizu, Y.; Kleinhans, M.G.

    2016-01-01

    Channel belt deposits from meandering river systems commonly display an internal architecture of stacked depositional features with scoured basal contacts due to channel and bedform migration across a range of scales. Recognition and correct interpretation of these bounding surfaces is essential to

  1. AGING MANAGMENT OF REACTOR COOLANT SYSTEM MECHANICAL COMPONENTS FOR LICENSE RENEWAL

    International Nuclear Information System (INIS)

    SUBUDHI, M.; MORANTE, R.; LEE, A.D.

    2002-01-01

    The reactor coolant system (RCS) mechanical components that require an aging management review for license renewal include the primary loop piping and associated connections to other support systems, reactor vessel, reactor vessel internals, pressurizer. steam generators, reactor coolant pumps, and all other inter-connected piping, pipe fittings, valves, and bolting. All major RCS components are located inside the reactor building. Based on the evaluation findings of recently submitted license renewal applications for pressurized water reactors, this paper presents the plant programs and/or activities proposed by the applicants to manage the effects of aging. These programs and/or activities provide reasonable assurance that the intended function(s) of these mechanical components will be maintained for the period of extended operation. The license renewal application includes identification of RCS subcomponents that are within the scope of license renewal and are vulnerable to age-related degradation when exposed to environmental and operational conditions. determination of the effects of aging on their intended safety functions. and implementation of the aging management programs and/or activities including both current and new programs. Industry-wide operating experience, including generic communication by the NRC, is part of the aging management review for the RCS components. In addition, this paper discusses time-limited aging analyses associated with neutron embrittlement of the reactor vessel beltline region and thermal fatigue

  2. Phenomena occurring in the reactor coolant system during severe core damage accidents

    International Nuclear Information System (INIS)

    Malinauskas, A.P.

    1989-01-01

    The reactor coolant system (RCS) of a nuclear power plant consists of the reactor pressure vessel and the piping and associated components that are required for the continuous circulation of the coolant which is used to maintain thermal equilibrium throughout the system. In the event of an accident, the RCS also serves as one of several barriers to the escape of radiotoxic material into the biosphere. In contrast to normal operating conditions, severe core damage accidents are characterized by significant temporal and spatial variations in heat and mass fluxes, and by eventual geometrical changes within the RCS. Furthermore, the difficulties in describing the system in the severe accident mode are compounded by the occurrence of chemical reactions. These reactions can influence both the thermal and the mass transport behavior of the system. In addition, behavior of the reactor vessel internals and of materials released from the core region (especially the radioactive fission products) in the course of the accident likewise become of concern to the analyst. This report addresses these concerns. 9 refs., 1 tab

  3. Numerical Simulation of a Coolant Flow and Heat Transfer in a Pebble Bed Reactor

    International Nuclear Information System (INIS)

    In, Wang-Kee; Kim, Min-Hwan; Lee, Won-Jae

    2008-01-01

    Pebble Bed Reactor(PBR) is one of the very high temperature gas cooled reactors(VHTR) which have been reviewed in the Generation IV International Forum as potential sources for future energy needs, particularly for a hydrogen production. The pebble bed modular reactor(PBMR) exhibits inherent safety features due to the low power density and the large amount of graphite present in the core. PBR uses coated fuel particles(TRISO) embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the PBR core during a reactor operation and the coolant flows around randomly distributed spheres. For the reliable operation and the safety of the PBR, it is important to understand the coolant flow structure and the fuel pebble temperature in the PBR core. There have been few experimental and numerical studies to investigate the fluid and heat transfer phenomena in the PBR core. The objective of this paper is to predict the fluid and heat transfer in the PBR core. The computational fluid dynamics (CFD) code, STAR-CCM+(V2.08) is used to perform the CFD analysis using the design data for the PBMR400

  4. Numerical analysis of coolant mixing in the pressure vessel of WWER-440 type nuclear reactors

    International Nuclear Information System (INIS)

    Boros, I.; Aszodi, A.

    2003-01-01

    The precise description of the coolant mixing processes taking place in the reactor pressure vessel (RPV) of pressurized water nuclear reactors has an essential importance during power operation, as well as in case of incidental or accidental conditions. In this paper the detailed CFD model of the pressure vessel of a WWER-440 type reactor and calculations performed with this RPV model are presented. The CFD model of the pressure vessel contains all the important internal structural elements of the RPV. Sensitivity study on the effect of these elements was also carried out. Both steady-state and transient calculation were performed using the CFD code CFX-5.5.1. The results of the steady-state calculations give the so called mixing factors, i.e. the effect of each single primary loop at the core inlet. The mixing factors can be given for nominal circumstances (i.e. all main coolant pumps are working) or in case of less than six working MCPs. In order to validate the model the calculated mixing factors are compared with the values measured in the Paks NPP (Authors)

  5. The coupling algorithm between fuel pin and coolant channel in the European Accident Code EAC-2

    International Nuclear Information System (INIS)

    Goethem, G. van; Lassmann, K.

    1989-01-01

    In the field of fast breeder reactors the Commission of the European Communities (CEC) is conducting coordination and harmonisation activities as well as its own research at the CEC's Joint Research Centre (JRC). The development of the modular European Accident Code (EAC) is a typical example of concerted action between EC Member States performed under the leadership of the JRC. This computer code analyzes the initiation phase of low-probability whole-core accidents in LMFBRs with the aim of predicting the rapidity of sodium voiding, the mode of pin failure, the subsequent fuel redistribution and the associated energy release. This paper gives a short overview on the development of the EAC-2 code with emphasis on the coupling mechanism between the fuel behaviour module TRANSURANUS and the thermohydraulics modules which can be either CFEM or BLOW3A. These modules are also briefly described. In conclusion some numerical results of EAC-2 are given: they are recalculations of an unprotected LOF accident for the fictitious EUROPE fast breeder reactor which was earlier analysed in the frame of a comparative exercise performed in the early 80s and organised by the CEC. (orig.)

  6. Fuel cladding interaction with water coolant in power reactors

    International Nuclear Information System (INIS)

    1985-11-01

    Water coolant chemistry and corrosion processes are important factors in reliable operation of NPP's, as at elevated temperatures water is aggressive towards structural materials. Water regimes for commercial Pressurized Water Reactors and Boiling Water Reactors were developed and proved to be satisfactory. Nevertheless, studies of operation experience continue and an amount of new Research and Development work is being conducted for further improvements of technology and better understanding of the physicochemical nature of those processes. In this report information is presented on the IAEA programme on fuel element cladding interaction with water coolant. Some results of this survey and recommendations made by the group of consultants who participated in this work are given as well as recommendations for continuation of this study. Separate abstracts were prepared for 6 papers of this report

  7. N13 - based reactor coolant pressure boundary leakage system

    International Nuclear Information System (INIS)

    Dissing, E.; Marbaeck, L.; Sandell, S.; Svansson, L.

    1980-05-01

    A system for the monitoring of leakage of coolant from the reactor coolant pressure boundary and auxiliary systems to the reactor containment, based on the detection of the N13 content in the atmosphere, has been tested. N13 is produced from the oxyegen of the reactor water via the recoil photon nuclear process H1 + 016 + He4. The generation of N13 is therefore independent of fuel element leakage and of the corrosion product content in the water. In the US AEC regulatory guide 1.45 has a leakage increase of 4 liter/ min been suggested as the response limit. The experiments carried out in Ringhals indicate, that with the accomplishment of minor improvements in the installation, a 4 liter/min leakage to the containment will give rise to a signal with a random error range of +- 0.25 liter/min, 99.7 % confidence level. (author)

  8. Coolant Chemistry Control: Oxygen Mass Transport in Lead Bismuth Eutectic

    International Nuclear Information System (INIS)

    Weisenburger, A.; Mueller, G.; Bruzzese, C.; Glass, A.

    2015-01-01

    In lead-bismuth cooled transmutation systems, oxygen, dissolved in the coolant at defined quantities, is required for stable long-term operation by assuring the formation of protective oxide scales on structural steel surfaces. Extracted oxygen must be permanently delivered to the system and distributed in the entire core. Therefore, coolant chemistry control involves detailed knowledge on oxygen mass transport. Beside the different flow regimes a core might have stagnant areas at which oxygen delivery can only be realised by diffusion. The difference between oxygen transport in flow paths and in stagnant zones is one of the targets of such experiments. To investigate oxygen mass transport in flowing and stagnant conditions, a dedicated facility was designed based on computational fluid dynamics (CFD). CFD also was applied to define the position of oxygen sensors and ultrasonic Doppler velocimetry transducers for flow measurements. This contribution will present the test facility, design relevant CFD calculations and results of first tests performed. (authors)

  9. Health physics aspects of processing EBR-I coolant

    International Nuclear Information System (INIS)

    Burke, L.L.; Thalgott, J.O.; Poston, J.W. Jr.

    1998-01-01

    The sodium-potassium reactor coolant removed from the Experimental Breeder Reactor Number One after a partial reactor core meltdown had been stored at the Idaho National Engineering and Environmental Laboratory for 40 years. The State of Idaho considered this waste the most hazardous waste stored in the state and required its processing. The reactor coolant was processed in three phases. The first phase converted the alkali metal into a liquid sodium-potassium hydroxide. The second phase converted this caustic to a liquid sodium-potassium carbonate. The third phase solidified the sodium-potassium carbonate into a form acceptable for land disposal. Health physics aspects and dose received during each phase of the processing are discussed

  10. Qualification test of a main coolant pump for SMART pilot

    International Nuclear Information System (INIS)

    Park, Sang Jin; Yoon, Eui Soo; Oh, Hyong Woo

    2006-01-01

    SMART Pilot is a multipurpose small capacity integral type reactor. Main Coolant Pump (MCP) of SMART Pilot is a canned-motor-type axial pump to circulate the primary coolant between nuclear fuel and steam generator in the primary system. The reactor is designed to operate under condition of 310 .deg. C and 14.7 MPa. Thus MCP has to be tested under same operating condition as reactor design condition to verify its performance and safety. In present work, a test apparatus to simulate real operating situations of the reactor has been designed and constructed to test MCP. And then functional tests, performance tests, and endurance tests have been carried out upon a prototype MCP. Canned motor characteristics, homologous head/torque curves, coast-down curves, NPSH curves and life-time performance variations were obtained from the qualification test as well as hydraulic performance characteristics of MCP

  11. Pressurized-water coolant nuclear reactor steam generator

    International Nuclear Information System (INIS)

    Mayer, H.; Schroder, H.J.

    1975-01-01

    A description is given of a pressurized-water coolant nuclear reactor steam generator having a vertical housing for the steam generating water and containing an upstanding heat exchanger to which the pressurized-water coolant passes and which is radially surrounded by a guide jacket supporting a water separator on its top. By thermosiphon action the steam generating water flows upward through and around the heat exchanger within the guide chamber to the latter's top from which it flows radially outwardly and downwardly through a down draft space formed between the outside of the jacket and the housing. The water separator discharges separated water downwardly. The housing has a feedwater inlet opening adjacent to the lower portion of the heat exchanger, providing preheating of the introduced feedwater. This preheated feedwater is conveyed by a duct upwardly to a location where it mixes with the water discharged from the water separator

  12. Multi-rod burst behavior under a loss-of-coolant accident condition, (1)

    International Nuclear Information System (INIS)

    Hashimoto, Masao; Otomo, Takashi; Furuta, Teruo; Kawasaki, Satoru; Uetsuka, Hiroshi

    1980-12-01

    Multi-rod burst tests have been planned since 1977 to estimate quantitative channel restriction during a LOCA transient in LWRs. For this purpose, many bundle tests have been making to burst in a steam in varying a few parameters which influence the degree of channel restriction. The purpose of this report is to provide a background document for final reports to be published in the future. This report includes the results of No. 7805 bundle test, namely temperature, internal pressure, burst behavior of rods and channel restriction of the bundle. (author)

  13. Computational fluid dynamics analyses of lateral heat conduction, coolant azimuthal mixing and heat transfer predictions in a BR2 fuel assembly geometry

    International Nuclear Information System (INIS)

    Tzanos, C.P.; Dionne, B.

    2011-01-01

    To support the analyses related to the conversion of the BR2 core from highly-enriched (HEU) to low-enriched (LEU) fuel, the thermal-hydraulics codes PLTEMP and RELAP-3D are used to evaluate the safety margins during steady-state operation (PLTEMP), as well as after a loss-of-flow, loss-of-pressure, or a loss of coolant event (RELAP). In the 1-D PLTEMP and RELAP simulations, conduction in the azimuthal and axial directions is not accounted. The very good thermal conductivity of the cladding and the fuel meat and significant temperature gradients in the lateral directions (axial and azimuthal directions) could lead to a heat flux distribution that is significantly different than the power distribution. To evaluate the significance of the lateral heat conduction, 3-D computational fluid dynamics (CFD) simulations, using the CFD code STAR-CD, were performed. Safety margin calculations are typically performed for a hot stripe, i.e., an azimuthal region of the fuel plates/coolant channel containing the power peak. In a RELAP model, for example, a channel between two plates could be divided into a number of RELAP channels (stripes) in the azimuthal direction. In a PLTEMP model, the effect of azimuthal power peaking could be taken into account by using engineering factors. However, if the thermal mixing in the azimuthal direction of a coolant channel is significant, a stripping approach could be overly conservative by not taking into account this mixing. STAR-CD simulations were also performed to study the thermal mixing in the coolant. Section II of this document presents the results of the analyses of the lateral heat conduction and azimuthal thermal mixing in a coolant channel. Finally, PLTEMP and RELAP simulations rely on the use of correlations to determine heat transfer coefficients. Previous analyses showed that the Dittus-Boelter correlation gives significantly more conservative (lower) predictions than the correlations of Sieder-Tate and Petukhov. STAR-CD 3-D

  14. Structural integrity analysis of reactor coolant pump flywheel(I)

    International Nuclear Information System (INIS)

    Kim, Young Jin

    1986-01-01

    A reactor coolant pump flywheel is an important machine element to provide the necessary rotational inertia in the event of loss of power to the pumps. This paper attempts to assess the influence of keyways on flywheel stresses and fracture behaviour in detail. The finite element method was used to determine stresses near keyways, including residual stresses, and to establish stress intensity factors for keyway cracks for use in fracture mechanics assessments. (Author)

  15. Evaluation of organic moderator/coolants for fusion breeder blankets

    International Nuclear Information System (INIS)

    Romero, J.B.

    1980-03-01

    Organic coolants have several attractive features for fusion breeder blanket design. Their apparent compatibility with lithium and their ideal physical and nuclear properties allows straight-forward, high performance designs. Radiolytic damage can be reduced to about the same order as comparable fission systems by using multiplier/stripper blanket designs. Tritium recovery from the organic should be straightforward, but additional data is needed to make a better assessment of the economics of the process

  16. Design of Reactor Coolant Pump Seal Online Monitoring System

    Energy Technology Data Exchange (ETDEWEB)

    Ah, Sang Ha; Chang, Soon Heung [KAIST, Daejeon (Korea, Republic of); Lee, Song Kyu [Korea Power Engineering Co., Yongin (Korea, Republic of)

    2008-05-15

    As a part of a Department of Korea Power Engineering Co., (KOPEC) Project, Statistical Quality Control techniques have been applied to many aspects of industrial engineering. An application to nuclear power plant maintenance and control is also presented that can greatly improve plant safety. As a demonstration of such an approach, a specific system is analyzed: the reactor coolant pumps (RCPs) and the fouling resistance of heat exchanger. This research uses Shewart X-bar, R charts, Cumulative Sum charts (CUSUM), and Sequential Probability Ratio Test (SPRT) to analyze the process for the state of statistical control. And the Control Chart Analyzer (CCA) has been made to support these analyses that can make a decision of error in process. The analysis shows that statistical process control methods can be applied as an early warning system capable of identifying significant equipment problems well in advance of traditional control room alarm indicators. Such a system would provide operators with enough time to respond to possible emergency situations and thus improve plant safety and reliability. RCP circulates reactor coolant to transfer heat from the reactor to the steam generators. RCP seals are in the pressure part of reactor coolant system, so if it breaks, it can cause small break LOCA. And they are running on high pressure, and high temperature, so they can be easily broken. Since the reactor coolant pumps operate within the containment building, physical access to the pumps occurs only during refueling outages. Engineers depend on process variables transmitted to the control room and through the station's data historian to assess the pumps' condition during normal operation.

  17. Design of Reactor Coolant Pump Seal Online Monitoring System

    International Nuclear Information System (INIS)

    Ah, Sang Ha; Chang, Soon Heung; Lee, Song Kyu

    2008-01-01

    As a part of a Department of Korea Power Engineering Co., (KOPEC) Project, Statistical Quality Control techniques have been applied to many aspects of industrial engineering. An application to nuclear power plant maintenance and control is also presented that can greatly improve plant safety. As a demonstration of such an approach, a specific system is analyzed: the reactor coolant pumps (RCPs) and the fouling resistance of heat exchanger. This research uses Shewart X-bar, R charts, Cumulative Sum charts (CUSUM), and Sequential Probability Ratio Test (SPRT) to analyze the process for the state of statistical control. And the Control Chart Analyzer (CCA) has been made to support these analyses that can make a decision of error in process. The analysis shows that statistical process control methods can be applied as an early warning system capable of identifying significant equipment problems well in advance of traditional control room alarm indicators. Such a system would provide operators with enough time to respond to possible emergency situations and thus improve plant safety and reliability. RCP circulates reactor coolant to transfer heat from the reactor to the steam generators. RCP seals are in the pressure part of reactor coolant system, so if it breaks, it can cause small break LOCA. And they are running on high pressure, and high temperature, so they can be easily broken. Since the reactor coolant pumps operate within the containment building, physical access to the pumps occurs only during refueling outages. Engineers depend on process variables transmitted to the control room and through the station's data historian to assess the pumps' condition during normal operation

  18. Detection of coolant void in lead-cooled fast reactors

    International Nuclear Information System (INIS)

    Wolniewicz, Peter; Håkansson, Ane; Jansson, Peter

    2015-01-01

    Highlights: • We model the ALFRED LFR using different Monte-Carlo codes. • We study the impact on coolant void on the fission cross section in fission chambers. • We develop a methodology to detect coolant void. • We study the impact of detector fissile coating burn-up. • We conclude that the developed methodology may be an attractive complement to LFR monitoring. - Abstract: Previous work (Wolniewicz et al., 2013) has indicated that using fission chambers coated with 242 Pu and 235 U, respectively, can provide the means of detecting changes in the neutron flux that are connected to coolant density changes in a small lead-cooled fast reactor. Such density changes may be due to leakages of gas into the coolant, which, over time, may coalesce to large bubbles implying a high risk of causing severe damage of the core. By using the ratio of the information provided by the two types of detectors a quantity is obtained that is sensitive to these density changes and, to the first order approximation, independent of the power level of the reactor. In this work we continue the investigation of this proposed methodology by applying it to the Advanced LFR European Demonstrator (ALFRED) and using realistic modelling of the neutron detectors. The results show that the methodology may be used to detect density changes indicating the initial stages of a coalescence process that may result in a large bubble. Also, it is shown that under certain circumstances, large bubbles passing through the core could be detected with this methodology

  19. Trends and experiences in reactor coolant pump motors

    International Nuclear Information System (INIS)

    Anon.

    1980-01-01

    A review of the requirements and features of these motors is given as background along with a discussion of trends and experiences. Included are a discussion of thrust bearings and a review of safety related requirements and design features. Primary coolant pump motors are vertical induction motors for pumps that circulate huge quantities of water through the reactor core to carry the heat generated there to steam generator heat exchangers. 4 refs

  20. Experimental and numerical study of hydrodynamics of flow-accelerated corrosion in CANDU primary coolant

    Energy Technology Data Exchange (ETDEWEB)

    Supa-Amornkul, S

    2006-07-01

    In CANDU-6 reactors, the pressurised high-temperature coolant flows through 380 fuel channels passing horizontally through the core. Each end of a fuel channel has a stainless steel annular end-fitting connected to a carbon steel feeder pipe. The outlet coolant, which is at 310{sup o}C with up to 0.30 steam voidage, turns through 90{sup o} as it passes from flow in the annular end-fitting to pipe flow in the feeder via a Grayloc connector. Since 1996, several CANDU stations reported excessive corrosion of their outlet feeder pipes, especially over the first metre, which consists of single or double bends. Early studies related the attack to the hydrodynamics of the coolant and verified that it was a type of flow-accelerated corrosion (FAC). Local shear stress, which is believed to be one of the important factors contributing to FAC, was approximated in the studies with standard empirical correlations. In order to understand the hydrodynamics of the coolant in the outlet feeders, flow-visualisation studies were done at AECL and UNB. At AECL, the observations were confined to a transparent simulation of an outlet feeder bend but at UNB a full-scale transparent test section simulating the geometry and orientation of an outlet feeder bend with its upstream annular end-fitting was fabricated. The feeder consisted of a 54 mm (inside diameter) acrylic pipe with a 73{sup o} bend, connected to an acrylic simulation of a Grayloc flanged fitting and annular end-fitting. The annular end-fitting consisted of an inner pipe, 110 mm outside diameter, and an outer pipe, 150 mm inside diameter, both 1.907 m long. The tests were performed with water and air at atmospheric pressure and room temperature. The maximum water volumetric flow rate was 0.019 m{sup 3}/s and the volume fraction of air varied from 0.05 to 0.56. In characterizing the flow in the UNB study, particular attention was paid to the patterns at the inside of the bend, where a CFD (computational fluid dynamics) code

  1. Experimental and numerical study of hydrodynamics of flow-accelerated corrosion in CANDU primary coolant

    International Nuclear Information System (INIS)

    Supa-Amornkul, S.

    2006-01-01

    In CANDU-6 reactors, the pressurised high-temperature coolant flows through 380 fuel channels passing horizontally through the core. Each end of a fuel channel has a stainless steel annular end-fitting connected to a carbon steel feeder pipe. The outlet coolant, which is at 310 o C with up to 0.30 steam voidage, turns through 90 o as it passes from flow in the annular end-fitting to pipe flow in the feeder via a Grayloc connector. Since 1996, several CANDU stations reported excessive corrosion of their outlet feeder pipes, especially over the first metre, which consists of single or double bends. Early studies related the attack to the hydrodynamics of the coolant and verified that it was a type of flow-accelerated corrosion (FAC). Local shear stress, which is believed to be one of the important factors contributing to FAC, was approximated in the studies with standard empirical correlations. In order to understand the hydrodynamics of the coolant in the outlet feeders, flow-visualisation studies were done at AECL and UNB. At AECL, the observations were confined to a transparent simulation of an outlet feeder bend but at UNB a full-scale transparent test section simulating the geometry and orientation of an outlet feeder bend with its upstream annular end-fitting was fabricated. The feeder consisted of a 54 mm (inside diameter) acrylic pipe with a 73 o bend, connected to an acrylic simulation of a Grayloc flanged fitting and annular end-fitting. The annular end-fitting consisted of an inner pipe, 110 mm outside diameter, and an outer pipe, 150 mm inside diameter, both 1.907 m long. The tests were performed with water and air at atmospheric pressure and room temperature. The maximum water volumetric flow rate was 0.019 m 3 /s and the volume fraction of air varied from 0.05 to 0.56. In characterizing the flow in the UNB study, particular attention was paid to the patterns at the inside of the bend, where a CFD (computational fluid dynamics) code - Fluent 6.1- had

  2. Method of injecting iron ion into reactor coolant

    International Nuclear Information System (INIS)

    Ito, Kazuyuki; Sawa, Toshio; Nishino, Yoshitaka; Adachi, Tetsuro; Osumi, Katsumi.

    1988-01-01

    Purpose: To form iron ions stably and inject them into nuclear reactor coolants with no substantial degradation of the severe water quality conditions for reactor coolants. Method: Iron ions are formed by spontaneous corrosion of iron type materials and electroconductivity is increased with the iron ions. Then, the liquids are introduced into an electrolysis vessel using iron type material as electrodes and, thereafter, incorporation of newly added ions other than the iron ions are prevented by supplying electric current. Further, by retaining the iron type material in the packing vessel by the magnetic force therein, only the iron ions are flow out substantially from the packing vessel while preventing the discharge of iron type materials per se or solid corrosion products and then introduced into the electrolysis vessel. Powdery or granular pure iron or carbon steel is used as the iron type material. Thus, iron ions and hydroxides thereof can be injected into coolants by using reactor water at low electroconductivity and incapable of electrolysis. (Kamimura, M.)

  3. On-line real time gamma analysis of primary coolant

    International Nuclear Information System (INIS)

    Kalechstein, W.; Kupca, S.; Lipsett, J.J.

    1985-10-01

    The evolution of failed fuel monitoring at CANDU power stations is briefly summarized and the design of the latest system for failed fuel detection at a multi-unit power station is described. At each reactor, the system employs a germanium spectrometer combined with a novel spectrum analyzer that simultaneously accumulates the gamma-ray spectrum of the coolant and provides the control room with the concentration of radioisotope activity in the coolant for the gaseous fission products Xe-133, Xe-135, Kr-88 and I-131 in real time and with statistical precision independent of count rate. A gross gamma monitor is included to provide independent information on the level of radioactivity in the coolant and extend the measurement range at very high count rates. A central computer system archives spectra received from all four spectrum analyzers and provides both the activity concentrations and the release rates of specified isotopes. Compared with previous systems the current design offers improvements in that the activity concentrations are updated much more frequently, improved tools are provided for long term surveillance of the heat transport system and the monitor is more reliable and less costly

  4. Design criteria of primary coolant chemistry in SMART-P

    International Nuclear Information System (INIS)

    Choi, Byung Seon; Kim, Ah Young; Kim, Seong Hoon; Yoon, Ju Hyeon; Zee, Sung Qunn

    2005-01-01

    SMART-P differs significantly from commercially designed PWRs. Materials inventories used in SMART-P differ from that at PWRs. All surfaces of the primary circuit with the primary coolant are either made from or plated with stainless steel. The material of steam generator (SG) is also different from that of the standard material of the commercially operating PWRs: titanium alloy for the steam generator tubes. Also, SMART-P primary coolant technology differs from that in PWRs: ammonia is used as a pH raising agent and hydrogen formed due to radiolytic processes is kept in specific range by ammonia dosing. Nevertheless, main objectives of the SMART-P primary coolant are the same as at PWRs: to assure primary system pressure boundary integrity, fuel cladding integrity and to minimize out-of-core radiation buildup. The objective of this work is to introduce the design criteria for the primary water chemistry for SMART-P from the viewpoint of the system characteristics and the chemical design concept

  5. Radiolytic reactions in the coolant of helium cooled reactors

    International Nuclear Information System (INIS)

    Tingey, G.L.; Morgan, W.C.

    1975-01-01

    The success of helium cooled reactors is dependent upon the ability to prevent significant reaction between the coolant and the other components in the reactor primary circuit. Since the thermal reaction of graphite with oxidizing gases is rapid at temperatures of interest, the thermal reactions are limited primarily by the concentration of impurity gases in the helium coolant. On the other hand, the rates of radiolytic reactions in helium are shown to be independent of reactive gas concentration until that concentration reaches a very low level. Calculated steady-state concentrations of reactive species in the reactor coolant and core burnoff rates are presented for current U. S. designed, helium cooled reactors. Since precise base data are not currently available for radiolytic rates of some reactions and thermal reaction rate data are often variable, the accuracy of the predicted gas composition is being compared with the actual gas compositions measured during startup tests of the Fort Saint Vrain high temperature gas-cooled reactor. The current status of these confirmatory tests is discussed. 12 references

  6. Sodium coolant of fast reactors: Experience and problems

    International Nuclear Information System (INIS)

    Kozlov, F.A.; Volchkov, L.G.; Drobyshev, A.V.; Nikulin, M.P.; Kochetkov, L.A.; Alexeev, V.V.

    1997-01-01

    In present report the following subjects are considered: state of the coolant and sodium systems under normal operating condition as well as under decommissioning, disclosing of sodium circuits and liquidation of its consequences, cleaning from sodium and decontamination under repairing works of equipment and circuits. Cleaning of coolant and sodium systems under normal operating conditions and under accident contamination. Cleaning of the equipment under repairing works and during decommissioning from sodium and products of its interaction with water and air. Treatment of sodium waste, taking into account a possibility of sodium fires. It is shown that the state of coolant, cover gas, surfaces of constructive materials which are in contact with them, cleaning systems, formed during installation operation require development of specific technologies. Developed technologies ensured safety operation of sodium cooled installations as in normal operating conditions so in abnormal situations. R and D activities in this field and experience gained provided a solid base for coping with problems arising during decommissioning. Prospective research problems are emphasized where the future efforts should be concentrated in order to improve characteristics of sodium cooled reactors and to make their decommissioning optimal and safe. (author)

  7. Crack stability analysis of low alloy steel primary coolant pipe

    Energy Technology Data Exchange (ETDEWEB)

    Tanaka, T.; Kameyama, M. [Kansai Electric Power Company, Osaka (Japan); Urabe, Y. [Mitsubishi Heavy Industries, Ltd., Takasago (Japan)] [and others

    1997-04-01

    At present, cast duplex stainless steel has been used for the primary coolant piping of PWRs in Japan and joints of dissimilar material have been applied for welding to reactor vessels and steam generators. For the primary coolant piping of the next APWR plants, application of low alloy steel that results in designing main loops with the same material is being studied. It means that there is no need to weld low alloy steel with stainless steel and that makes it possible to reduce the welding length. Attenuation of Ultra Sonic Wave Intensity is lower for low alloy steel than for stainless steel and they have advantageous inspection characteristics. In addition to that, the thermal expansion rate is smaller for low alloy steel than for stainless steel. In consideration of the above features of low alloy steel, the overall reliability of primary coolant piping is expected to be improved. Therefore, for the evaluation of crack stability of low alloy steel piping to be applied for primary loops, elastic-plastic future mechanics analysis was performed by means of a three-dimensioned FEM. The evaluation results for the low alloy steel pipings show that cracks will not grow into unstable fractures under maximum design load conditions, even when such a circumferential crack is assumed to be 6 times the size of the wall thickness.

  8. The operating reliability of the reactor coolant pump

    International Nuclear Information System (INIS)

    Grancy, W.

    1996-01-01

    There is a strong tendency among operating companies and manufacturers of nuclear power stations to further increase safety and operating availability of the plant and of its components. This applies also and particularly to reactor coolant pumps for the primary circuit of nuclear power stations of the type PWR. For 3 decades, ANDRITZ has developed and built such pumps and has attached great importance to the design of the complete pump rotor and of its essential surrounding elements, such as bearing and shaft seal. Apart from questions connected with design functioning of the pump there is one question of top priority: the operating reliability of the reactor coolant pump. The pump rotor (together with the rotor of the drive motor) is the only component within the primary system that permanently rotates at high speed during operation of the reactor plant. Many questions concerning design and configuration of such components cannot be answered purely theoretically, or they can only be answered partly. Therefore comprehensive development work and testing was necessary to increase the operating reliability of the pump rotor itself and of its surrounding elements. This contribution describes the current status of development and, as a focal point, discusses shaft sealing solutions elaborated so far. In this connection also a sealing system will be presented which aims for the first time at using a two-stage mechanical seal in reactor coolant pumps

  9. Feasibility study on the type of KALIMER coolant circulation pump

    International Nuclear Information System (INIS)

    Nam, H. Y.; Kim, Y. K.; Lee, Y. B.; Hwang, J. S.; Choi, S. K.

    1997-07-01

    The characteristics of mechanical pump and electromagnetic (EM) pump for liquid sodium coolant in a liquid metal reactor are compared and analysed as a design concept of KALIMER coolant pumps. The type of coolant circulation pump affects the selection of reactor type, economics, and reliability of reactor. Though the mechanical pump has much application experience and give satisfaction to the reliability of developed reactor type, the possibility of development is limited and its large weight and volume have a negative effect on the design of the economical liquid metal reactor. The large scale electromagnetic pump has not been verified yet, but it is expected to be demonstrated in time. Because the size of EM pump is small relative to the mechanical pump, the compact reactor design is possible. Therefore the selection of EM pump can be one of the methods to improve the economics. Since the shape of EM pump can be varied according to the arrangement of electromagnet coils, a new or unique reactor type can be developed easily in the process of KALIMER development. In the view point of economic LMR development, it is desirable to adopt the electromagnetic pump. (author). 50 refs., 11 tabs., 24 figs

  10. A open-quotes zero wasteclose quotes coolant management strategy

    International Nuclear Information System (INIS)

    Kennicott, M.A.

    1994-01-01

    In June of 1992 the Waste Minimization Program at Rocky Flats Plant (RFP) began a study to determine the best methods of managing water-based industrial metalworking fluids in the plant's Tool Manufacturing Shop. The shop was faced with the challenge of managing fluids that could no longer be disposed of in the traditional manner, through the plant's liquid process waste drains, due to a problem they, were having causing in the Liquid Waste Operations Evaporator. The study's goal was to reduce the waste coolants being generated and to reduce worker exposure to a serious health risk. Results of this study and those of a subsequent study to determine relative compatibilities of various coolants and metals, led to the application of a open-quotes zero wasteclose quotes machine coolant management program. This program is currently saving the generation of 10,000 gallons of liquid waste annually, has eliminated worker exposure to harmful bacteria and biocides, and should result in extended machine tool life, increased product quality, fewer rejected parts, and decreases labor costs

  11. Analysis of molten fuel-coolant interaction during a reactivity-initiated accident experiment

    International Nuclear Information System (INIS)

    El-Genk, M.S.; Hobbins, R.R.

    1981-01-01

    The results of a reactivity-initiated accident experiment, designated RIA-ST-4, are discussed and analyzed with regard to molten fuel-coolant interaction (MFCI). In this experiment, extensive amounts of molten UO 2 fuel and zircaloy cladding were produced and fragmented upon mixing with the coolant. Coolant pressurization up to 35 MPa and coolant overheating in excess of 940 K occurred after fuel rod failure. The initial coolant conditions were similar to those in boiling water reactors during a hot startup (that is, coolant pressure of 6.45 MPa, coolant temperature of 538 K, and coolant flow rate of 85 cm 3 /s). It is concluded that the high coolant pressure recorded in the RIA-ST-4 experiment was caused by an energetic MFCI and was not due to gas release from the test rod at failure, Zr/water reaction, or to UO 2 fuel vapor pressure. The high coolant temperature indicated the presence of superheated steam, which may have formed during the expansion of the working fluid back to the initial coolant pressure; yet, the thermal-to-mechanical energy conversion ratio is estimated to be only 0.3%

  12. Analysis of actual status of works on technology of heavy liquid metal coolants

    International Nuclear Information System (INIS)

    Martynov, P.N.; Askhadullin, R.Sh.; Orlov, Yu.I.; Storozhenko, A.N.

    2014-01-01

    Principle duties in heavy liquid metal coolant technology (HLMC) are provision of the purity of coolant and surfaces of circulation loop for maintenance of design thermohydraulic characteristics, prevention of structural materials corrosion and erosion during long service life and present-day safety precautions on different stages of reactor facility operation. For this reason, current HLMC (Pb-Bi, Pb) technology must include coolant pre-operation and charging; monitoring and regulating of coolant oxygen potential; hydrogen purification of coolant and surfaces of circulation loop from lead oxides-based slags; coolant filtration; reactor cover gas purification from coolant aerosols. The current topical problem is personnel training on the questions of HLMC technology [ru

  13. Feeding and purge systems of coolant primary circuit and coolant secondary circuit control of the I sup(123) target

    International Nuclear Information System (INIS)

    Almeida, G.L. de.

    1986-01-01

    The Radiation Protection Service of IEN (Brazilian-CNEN) detected three faults in sup(123)I target cooling system during operation process for producing sup(123)I: a) non hermetic vessel containing contaminated water from primary coolant circuit; possibility of increasing radioactivity in the vessel due to accumulation of contaminators in cooling water and; situation in region used for personnels to arrange and adjust equipments in nuclear physics area, to carried out maintenance of cyclotron and target coupling in irradiation room. The primary circuit was changed by secondary circuit for target coolant circulating through coil of tank, which receive weater from secondary circuit. This solution solved the three problems simultaneously. (M.C.K.)

  14. Channel box

    International Nuclear Information System (INIS)

    Tanabe, Akira.

    1993-01-01

    In a channel box of a BWR type reactor, protruding pads are disposed in axial position on the lateral side of a channel box opposing to a control rod and facing the outer side portion of the control rod in a reactor core loaded state. In the initial loading stage of fuel assemblies, channel fasteners and spacer pads are abutted against each other in the upper portion between the channel boxes sandwiching the control rod therebetween. Further, in the lower portion, a gap as a channel for the movement of the control rod is ensured by the support of fuel support metals. If the channel box is bent toward the control rod along with reactor operation, the pads are abutted against each other to always ensure the gap through which the control rod can move easily. Further, when the pads are brought into contact with each other, the bending deformation of the channel box is corrected by urging to each other. Thus, the control rod can always be moved smoothly to attain reactor safety operation. (N.H.)

  15. Simplified numerical simulation of hot channel in sodium cooled reactor

    International Nuclear Information System (INIS)

    Fonseca, F. de A.S. da; Silva Filho, E.

    1988-12-01

    The thermal-hydraulic parameter values that restrict the operation of a liquid sodium cooled reactor are not established by the average conditions of the coolant in the reactor core but by the extreme conditions of the hot channel. The present work was developed to analysis of hot channel of a sodium cooled reactor, adapting to this reactor an existent simplified model for hot channel of pressurized water reactor. The model was applied for a standard sodium reactor and the results are considered satisfatory. (author) [pt

  16. Membrane technology for treating of waste nanofluids coolant: A review

    Science.gov (United States)

    Mohruni, Amrifan Saladin; Yuliwati, Erna; Sharif, Safian; Ismail, Ahmad Fauzi

    2017-09-01

    The treatment of cutting fluids wastes concerns a big number of industries, especially from the machining operations to foster environmental sustainability. Discharging cutting fluids, waste through separation technique could protect the environment and also human health in general. Several methods for the separation emulsified oils or oily wastewater have been proposed as three common methods, namely chemical, physicochemical and mechanical and membrane technology application. Membranes are used into separate and concentrate the pollutants in oily wastewater through its perm-selectivity. Meanwhile, the desire to compensate for the shortcomings of the cutting fluid media in a metal cutting operation led to introduce the using of nanofluids (NFs) in the minimum quantity lubricant (MQL) technique. NFs are prepared based on nanofluids technology by dispersing nanoparticles (NPs) in liquids. These fluids have potentially played to enhance the performance of traditional heat transfer fluids. Few researchers have studied investigation of the physical-chemical, thermo-physical and heat transfer characteristics of NFs for heat transfer applications. The use of minimum quantity lubrication (MQL) technique by NFs application is developed in many metal cutting operations. MQL did not only serve as a better alternative to flood cooling during machining operation and also increases better-finished surface, reduces impact loads on the environment and fosters environmental sustainability. Waste coolant filtration from cutting tools using membrane was treated by the pretreated process, coagulation technique and membrane filtration. Nanomaterials are also applied to modify the membrane structure and morphology. Polyvinylidene fluoride (PVDF) is the better choice in coolant wastewater treatment due to its hydrophobicity. Using of polyamide nanofiltration membranes BM-20D and UF-PS-100-100, 000, it resulted in the increase of permeability of waste coolant filtration. Titanium dioxide

  17. Evaluation of the fuel rod integrity in PWR reactors from the spectrometric analysis of the primary coolant

    International Nuclear Information System (INIS)

    Monteiro, Iara Arraes

    1999-02-01

    The main objective of this thesis is to provide a better comprehension of the phenomena involved in the transport of fission products, from the fuel rod to the coolant of a PWR reactor. To achieve this purpose, several steps were followed. Firstly, it was presented a description of the fuel elements and the main mechanisms of fuel rod failure, indicating the most important nuclides and their transport mechanisms. Secondly, taking both the kinetic and diffusion models for the transport of fission products as a basis, a simple analytical and semi-empirical model was developed. This model was also based on theoretical considerations and measurements of coolant's activity, according to internationally adopted methodologies. Several factors are considered in the modelling procedures: intrinsic factors to the reactor itself, factors which depend on the reactor's operational mode, isotope characteristic factors, and factors which depend on the type of rod failure. The model was applied for different reactor's operational parameters in the presence of failed rods. The main conclusions drawn from the analysis of the model's output are relative to the variation on the coolant's water activity with the fuel burnup, the linear operation power and the primary purification rate and to the different behaviour of iodine and noble gases. The model was saturated from a certain failure size and showed to be unable to distinguish between a single big fail and many small ones. (author)

  18. Distributed Nonstationary Heat Model of Two-Channel Solar Air Heater

    International Nuclear Information System (INIS)

    Klychev, Sh. I.; Bakhramov, S. A.; Ismanzhanov, A. I.; Tashiev, N.N.

    2011-01-01

    An algorithm for a distributed nonstationary heat model of a solar air heater (SAH) with two operating channels is presented. The model makes it possible to determine how the coolant temperature changes with time along the solar air heater channel by considering its main thermal and ambient parameters, as well as variations in efficiency. Examples of calculations are presented. It is shown that the time within which the mean-day efficiency of the solar air heater becomes stable is significantly higher than the time within which the coolant temperature reaches stable values. The model can be used for investigation of the performances of solar water-heating collectors. (authors)

  19. Surface channeling

    International Nuclear Information System (INIS)

    Sizmann, R.; Varelas, C.

    1976-01-01

    There is experimental evidence that swift light ions incident at small angles towards single crystalline surfaces can lose an appreciable fraction of their kinetic energy during reflection. It is shown that these projectiles penetrate into the bulk surface region of the crystal. They can travel as channeled particles along long paths through the solid (surface channeling). The angular distribution and the depth history of the re-emerged projectiles are investigated by computer simulations. A considerable fraction of the penetrating projectiles re-emerges from the crystal with constant transverse energy if the angle of incidence is smaller than the critical angle for axial channeling. Analytical formulae are derived based on a diffusion model for surface channeling. A comparison with experimental data exhibits the relevance of the analytical solutions. (Auth.)

  20. Conceptual designing of reduced-moderation water reactor with heavy water coolant

    Energy Technology Data Exchange (ETDEWEB)

    Hibi, Kohki; Shimada, Shoichiro; Okubo, Tsutomu E-mail: okubo@hems.jaeri.go.jp; Iwamura, Takamichi; Wada, Shigeyuki

    2001-12-01

    The conceptual designing of reduced-moderation water reactors, i.e. advanced water-cooled reactors using plutonium mixed-oxide fuel with high conversion ratios more than 1.0 and negative void reactivity coefficients, has been carried out. The core is designed on the concept of a pressurized water reactor with a heavy water coolant and a triangular tight lattice fuel pin arrangement. The seed fuel assembly has an internal blanket region inside the seed fuel region as well as upper and lower blanket regions (i.e. an axial heterogeneous core). The radial blanket fuel assemblies are introduced in a checkerboard pattern among the seed fuel assemblies (i.e. a radial heterogeneous core). The radial blanket region is shorter than the seed fuel region. This study shows that the heavy water moderated core can achieve negative void reactivity coefficients and conversion ratios of 1.06-1.11.

  1. Development of a reactor coolant pump monitoring and diagnostic system. Progress report, June 1982-July 1983

    International Nuclear Information System (INIS)

    Morris, D.J.; Sommerfield, G.A.

    1983-12-01

    The quality of operating data has been insufficient to allow proper evaluation of theoretical reactor coolant (RC) pump seal failure mechanisms. The RC pump monitoring and diagnostic system being developed and installed at Toledo Edison's Davis-Besse Nuclear Power Station will examine the relationship between seal failures and three other variables: The rotordynamic behavior of the pump shaft and related components, the internal conditions and performance of the seals, and the plant or pump operating environment (controlled by the plant operator). Interrelationships between these areas will be developed during the data collection task, scheduled to begin in October 1983 (for a full fuel cycle at Davis-Besse). This report describes system software and hardware development, testing, and installation work performed during this period. Also described is a parallel effort being conducted by a B and W/Byron Jackson/Utility group to improve pump seal performance

  2. Spark Channels

    Energy Technology Data Exchange (ETDEWEB)

    Haydon, S. C. [Department of Physics, University of New England, Armidale, NSW (Australia)

    1968-04-15

    A brief summary is given of the principal methods used for initiating spark channels and the various highly time-resolved techniques developed recently for studies with nanosecond resolution. The importance of the percentage overvoltage in determining the early history and subsequent development of the various phases of the growth of the spark channel is discussed. An account is then given of the recent photographic, oscillographic and spectroscopic investigations of spark channels initiated by co-axial cable discharges of spark gaps at low [{approx} 1%] overvoltages. The phenomena observed in the development of the immediate post-breakdown phase, the diffuse glow structure, the growth of the luminous filament and the final formation of the spark channel in hydrogen are described. A brief account is also given of the salient features emerging from corresponding studies of highly overvolted spark gaps in which the spark channel develops from single avalanche conditions. The essential differences between the two types of channel formation are summarized and possible explanations of the general features are indicated. (author)

  3. Report on the Survey of the Design Review of New Reactor Applications. Volume 4: Reactor Coolant and Associated Systems

    International Nuclear Information System (INIS)

    Downey, Steven; Monninger, John; Nevalainen, Janne; Joyer, Philippe; Koley, Jaharlal; Kawamura, Tomonori; Chung, Yeon-Ki; Haluska, Ladislav; Persic, Andreja; Reierson, Craig; Monninger, John; Choi, Young-Joon; )

    2017-01-01

    At the tenth meeting of the Committee on Nuclear Regulatory Activities (CNRA) Working Group on the Regulation of New Reactors (WGRNR) in March 2013, the Working Group agreed to present the responses to the Second Phase, or Design Phase, of the licensing process survey as a multi-volume text. As such, each report will focus on one of the eleven general technical categories covered in the survey. The general technical categories were selected to conform to the topics covered in the International Atomic Energy Agency (IAEA) Safety Guide GS-G-4.1. This report provides a discussion of the survey responses related to the Reactor Coolant and Associated Systems category. The Reactor Coolant and Associated Systems category includes the following technical topics: overpressure protection, reactor coolant pressure boundary, reactor vessel, and design of the reactor coolant system. For each technical topic, the member countries described the information provided by the applicant, the scope and level of detail of the technical review, the technical basis for granting regulatory authorisation, the skill sets required and the level of effort needed to perform the review. Based on a comparison of the information provided by the member countries in response to the survey, the following observations were made: - Although the description of the information provided by the applicant differs in scope and level of detail among the member countries that provided responses, there are similarities in the information that is required. - All of the technical topics covered in the survey are reviewed in some manner by all of the regulatory authorities that provided responses. - It is common to consider operating experience and lessons learnt from the current fleet during the review process. - The most commonly and consistently identified technical expertise needed to perform design reviews related to this category are mechanical engineering and materials engineering. The complete survey

  4. Analysis of an Advanced Test Reactor Small-Break Loss-of-Coolant Accident with an Engineered Safety Feature to Automatically Trip the Primary Coolant Pumps

    International Nuclear Information System (INIS)

    Polkinghorne, Steven T.; Davis, Cliff B.; McCracken, Richard T.

    2000-01-01

    A new engineered safety feature that automatically trips the primary coolant pumps following a low-pressure reactor scram was recently installed in the Advanced Test Reactor (ATR). The purpose of this engineered safety feature is to prevent the ATR's surge tank, which contains compressed air, from emptying during a small-break loss-of-coolant accident (SBLOCA). If the surge tank were to empty, the air introduced into the primary coolant loop could potentially cause the performance of the primary and/or emergency coolant pumps to degrade, thereby reducing core thermal margins. Safety analysis performed with the RELAP5 thermal-hydraulic code and the SINDA thermal analyzer shows that adequate thermal margins are maintained during an SBLOCA with the new engineered safety feature installed. The analysis also shows that the surge tank will not empty during an SBLOCA even if one of the primary coolant pumps fails to trip

  5. Numerical study of the thermo-hydraulic behavior for the Candu type fuel channel

    International Nuclear Information System (INIS)

    Lazaro, Pavel Gabriel; Balas Ghizdeanu, Elena Nineta

    2008-01-01

    Candu type reactors use fuel channel in a horizontal lattice. The fuel bundles are positioned in two Zircaloy tubes: the pressure tube surrounded by calandria tube. Inside the pressure tube the coolant heavy water flows. The coolant reaches high temperatures and pressures. Due to irregular neutron spatial distribution, the fuel channel stress differs from one channel to other. In one improbable event of severe accident, the fuel channel behaves differently according to its normal function history. Over the years, there have been many research projects trying to analyze thermal hydraulic performance of the design and to add some operational improvements in order to achieve an efficient thermal hydraulic distribution. This paper discusses the thermo hydraulic behavior (influence of the temperature and velocity distribution) of the most solicited channel, simulated with Fluent 6.X. Code. Moreover it will be commented the results obtained using different models and mesh applied. (authors)

  6. SALLY, Dynamic Behaviour of Reactor Cooling Channel by Point Model

    International Nuclear Information System (INIS)

    Reiche, Chr.; Ziegenbein, D.

    1981-01-01

    1 - Nature of the physical problem solved: The dynamical behaviour of a cooling channel is calculated. Starting from an equilibrium state a perturbation is introduced into the system. That may be an outer reactivity perturbation or a change in the coolant velocity or in the coolant temperature. The neutron kinetics is treated in the framework of the one-point model. The cooling channel consists of a cladded and cooled fuel rod. The temperature distribution is taken into account as an array above a mesh of radial zones and axial layers. Heat transfer is considered in radial direction only, the thermodynamical coupling of the different layers is obtained by the coolant flow. The thermal material parameters are considered to be temperature independent. Reactivity feedback is introduced by means of reactivity coefficients for fuel, canning, and coolant. Doppler broadening is included. The first cooling cycle can be taken into account by a simple model. 2 - Method of solution: The integration of the point kinetics equations is done numerically by the P11 scheme. The system of temperature equations with constant heat resistance coefficients is solved by the method of factorization. 3 - Restrictions on the complexity of the problem: Given limits are: 10 radial fuel zones, 25 axial layers, 6 groups of delayed neutrons

  7. Flow field analysis inside a gas turbine trailing edge cooling channel under static and rotating conditions

    International Nuclear Information System (INIS)

    Armellini, A.; Casarsa, L.; Mucignat, C.

    2011-01-01

    The flow field inside a modern internal cooling channel specifically designed for the trailing edge of gas turbine blades has been experimentally investigated under static and rotating conditions. The passage is characterized by a trapezoidal cross-section of high aspect-ratio and coolant discharge at the blade tip and along the wedge-shaped trailing edge, where seven elongated pedestals are also installed. The tests were performed under engine similar conditions with respect to both Reynolds (Re = 20,000) and Rotation (Ro = 0, 0.23) numbers, while particular care was put in the implementation of proper pressure conditions at the channel exits to allow the comparison between data under static and rotating conditions. The flow velocity was measured by means of 2D and Stereo-PIV techniques applied in the absolute frame of reference. The relative velocity fields were obtained through a pre-processing procedure of the PIV images developed on purpose. Time averaged flow fields inside the stationary and rotating channels are analyzed and compared. A substantial modification of the whole flow behavior due to rotational effects is commented, nevertheless no trace of rotation induced secondary Coriolis vortices has been found because of the progressive flow discharge along the trailing edge. For Ro = 0.23, at the channel inlet the high aspect-ratio of the cross section enhances inviscid flow effects which determine a mass flow redistribution towards the leading edge side. At the trailing edge exits, the distortion of the flow path observed in the channel central portion causes a strong reduction in the dimensions of the 3D separation structures that surround the pedestals.

  8. Flow behaviour in a CANDU horizontal fuel channel from stagnant subcooled initial conditions

    International Nuclear Information System (INIS)

    Caplan, M.Z.; Gulshani, P.; Holmes, R.W.; Wright, A.C.D.

    1984-01-01

    The flow behaviour in a CANDU primary system with horizontal fuel channels is described following a small inlet header break. With the primary pumps running, emergency coolant injection is in the forward direction so that the channel outlet feeders remain warmer than the inlet thereby promoting forward natural circulation. However, the break force opposes the forward driving force. Should the primary pumps run down after the circuit has refilled, there is a break size for which the natural circulation force is balanced by the break force and channels could, theoretically, stagnate. Result of visualization and of full-size channel tests on channel flow behaviour from an initially stagnant channel condition are discussed. After a channel stagnation, the decay power heats the coolant to saturation. Steam is then formed and the coolant stratifies. The steam expands into the subcooled water in the end fitting in a chugging type of flow regime due to steam condensation. After the end fitting reaches the saturation temperature, steam is able to penetrate into the vertical feeder thereby initiating a large buoyancy induced flow which refills the channel. The duration of stagnation is shown to be sensitive to small asymmetries in the initial conditions. A small initial flow can significantly shorten the occurrence and/or duration of boiling as has been confirmed by reactor experience. (author)

  9. Numerical FEM Analyses of primary coolant system at NPP Temelin

    International Nuclear Information System (INIS)

    Junek, L.; Slovacek, M.; Ruzek, L.; Moulis, P.

    2003-01-01

    The main goal of this paper is to inform about the beginning and first steps of implementation of an aging management system at the Temelin NPP. The aging management system is important not only for achieving the current safety level but also for reaching operational reliability of a production unit equipment above the life time assumed by the original design, typically over 40 years. A method to locate the most prominent degradation regions is described. A global shell model of the primary coolant system including all loops and their components - reactor pressure vessel (RPV), steam generator (SG), main coolant pump (MCP), pressurizer, feed water and steam pipelines system is presented. The results of stress-strain analysis on the measured service parameters base are given. Validation of the results is very important and the method to compare the service measurement data with the numerical results is described. The global/local approach is mentioned and discussed. The effects of the complete global system on the individual components under monitoring are transformed into more accurate local spatial models. The local spatial models are used to analyze the gradual lifetime exhaustion of a facility during its service operation. Two spatial local models are presented, viz. feed water nozzle of SG and main coolant piping system T-brunch. The results of analysis of the local spatial models are processed by the neural network computing method, which is also described. The actual gradual damage of the material of the components under monitoring can be obtained based on the analyses performed and on the results from the neural network in combination with the knowledge of the real material characteristics. The procedures applied are included in the DIALIFE diagnostic system

  10. Fuel-coolant interaction-phenomena under prompt burst conditions

    International Nuclear Information System (INIS)

    Jacobs, H.; Young, M.F.; Reil, K.O.

    1979-01-01

    The Prompt Burst Energetics (PBE) experiments conducted at Sandia Laboratories are a series of in-pile tests with fresh uranium oxide or uranium carbide fuel pins in stagnant sodium. Fuel-coolant-interactions in PBE-9S (oxide/sodium system) and PBE-SG2 (carbide/sodium) have been analyzed with the MURTI parametric FCI code. The purpose is to gain insight into possible FCI scenarios in the experiments and sensitivity of results to input parameters. Results are in approximate agreement for the second (triggered) event in PBE-9S (32 MPa peak) and the initial interaction in PBE-SG2

  11. Modelling transient energy release from molten fuel coolant interaction debris

    International Nuclear Information System (INIS)

    Fletcher, D.F.

    1984-05-01

    A simple model of transient energy release in a Molten Fuel Coolant Interaction is presented. A distributed heat transfer model is used to examine the effect of heat transfer coefficient, time available for rapid energy heat transfer and particle size on transient energy release. The debris is assumed to have an Upper Limit Lognormal distribution. Model predictions are compared with results from the SUW series of experiments which used thermite-generated uranium dioxide molybdenum melts released below the surface of a pool of water. Uncertainties in the physical principles involved in the calculation of energy transfer rates are discussed. (author)

  12. In-operation diagnostic system for reactor coolant pump

    International Nuclear Information System (INIS)

    Sugiyama, Mitsunobu; Hasegawa, Ichiro; Kitahara, Hiromichi; Shimamura, Kazuo; Yasuda, Chiaki; Ikeda, Yasuhiro; Kida, Yasuo.

    1996-01-01

    A reactor coolant pump (RCP) is one of the most important rotating machines in the primary loop nuclear power plants. To improve the reliability and of nuclear power plants, a new diagnostic system that enables early detection of RCP faults has been developed. This system is based on continuous monitoring of vibration and other process data. Vibration is an important indicator of mechanical faults providing information on physical phenomena such as changes in dynamic characteristics and excitation forces changes that signal failure or incipient failure. This new system features comparative vibration analysis and simulation to anticipate equipment failure. (author)

  13. PUMP: analog-hybrid reactor coolant hydraulic transient model

    International Nuclear Information System (INIS)

    Grandia, M.R.

    1976-03-01

    The PUMP hybrid computer code simulates flow and pressure distribution; it is used to determine real time response to starting and tripping all combinations of PWR reactor coolant pumps in a closed, pressurized, four-pump, two-loop primary system. The simulation includes the description of flow, pressure, speed, and torque relationships derived through pump affinity laws and from vendor-supplied pump zone maps to describe pump dynamic characteristics. The program affords great flexibility in the type of transients that can be simulated

  14. Graphite beds for coolant filtration at high temperature

    International Nuclear Information System (INIS)

    Heathcock, R.E.; Lacy, C.S.

    1978-01-01

    High temperature filtration will be provided for new Ontario Hydro CANDU heat transport systems. Filtration has been shown to effectively reduce the concentration of circulating corrosion products in our heat transport systems, hence, minimizing the processes of activity transport. This paper will present one option we have for this application; Deep Bed Granular Graphite Filters. The filter system is described by discussing pertinent aspects of its development programme. The compatibility of the filter and the heat transport coolant are demonstrated by results from loop tests, both out- and in-reactor, and by subsequent results from a large filter installation in the NPD NGS heat transport system. (author)

  15. Effects contributing to positive coolant void reactivity in CANDU

    International Nuclear Information System (INIS)

    Whitlock, J.J.; Garland, W.J.; Milgram, M.S.

    1995-01-01

    The lattice cell code WIMS-AECL (Ref. 3) is used to model a typical CANDU lattice cell, using nominal geometric bucklings, the PIJ option, and 69-group Winfrith library. The effect of cell voiding is modeled as a 100% instantaneous removal of coolant from the lattice. This is conservative because of the neglect of time dependence and partial core voiding, considered more plausible in CANDU. Results are grouped into three spectral groups: fast neutrons (0.821 to 10 MeV), epithermal neutrons (0.625 eV to 0.821 MeV), and thermal neutrons (<0.625 eV)

  16. COPDIRC - calculation of particle deposition in reactor coolants

    International Nuclear Information System (INIS)

    Reeks, M.W.

    1982-06-01

    A description is given of a computer code COPDIRC intended for the calculation of the deposition of particulate onto smooth perfectly sticky surfaces in a gas cooled reactor coolant. The deposition is assumed to be limited by transport in the boundary layer adjacent to the depositing surface. This implies that the deposition velocity normalised with respect to the local friction velocity, is an almost universal function of the normalised particle relaxation time. Deposition is assumed similar to deposition in an equivalent smooth perfectly absorbing pipe. The deposition is calculated using 2 models. (author)

  17. Use of flow models to analyse loss of coolant accidents

    International Nuclear Information System (INIS)

    Pinet, Bernard

    1978-01-01

    This article summarises current work on developing the use of flow models to analyse loss-of-coolant accident in pressurized-water plants. This work is being done jointly, in the context of the LOCA Technical Committee, by the CEA, EDF and FRAMATOME. The construction of the flow model is very closely based on some theoretical studies of the two-fluid model. The laws of transfer at the interface and at the wall are tested experimentally. The representativity of the model then has to be checked in experiments involving several elementary physical phenomena [fr

  18. Time-dependent coolant velocity measurements in an operating BWR

    International Nuclear Information System (INIS)

    Luebbesmeyer, D.; Crowe, R.D.

    1980-01-01

    A method to measure time-dependent fluid velocities in BWR-bundle elements by noise analysis of the incore-neutron-detector signals is shown. Two application examples of the new method are given. The time behaviour of the fluid velocity in the bundle element during a scheduled power excursion of the plant. The change of power was performed by changing the coolant flow through the core The apparent change of the fluid velocity due to thermal elongation of the helix-drive of the TIP-system. A simplified mathematical model was derived for this elongation to use as a reference to check the validity of the new method. (author)

  19. Definition of loss-of-coolant accident radiation source

    International Nuclear Information System (INIS)

    1978-02-01

    Meaningful qualification testing of nuclear reactor components requires a knowledge of the radiation fields expected in a loss-of-coolant accident (LOCA). The overall objective of this program is to define the LOCA source terms and compare these with the output of various simulators employed for radiation qualification testing. The basis for comparison will be the energy deposition in a model reactor component. The results of the calculations are presented and some interpretation of the results given. The energy release rates and spectra were validated by comparison with other calculations using different codes since experimental data appropriate to these calculations do not exist

  20. Calculational advance in the modeling of fuel-coolant interactions

    International Nuclear Information System (INIS)

    Bohl, W.R.

    1982-01-01

    A new technique is applied to numerically simulate a fuel-coolant interaction. The technique is based on the ability to calculate separate space- and time-dependent velocities for each of the participating components. In the limiting case of a vapor explosion, this framework allows calculation of the pre-mixing phase of film boiling and interpenetration of the working fluid by hot liquid, which is required for extrapolating from experiments to a reactor hypothetical accident. Qualitative results are compared favorably to published experimental data where an iron-alumina mixture was poured into water. Differing results are predicted with LMFBR materials